WorldWideScience

Sample records for analysis code system

  1. Systems Improved Numerical Fluids Analysis Code

    Science.gov (United States)

    Costello, F. A.

    1990-01-01

    Systems Improved Numerical Fluids Analysis Code, SINFAC, consists of additional routines added to April, 1983, version of SINDA. Additional routines provide for mathematical modeling of active heat-transfer loops. Simulates steady-state and pseudo-transient operations of 16 different components of heat-transfer loops, including radiators, evaporators, condensers, mechanical pumps, reservoirs, and many types of valves and fittings. Program contains property-analysis routine used to compute thermodynamic properties of 20 different refrigerants. Source code written in FORTRAN 77.

  2. Code system for fast reactor neutronics analysis

    International Nuclear Information System (INIS)

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

  3. Systemization of burnup sensitivity analysis code

    International Nuclear Information System (INIS)

    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

  4. Systemization of burnup sensitivity analysis code. 2

    International Nuclear Information System (INIS)

    Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant 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

  5. SINFAC - SYSTEMS IMPROVED NUMERICAL FLUIDS ANALYSIS CODE

    Science.gov (United States)

    Costello, F. A.

    1994-01-01

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

  6. Performance Analysis of Turbo Coded OFDM System

    Directory of Open Access Journals (Sweden)

    Jyoti Chand

    2014-05-01

    Full Text Available Orthogonal Frequency Division Multiplexing (OFDM has become a popular modulation method in high speed wireless communication system. By partitioning a wideband fading channel into a flat narrowband channels, OFDM is able to mitigate the detrimental effects of multipath fading using a simple one- tap equalizer. There is a growing need to quickly transmit information wirelessly and accurately. Engineers have already combine techniques such as OFDM suitable for high data rate transmission with forward error correction (FEC methods over wireless channels. In this thesis, we enhance the system throughput of a working OFDM system by adding turbo codes. The smart use of coding and power allocation in OFDM will be useful to the desired performance at higher data rates. Simulation is to be done over Rayleigh and additive white Gaussian noise (AWGN channels. Here we also compare the two different generator polynomials. This project increases the system throughput at the same time maintaining system performance. The performance is improved by convolution coding [1].

  7. System analysis of bar code laser scanner

    Science.gov (United States)

    Wang, Jianpu; Chen, Zhaofeng; Lu, Zukang

    1996-10-01

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

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

    International Nuclear Information System (INIS)

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

  9. 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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  13. Development of a nuclear power plant system analysis code

    International Nuclear Information System (INIS)

    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

  14. 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)

  15. Validation of IRBURN calculation code system through burnup benchmark analysis

    International Nuclear Information System (INIS)

    Assessment of the reactor fuel composition during the irradiation time, fuel management and criticality safety analysis require the utilization of a validated burnup calculation code system. In this work a newly developed burnup calculation code system, IRBURN, is introduced for the estimation and analysis of the fuel burnup in LWR reactors. IRBURN provides the full capabilities of the Monte Carlo neutron and photon transport code MCNP4C as well as the versatile code for calculating the buildup and decay of nuclides in nuclear materials, ORIGEN2.1, along with other data processing and linking subroutines. This code has the capability of using different depletion calculation schemes. The accuracy and precision of the implemented algorithms to estimate the eigenvalue and spent fuel isotope concentrations are demonstrated by validation against reliable benchmark problem analyses. A comparison of IRBURN results with experimental data demonstrates that the code predicts the spent fuel concentrations within 10% accuracy. Furthermore, standard deviations of the average values for isotopic concentrations including IRBURN data decreases considerably in comparison with the same parameter excluding IRBURN results, except for a few sets of isotopes. The eigenvalue comparison between our results and the benchmark problems shows a good prediction of the k-inf values during the entire burnup history with the maximum difference of 1% at 100 MWd/kgU.

  16. CALOR89: The code system for calorimeter analysis and design

    International Nuclear Information System (INIS)

    As part of a strong experimental high energy physics program, a substantial effort must be involved in calculational analysis of the detector system. This calculational capability must be fundamentally sound and based on previous interchange between theoretical calculations and experimental test programs. The CALOR89[1-6] system for analyzing calorimeters offers a solid approach for investigating all facets of detector systems and has been used in many calculational studies. CALOR89 is one of two major code systems recommended for analysis of SSC detector systems

  17. Sub-channel analysis by RELAP5 system code

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

  18. A new neutronics analysis code system for fast reactors and validation

    International Nuclear Information System (INIS)

    A new neutronics analysis code system has been developed for detailed analysis of fast reactor cores. The code system is composed of a calculation code of effective cross sections, an assembly calculation code based on the method of characteristics, and a full core transport/diffusion calculation code. The validity of the code system is investigated by applying it to the prototype fast reactor Monju, and by comparing the calculation results with measured ones. (author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  1. THYDE-NEU: Nuclear reactor system analysis code

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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)

  3. 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.

  4. Code Based Analysis for Object-Oriented Systems

    Institute of Scientific and Technical Information of China (English)

    Swapan Bhattacharya; Ananya Kanjilal

    2006-01-01

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

  5. Experimental transport analysis code system in JT-60

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  7. 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.

  8. ATHLET-MF: a system analysis code for multi-fluid nuclear systems

    International Nuclear Information System (INIS)

    The system analysis code ATHLET-MF has been developed for the application to the accelerator driven subcritical reactor system (ADS) in Forschungszentrum Karlsruhe on the basis of ATHLET. The new code has an improved numerical scheme to cope with the application to multi-fluid nuclear systems. The code structure was modified in such a way that the user can easily adapt the code for various fluids. For its application to LBE-cooled ADS, correlations of LBE thermophysical properties as well as heat transfer were implemented in ATHLET-MF. The new developed code ATHLETMF is applied to the target cooling systems of the European Experimental ADS (XADS) and 1 MEGAwatt PIlot Experiment (MEGAPIE). Analyses were performed for steady state operation and various transient scenarios to study the thermalhydraulic behavior of the target cooling system. The transient results are presented and discussed. The present study clearly indicates the feasibility of the ATHLET-MF to ADS applications. (authors)

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

  11. Abnormality transient analysis of Monju using a plant system code

    International Nuclear Information System (INIS)

    The objectives of the present study are to analyze plant transients caused by small abnormalities and to find plant parameters by which operators can recognize these small abnormalities. In order to evaluate the plant transient during an abnormal situation in the water system using the plant system code NETFLOW++, the turbine and feedwater systems should be analyzed with good precision. The code is validated using the measured data at Monju. Several abnormalities in the water system are candidates of the present study, e.g., feedwater control valve degradation, feedwater pump degradation, heat transfer degradation due to fouling on heat transfer tubes of the evaporator, loss-of-feedwater-heating, etc. All major components in the tertiary system are included in the calculation model such as the steam generators, the high-pressure turbine, the deaerator, the feedwater pump, the feedwater heaters, the feedwater control valves, the steam control valve, extraction lines and drainpipes. (author)

  12. Automotive Gas Turbine Power System-Performance Analysis Code

    Science.gov (United States)

    Juhasz, Albert J.

    1997-01-01

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

  13. Phase-Space Analysis of Wavefront Coding Imaging Systems

    Institute of Scientific and Technical Information of China (English)

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

    2006-01-01

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

  14. Crack propagation analysis of WECS (Wind Energy Conversion System) components using the LIFE2 computer code

    Science.gov (United States)

    Sutherland, Herbert J.; Schluter, Larry L.

    1988-06-01

    The LIFE2 code is a fatigue/fracture analysis code that is specialized to the analysis of wind energy conversion system components. It is a PC-compatible FORTRAN code that is written in a top-down, modular format. This paper discusses the additions to the code that permit WECS components to be analyzed using linear fracture mechanics. To illustrate the capabilities of the numerical techniques employed here, two example problems are presented.

  15. 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.

  16. Novel Codes Family for Modified Spectral-Amplitude-Coding OCDMA Systems and Performance Analysis

    CERN Document Server

    Noshad, Mohammad

    2010-01-01

    In this paper a novel family of codes for modified spectral-amplitude-coding optical code division multiple access (SAC-OCDMA) is introduced. The proposed codes exist for more number of processing gains comparing to the previously reported codes. In the network using these codes, the number of users can be extended without any essential changes in the previous transmitters. In this study, we propose a construction method for these codes and compare their performance with previously reported codes.

  17. 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.

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

    International Nuclear Information System (INIS)

    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)

  19. Development of dynamic analysis code for HTTR hydrogen production system (Contract research)

    International Nuclear Information System (INIS)

    A heat and mass balance analysis code (N-HYPAC) has been developed to investigate transient behavior in the HTTR hydrogen production system. The code can analyze heat and mass transfer (temperature and mass and pressure distributions of process and helium gases) and behavior of the control system under both static state (case of steady operation) and dynamic state (case of transient operation). Analysis model of helium and process gases from IHX to secondary helium loop and hydrogen production system has been constructed. This report describes analytical flow sheet, construction of the code, basic equations, method to treat the input data, estimation of the preliminary analysis. (author)

  20. Investigation and proposal of the system to affect nuclear fuel type authorization and analysis code certification

    International Nuclear Information System (INIS)

    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)

  1. 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.

  2. A study on the interlink of CANDU safety analysis codes with development of GUI system

    International Nuclear Information System (INIS)

    Recently, safety analysis codes show tendencies to modularize and regard user environment as important. In this manner, the interlinking system of containment analysis code, PRESCON2 and the thermal hydraulics analysis code, CATHENA has been implemented with development of the GUI system in order to maximize user's convenience and save time in analyzing the transient in the heat transport system and containment. Before the development of GUI system, named CANVAS (CANDU Visual Analysis System), two codes are partly corrected to optimize on the PC environment, and carried out its verification. The interlinking system of two codes was executed by introducing three interlinking variables, which are mass flux, mixture enthalpy, and mixture specific volume. By a once-through method, the data from CATHENA output are transferred to PRESCON2 input deck, and then PRESCON2 runs in order. This GUI system provides a dialog box that consists of four tab sheets to control information for the input file and the post processing. After completion of these menus, CANVAS shows the multi-graph on real-time for output data of CATHENA and PRESCON2. This system has extra rooms for additional codes or functions and can be improved step by step. This study can be eventually expected to improve the safety assessment system and technology for CANDU safety analysis fields. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-10-15

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

  4. An integrated multi-functional neutronics calculation and analysis code system: VisualBUS

    International Nuclear Information System (INIS)

    Neutronics calculation and analysis are the bases of reactor physics design, radiation protection, fuel management optimization, nuclear safety analysis, etc. After surveying and evaluating the status and trend of development of neutronics calculation and analysis codes, a network-based integrated multi-functional neutronics calculation and analysis code system has been designed and developed for applications in fusion, fission and various hybrid systems based on the adoption of advanced neutronics calculating approaches and modern computer' software technologies. A series of benchmark tests and applications have shown the maturity and effectiveness of the system. This paper gives a brief overview about main technical features of the system, the benchmark tests and applications. (authors)

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

    International Nuclear Information System (INIS)

    Highlights: ► Modification of system code ATHLET for supercritical water application. ► Development and assessment of a heat transfer package for supercritical water. ► Validation of the modified code at supercritical pressures with the theoretical point-hydraulics model and the SASC code. ► 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 safety system. All the results achieved indicate that the modified ATHLET code has good feasibility in

  6. 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

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

    International Nuclear Information System (INIS)

    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

  8. 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.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

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

  11. HDL code analysis for ASICs in mobile systems

    OpenAIRE

    Wickberg, Fredrik

    2007-01-01

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

  12. Comparison for the interfacial and wall friction models in thermal-hydraulic system analysis codes

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Moon Kyu; Park, Jee Won; Chung, Bub Dong; Kim, Soo Hyung; Kim, See Dal

    2007-07-15

    The average equations employed in the current thermal hydraulic analysis codes need to be closed with the appropriate models and correlations to specify the interphase phenomena along with fluid/structure interactions. This includes both thermal and mechanical interactions. Among the closure laws, an interfacial and wall frictions, which are included in the momentum equations, not only affect pressure drops along the fluid flow, but also have great effects for the numerical stability of the codes. In this study, the interfacial and wall frictions are reviewed for the commonly applied thermal-hydraulic system analysis codes, i.e. RELAP5-3D, MARS-3D, TRAC-M, and CATHARE.

  13. Comparison for the interfacial and wall friction models in thermal-hydraulic system analysis codes

    International Nuclear Information System (INIS)

    The average equations employed in the current thermal hydraulic analysis codes need to be closed with the appropriate models and correlations to specify the interphase phenomena along with fluid/structure interactions. This includes both thermal and mechanical interactions. Among the closure laws, an interfacial and wall frictions, which are included in the momentum equations, not only affect pressure drops along the fluid flow, but also have great effects for the numerical stability of the codes. In this study, the interfacial and wall frictions are reviewed for the commonly applied thermal-hydraulic system analysis codes, i.e. RELAP5-3D, MARS-3D, TRAC-M, and CATHARE

  14. Capability of the coupled code system ATHLET-QUABOX/CUBBOX for safety analysis

    International Nuclear Information System (INIS)

    There exists a tendency to perform accident analysis of NPP by best estimate codes, that decreases the conservatism of performing calculations and allows more realistic simulation of transients. A necessary step for this approach is the comprehensive validation of the computer codes and the coupling of thermo-hydraulic plant system codes with 3D neutronics models. The application of coupled code systems is mandatory for the analysis of accident conditions which are determined by a strong coupling between neutronics of the reactor core and thermo-hydraulics of the primary circuit, especially, when asymmetrical processes take place in the core leading to strongly space-dependent power generation. The paper gives an overview on the development of the coupled 3D neutronics and fluid-dynamic system code in GRS, in particular, the work is presented performed on the basis of the system code ATHLET and the 3D reactor core model QUABOX/CUBBOX, both developed in GRS. In addition, the experiences from applications in accident analysis are summarized and further developments are discussed. (author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  17. RAVE code system for 3-D core non-LOCA accident analysis

    International Nuclear Information System (INIS)

    Full text of publication follows: This paper provides an overview of the application of the Westinghouse updated RAVE three dimensional (3-D) core transient analysis code system for PWR non-LOCA accident analysis. The RAVE code system consists of a linkage of the following USNRC-approved codes: the EPRI RETRAN-02 (RETRAN) system transient analysis code, the Westinghouse SPNOVA (also referred to as ANC-K) reactor core neutron kinetic nodal code, and the EPRI VIPRE-01 (VIPRE) reactor core thermal-hydraulic (T/H) code. The RETRAN code is used for calculating transient conditions in the reactor coolant system (RCS), including reactor vessel, RCS loops, pressurizer and steam generators. RETRAN also models reactor trips, engineering safety feature (ESF) functions, and the control systems. The SPNOVA code is used to perform 3-D core neutronic calculations for core average power and power distributions in the core. Its reactivity feedback calculation is based on transient fluid conditions and fuel temperatures obtained from the VIPRE code. Based on core inlet temperature, inlet flow and core exit pressure from RETRAN, and the nodal nuclear power from SPNOVA, VIPRE provides back to RETRAN transient nodal heat flux in the reactor core region. An effective 3-D analysis requires RETRAN, SPNOVA and VIPRE calculations to be closely linked for the entire reactor core. The linking architecture uses a standard external communication interface protocol for communication among the running programs on the same or different computers. The RAVE code system currently uses the Parallel Virtual Machine (PVM) software for the data transfer. Besides the necessary changes for data transfer, no other changes were made to RETRAN, SPNOVA or VIPRE fundamental code algorithms or solution methods. The RETRAN model in the RAVE system uses the same detailed reactor vessel, RCS loops, pressurizer, and steam generator, and control and protection models as has been licensed for current plant Safety

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-03-15

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

  19. Application of the FORSS sensitivity code system to fast reactor analysis

    International Nuclear Information System (INIS)

    The FORSS Sensitivity Analysis Code System is described in terms of its objectives and present capabilities. An example is made of a problem specified by the Processing Methods Testing Subcommittee of the Code Evaluation Working Group, i.e., the determination of integral parameters, sensitivities to cross-section data, methods and data uncertainties, and required cross-section accuracies for an infinite media of ZPR 6/7 core composition

  20. Performance Analysis of Spectral Amplitude Coding Based OCDMA System with Gain and Splitter Mismatch

    Science.gov (United States)

    Umrani, Fahim A.; Umrani, A. Waheed; Umrani, Naveed A.; Memon, Kehkashan A.; Kalwar, Imtiaz Hussain

    2013-09-01

    This paper presents the practical analysis of the optical code-division multiple-access (O-CDMA) systems based on perfect difference codes. The work carried out use SNR criterion to select the optimal value of avalanche photodiodes (APD) gain and shows how the mismatch in the splitters and gains of the APD used in the transmitters and receivers of network can degrade the BER performance of the system. The investigations also reveal that higher APD gains are not suitable for such systems even at higher powers. The system performance, with consideration of shot noise, thermal noise, bulk and surface leakage currents is also investigated.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  3. Model verification of the system dynamics analysis code NALAP-II

    International Nuclear Information System (INIS)

    Japan Nuclear Energy Safety Organization, JNES, is developing a system dynamics analysis code NALAP-II, to apply analyses for the prototype fast breeder reactor MONJU. The NALAP-II code is a system dynamics analysis code, and consists of a multi-channel and single-pin model coupled with a one-point neutron kinetics model to represent the core system and of a flow network model to describe the heat transport system of the reactor. Major analysis models have been verified based on the MONJU start-up tests such as the turbine-trip tests conducted at the 40% of the rated power and the natural convection test for the SHTS. Results of the examination indicate that the models of the primary and secondary heat transport systems including the steam generator are satisfactory verified. The applicability of the NALAP-II code has also been investigated for typical design-basis events nd for the beyond design-basis events such as the total station blackout. It is emphasized that the passive decay heat removal capability by the natural convection of the sodium, one of the major safety features of the MONJU plant, has been appropriately analyzed and confirmed by using the NALAP-II code. (author)

  4. Computer code DYSAC for identification and dynamics analysis of multivariable systems based on the autoregressive model

    International Nuclear Information System (INIS)

    Usage is described of the computer code DYSAC (Dynamic System Analysis Code) developed for a hybrid computer for the identification and the analysis of system dynamics. A multivariable linear dynamic system is identified based on the autoregressive model using the time series data obtained from a system in operation and the system dynamics thus identified are analyzed. This code includes subroutines for the analysis of step response, frequency response, power spectrum, etc. In order to facilitate handling a large number of various experimental data and to perform the analysis in perspective, considerations for effective utilization of hybrid computer functions and terminal devices are taken in this code, such as; The experimental data record in an analog data recorder are directly input to the analog part of the hybrid computer. The computed results can be plotted on the graphic display and its hard copy is readily available. A series of messages for guidance is given on the display terminal by which the analysis though man-machine interactive computation can be performed. Thus, the required results can be obtained by performing case studies for which necessary parameters are input through the keyboard and the results displayed are checked. (auth.)

  5. Hydra-II, a computer code for hydrothermal analysis of spent fuel storage systems

    Energy Technology Data Exchange (ETDEWEB)

    McCann, R.A.

    1988-03-01

    HYDRA-II is a hydrothermal computer code designed to accurately predict steady-state fluid flow and temperature distributions in spent nuclear fuel storage and transportation systems. The code is capable of three-dimensional analysis of coupled conduction, convection, and thermal radiation problems. It provides a finite difference solution in cartesian coordinates to the equations governing the conservation of mass, momentum, and energy. A cylindrical coordinate system may also be used to enclose the cartesian coordinate system. This exterior coordinate system is useful for modeling cylindrical cask bodies. The basic equations and an overview of the numerics employed in their solution are presented. Selected results from an extensive code verification/validation effort are also presented. Comparisons are made between the results of simulations of a multiassembly storage system and actual experimental data. The effects of backfill gas composition and pressure and cask orientation are illustrated.

  6. Hydra-II, a computer code for hydrothermal analysis of spent fuel storage systems

    International Nuclear Information System (INIS)

    HYDRA-II is a hydrothermal computer code designed to accurately predict steady-state fluid flow and temperature distributions in spent nuclear fuel storage and transportation systems. The code is capable of three-dimensional analysis of coupled conduction, convection, and thermal radiation problems. It provides a finite difference solution in cartesian coordinates to the equations governing the conservation of mass, momentum, and energy. A cylindrical coordinate system may also be used to enclose the cartesian coordinate system. This exterior coordinate system is useful for modeling cylindrical cask bodies. The basic equations and an overview of the numerics employed in their solution are presented. Selected results from an extensive code verification/validation effort are also presented. Comparisons are made between the results of simulations of a multiassembly storage system and actual experimental data. The effects of backfill gas composition and pressure and cask orientation are illustrated

  7. DNBR calculation in digital core protection system by a subchannel analysis code

    International Nuclear Information System (INIS)

    The DNBR calculation uncertainty and DNBR margin were evaluated in digital core protection system by a thermal-hydrualic subchannel analysis code MATRA. A simplified thermal-hydraulic code CETOP is used to calculate on-line DNBR in core protection system at a digital PWR. The DNBR tuning process against a best-estimate subchannel analysis code is required for CETOP to ensure accurate and conservative DNBR calculation but not necessary for MATRA. The DNBR calculations by MATRA and CETOP were performed for a large number of operating condition in Yonggwang nulcear units 3-4 where the digitial core protection system is initially implemented in Korea. MATRA resulted in a less negative mean value (i.e., reduce the overconservatism) and a somewhat larger standard deviation of the DNBR error. The uncertainty corrected minimum DNBR by MATRA was shown to be higher by 1.8% -9.9% that the CETOP DNBR

  8. Development of Validation System for Subchannel Analysis Codes under Steady-State PWR Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hyun; Seo, Kyung Won; Kwon, Hyouk [KAERI, Daejeon (Korea, Republic of)

    2010-10-15

    Subchannel analysis code plays an essential role for the thermal hydraulic design of PWR cores. It calculates subchannel-wise local properties under single- and two-phase flow conditions which are used as input parameters for evaluating design parameters such as minimum DNBR and maximum fuel temperature. A validation system for subchannel analysis codes for steady-state conditions is provided in this study through investigation of validation status of existing subchannel codes, review of licensing guideline for thermal-hydraulic codes, and establishment of thermal hydraulic data base for test bundles. Thermal hydraulic test data has been procured for the following experiments: CNEN 4x4 flow distribution test, CU 4x4 flow and enthalpy distribution test, PNL 7x7 flow blockage test, GE 3x3 two-phase flow and quality distribution test, CE 15x15 inlet jetting test, WH 14x14 inlet blockage test, PNL 2x6 low flow test, ISPRA 4x4 two-phase flow and quality distribution test, FRIGG 36-rod void distribution test, Zion-1 plant FA exit temperature distribution test. Sampling analysis for each test has been conducted by employing KAERI inhouse subchannel analysis code MATRA. The state-of-the-art for subchannel thermal hydraulic analysis models such as void fraction correlations, crossflow and turbulent mixing models, and heat transfer correlations was also investigated

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

    International Nuclear Information System (INIS)

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

  10. Performance Analysis with Space-time coding in MIMO-OFDM Systems with Multiple Antennas

    OpenAIRE

    itendra K umar D aksh; Ravi Mohan; Sumit Sharma

    2013-01-01

    In this paper we show the performance analysis from the previous research and find the drawbacks for the further enhancement. We are considering the case of multiple antennas so that we achieve the higher space-time coding diversity gain and the better performance of system.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1987-01-01

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

  13. Verification of Gas System Analysis Code GAMMA+ with S-CO2 Compressor Test Data

    International Nuclear Information System (INIS)

    In this study, the validation and verification of the GAMMA+ code, which is gas system transient analysis code was conducted by using the obtained SCO2PE experimental results. Before performing a SCO2PE loop transient simulation with the updated GAMMA+ code, major components, the compressor and the heat exchanger, were separately modeled. For the transient experiment, the reduction in cooling event was experimented in the SCO2PE. The results of GAMMA+ code show reasonable agreement with SCO2PE experimental data. However, there is a minute difference between the GAMMA+ prediction and the experimental data, especially at the compressor outlet condition because the heat transfer value from the experimental data was uncertain due to the measurement uncertainties and the CO2 properties near the critical point. To reduce the difference between the experimental data and GAMMA+ results, the modeling of SCO2PE and the methodology for turbomachinery analyses will have to be updated in the future

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

    International Nuclear Information System (INIS)

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

  15. Development and assessment of a subchannel analysis code system for SMART core design

    International Nuclear Information System (INIS)

    A subchannel code system is developed for the thermal-hydraulic analysis of SMART core, and the applicability and accuracy of the code is assessed for various experimental data with rod bundles. MATRA is a subchannel analysis code calculating the enthalpy and flow distribution in fuel assemblies and reactor cores for both steady-state and transient conditions. MATRA has been developed to be run on an IBM PC or HP WS based on the existing CDC CYBER mainframe version of COBRA-IV-I. MATRA has been provided with an improved structure and code functions to give more convenient user environment. Improvement of various models enhances the convergence and accuracy of the code: those include the numerical solution scheme for the crossflow, the void fraction model, and the lateral transport model, and so on. A turbulent mixing model considering void drift phenomenon is devised by employing the two-phase mixing test data under PWR and BWR conditions. MATRA/SR-1 CHF correlation system is developed from local conditions of rod bundle CHF data calculated by MATRA. The optimized 1/8 core lumping models are developed for the analysis of the thermal margins of SMART core at steady-state and transient conditions

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-12-01

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

  18. Development of system analysis code for pyrochemical process using molten salt electrorefining

    International Nuclear Information System (INIS)

    This report describes accomplishment of development of a cathode processor calculation code to simulate the mass and heat transfer phenomena with the distillation process and development of an analytical model for cooling behavior of the pyrochemical process cell on personal computers. The pyrochemical process using molten salt electrorefining would introduce new technologies for new fuels of particle oxide, particle nitride and metallic fuels. The cathode processor calculation code with distillation process was developed. A code validation calculation has been conducted on the basic of the benchmark problem for natural convection in a square cavity. Results by using the present code agreed well for the velocity-temperature fields, the maximum velocity and its location with the benchmark solution published in a paper. The functions have been added to advance the reality in simulation and to increase the efficiency in utilization. The test run has been conducted using the code with the above modification for an axisymmetric enclosed vessel simulating a cathode processor, and the capability of the distillation process simulation with the code has been confirmed. An analytical model for cooling behavior of the pyrochemical process cell was developed. The analytical model was selected by comparing benchmark analysis with detailed analysis on engineering workstation. Flow and temperature distributions were confirmed by the result of steady state analysis. In the result of transient cooling analysis, an initial transient peak of temperature occurred at balanced heat condition in the steady-state analysis. Final gas temperature distribution was dependent on gas circulation flow in transient condition. Then there were different final gas temperature distributions on the basis of the result of steady-state analysis. This phenomenon has a potential for it's own metastable condition. Therefore it was necessary to design gas cooling flow pattern without cooling gas circulation

  19. One-dimensional system analysis code for reflood phase during LOCA

    International Nuclear Information System (INIS)

    A system code named REFLA-1D was developed by coupling the core thermo-hydrodynamic code and the primary system model for the analysis of the reflood phenomena. In order to assess the calculation method of this system code, the results of a base case test and parametric tests, which were run for the conditions of the base case test by varying only one parameter at a time, were compared with the results calculated with the system code. The calculation of the base case test showed a good agreement with the data for the core collapsed liquid level, the quench front elevation, and the heat transfer coefficient near the quench front. The calculation of the parametric test showed a good agreement with the data for the effect of the initial clad temperature and of the peak power, however, a good agreement was not obtained for the effect of system pressure. Further study of the two-phase flow modeling in the core and the quench front correlation against the pressure dependence is necessary for a better prediction of the system behaviors. (author)

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

    International Nuclear Information System (INIS)

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

  1. Modification and application of the system analysis code ATHLET to trans-critical simulations

    International Nuclear Information System (INIS)

    Highlights: ► The pseudo two-phase method is proposed and utilized to modify ATHLET code. ► A smooth transition of void fraction under trans-critical transient can be realized by this method. ► The newly developed ATHLET-SC code can be adopted to simulate the blowdown process of a simplified model. - Abstract: During the loss of coolant accident (LOCA) of supercritical water cooled reactor (SCWR), the pressure in the reactor system will undergo a rapid decrease from supercritical to subcritical condition. This process is called trans-critical transients, which is of crucial importance for the LOCA analysis of SCWR. Using the current version of system code (e.g. ATHLET, REALP), calculation will be terminated due to the abrupt change of void fraction across the critical point (22.064 MPa). To solve this problem, a pseudo two-phase method is proposed by introducing a fictitious region of latent heat (enthalpy of vaporization hfg∗) at pseudo-critical temperatures. A smooth transition of void fraction can be realized by using liquid-field conservation equations at temperatures lower than the pseudo-critical temperature, and vapor-field conservation equations at temperatures higher than the pseudo-critical temperature. Adopting this method, the system code ATHLET is modified to ATHLET-SC mod 2 on the basic of the previous version ATHLET-SC mod 1 modified by Shanghai Jiao Tong University. When the fictitious region of latent heat is kept as a small region, the code can achieve an acceptable accuracy. Moreover, the ATHLET-SC mod 2 code is applied to simulate the blowdown process of a simplified model. The results achieved so far indicate a good applicability of the new modified code for the trans-critical transient.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-01-01

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

  5. Development of sensitivity analysis code for elastic-plastic dynamic response of piping systems

    International Nuclear Information System (INIS)

    A sensitivity analysis code for the elastic-plastic dynamic response of piping systems is developed in the context of implicit time integration schemes. The formulation of sensitivity analysis originates in the perturbation method and retain its advantage. Namely, the decomposed effective stiffness matrix after the iteration for each increment is efficiently used to compute the sensitivity of any parameter involved in the piping system. A residual heat remover of PWR under an earthquake loading is analyzed based on the present method with the isoparametric elbow element proposed by Bathe, and satisfactory results are obtained. (author)

  6. Deterministic model for the analysis of YALINA-booster experiments with the ERANOS code system

    International Nuclear Information System (INIS)

    An experimental program has been launched by the Joint Institute for Power and Nuclear Research - Sosny (JIPNR-Sosny), National Academy of Sciences of Belarus with the purpose to study the physics of Accelerator Driven Systems. This paper gives an overview of the analysis of YALINA-Booster and it provides a detailed description of the adopted approach to create a calculational model based on the use of a deterministic code. (author)

  7. Tokamak Systems Code

    International Nuclear Information System (INIS)

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

  8. Monte Carlo burnup analysis code development and application to an incore thermionic space nuclear power system

    International Nuclear Information System (INIS)

    In the design of the incore thermionic reactor system developed under the Advanced Thermionic Initiative (ATI), the fuel is highly enriched uranium dioxide and the moderating medium is zirconium hydride. The traditional burnup and fuel depletion analysis codes have been found to be inadequate for these calculations, largely because of the material and geometry modeled and because the neutron spectra assumed for the codes such as LEOPARD and ORIGEN do not even closely fit that for a small, thermal reactor using ZrH as moderator. More sophisticated codes such as the transport lattice type code WIMS often lack some materials, such as ZrH. Thus a new method which could accurately calculate the neutron spectrum and the appropriate reaction rates within the fuel element is needed. The method developed utilizes and interconnects the accuracy of the Monte Carlo Neutron/Photon (MCNP) method to calculate reaction rates for the important isotopes, and a time dependent depletion routine to calculate the temporal effects on isotope concentrations. This effort required the modification of MCNP itself to perform the additional task of accomplishing burnup calculations. The modified version called, MCNPBURN, evolved to be a general dual purpose code which can be used for standard calculations as well as for burn-up

  9. Verification of DeCART/CAPP code system for VHTR by HTTR core analysis

    International Nuclear Information System (INIS)

    The DeCART/CAPP code system has been developed and verified against the numerical benchmark calculations for an HTTR. The reference calculations have been carried out by the Monte Carlo McCARD code in which a double heterogeneity model was used. Verification results show that the DeCART/CAPP code system gives less negative MTC and RTC than the McCARD code, and thus the DeCART code overestimates the multiplication factors at states with a high moderator and reflector temperature. However, the DeCART/CAPP code system predicts more negative FTC than McCARD code does. In the depletion calculation for the HTTR single cell and single block, the error of DeCART/CAPP code system increases with the burnup (authors)

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

    Institute of Scientific and Technical Information of China (English)

    Zhu Qi; Li Hao; Feng Guangzeng

    2006-01-01

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

  11. Integration of CFD into systems analysis codes for modeling thermal stratification during SFR transients

    International Nuclear Information System (INIS)

    The whole-plant systems analysis code SAS4A/SASSYS-1 has been coupled with a computational fluid dynamics code to assess the impact of high-fidelity simulations on safety-related performance for a sodium-cooled fast reactor (SFR). With the coupled capability, it is possible to identify critical safety-related phenomenon that cannot be resolved accurately with existing tools. In this work, the impact of coupling is demonstrated by evaluating plenum thermal stratification during a protected loss of flow transient. Stratification is shown to significantly alter core temperatures and flows predicted during natural circulation conditions. Significant temperature and flow impacts were also observed in the secondary coolant system, suggesting that resolving thermal stratification has far-reaching impacts on the whole plant. (author)

  12. Recent code systems at JAERI

    International Nuclear Information System (INIS)

    Recently several code systems have been developed and utilized at Japan Atomic Energy Research Institute. It is not easy to construct and maintain a code system, but the fact is not well known in the researchers. For this reason, it will be very useful to publish informations about design concepts, characteristics, necessary computer features and amounts of invested manpower for the developments of some recent JAERI code systems. In this report, a general view of required manpower on unification of nuclear codes is discussed and four code systems, i.e., SRAC for thermal reactor analysis, TRITON for Tokamak MHD analysis, SPEEDI for emergent environmental dose prediction and RADHEAT for radiation shielding analysis are presented. They are described from aspects of (1) purpose and schedule of development, (2) outline of system, (3) results of benchmark tests, (4) utilized computer features, (5) invested manpower, and (6) desirable computer features. Finally common aspects of four code systems from viewpoint of necessary computer hardwares and softwares are discussed for future development of code systems. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Carlos; Salgado, Jose

    1998-12-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  17. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-12-01

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

  19. 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.

  20. Polaris. A new two-dimensional lattice physics analysis capability for the SCALE code system

    International Nuclear Information System (INIS)

    Polaris is a new 2-dimensional (2-D) lattice physics capability in the SCALE code system for the analysis of light water reactor fuel designs. In this paper, the Polaris calculational methods are summarized and results are provided for a series of computational benchmarks. The summary includes the implementation of the relatively new resonance self-shielding approach called the embedded self-shielding method, the implementation of a new 2-D method-of-characteristics neutron transport solver, and the integration of the SCALE/ORIGEN depletion and decay solver for depleting material compositions. Polaris calculations are compared with reference continuous energy Monte Carlo solutions for a UO2 fuel computational benchmark. Because Polaris is integrated into the SCALE code system, Polaris has been utilized as part of the new SCALE/Sampler code sequence, which provides stochastic uncertainty analysis for the impact of cross-section uncertainties on lattice physics calculations. SCALE/Sampler results are summarized for selected benchmark calculations. (author)

  1. Thermal hydraulics safety analysis of Candu reactor using the RELAP5 system code

    International Nuclear Information System (INIS)

    In this study, the response of CANDU-6 nuclear reactor to several transients are investigated. The simulation of the system is performed by using RELAP5 thermalhydraulic system code. AECL performes the transient simulations of CANDU reactor by using the FIREBIRD code, developed by AECL for thermal hydraulic analysis of CANDU. All analysis for LOCA and ECCS effectiveness were done by using the FIREBIRD code. The investigations concerning the RELAP5 analysis of CANDU system are too few. Better normal operating conditions are achieved, the effect of pipe interconnecting the outlet headers in a loop is observed. It is found that, with the reactor outlet headers interconnected, the system is stable to perturbations but would exhibit divergent pressure, quality and flow oscillations if the interconnection is removed and if the quality at the reactor outlet header region is greater than 1-2% but less than 8%, specific large (100% of flow area) and small (10% of flow area) breaks in both inlet and outlet headers and in the pump suction are analysed. Results indicate that, l00% break in the inlet header has more probability of fuel failure than the same size break in the outlet header. The worst break location is found to be the pump suction with a break size of 100%. Higher void fractions, higher outlet header quality and heat temperatures are observed in the large break transients than that of small break transients. For small break transients, the break location in the inlet header results higher void fraction, outlet header quality and sheath temperatures than that of outlet header break transients. Emergency core cooling system (ECCS) is found to be effective for the cases analysed. Initiating trip parameters and time for scram and ECCS injection is also investigated

  2. A study on cooling efficiency using 1-d analysis code suitable for cooling system of thermoforming

    International Nuclear Information System (INIS)

    Thermoforming is one of the most versatile and economical processes available for polymer products, but cycle time and production cost must be continuously reduced in order to improve the competitive power of products. In this study, water spray cooling was simulated to apply to a cooling system instead of compressed air cooling in order to shorten the cycle time and reduce the cost of compressed air used in the cooling process. At first, cooling time using compressed air was predicted in order to check the state of mass production. In the following step, the ratio of removed energy by air cooling or water spray cooling among the total removed energy was found by using 1-D analysis code of the cooling system under the condition of checking the possibility of conversion from 2-D to 1-D problem. The analysis results using water spray cooling show that cycle time can be reduced because of high cooling efficiency of water spray, and cost of production caused by using compressed air can be reduced by decreasing the amount of the used compressed air. The 1-D analysis code can be widely used in the design of a thermoforming cooling system, and parameters of the thermoforming process can be modified based on the recommended data suitable for a cooling system of thermoforming

  3. Verification of Gas System Analysis Code GAMMA+ with S-CO{sub 2} Compressor Test Data

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Seong Junl; Ahn, Yoonhan; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    In this study, the validation and verification of the GAMMA+ code, which is gas system transient analysis code was conducted by using the obtained SCO{sub 2}PE experimental results. Before performing a SCO{sub 2}PE loop transient simulation with the updated GAMMA+ code, major components, the compressor and the heat exchanger, were separately modeled. For the transient experiment, the reduction in cooling event was experimented in the SCO{sub 2}PE. The results of GAMMA+ code show reasonable agreement with SCO{sub 2}PE experimental data. However, there is a minute difference between the GAMMA+ prediction and the experimental data, especially at the compressor outlet condition because the heat transfer value from the experimental data was uncertain due to the measurement uncertainties and the CO{sub 2} properties near the critical point. To reduce the difference between the experimental data and GAMMA+ results, the modeling of SCO{sub 2}PE and the methodology for turbomachinery analyses will have to be updated in the future.

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

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    Wonkyeong Kim

    2015-01-01

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

  6. Development of a High Fidelity System Analysis Code for Generation IV Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hongbin Zhang; Vincent Mousseau; Haihua Zhao

    2008-06-01

    Traditional nuclear reactor system analysis codes such as RELAP and TRAC employ an operator split methodology. In this approach, each of the physics (fluid flow, heat conduction and neutron diffusion) is solved separately and the coupling terms are done explicitly. This approach limits accuracy (first order in time at best) and makes the codes slow in running since the explicit coupling imposes stability restrictions on the time step size. These codes have been extensively tested and validated for the existing LWRs. However, for GEN IV nuclear reactor designs which tend to have long lasting transients resulting from passive safety systems, the performance is questionable and modern high fidelity simulation tools will be required. The requirement for accurate predictability is the motivation for a large scale overhaul of all of the models and assumptions in transient nuclear reactor safety simulation software. At INL we have launched an effort with the long term goal of developing a high fidelity system analysis code that employs modern physical models, numerical methods, and computer science for transient safety analysis of GEN IV nuclear reactors. Modern parallel solution algorithms will be employed through utilizing the nonlinear solution software package PETSc developed by Argonne National Laboratory. The physical models to be developed will have physically realistic length scales and time scales. The solution algorithm will be based on the physics-based preconditioned Jacobian-free Newton-Krylov solution methods. In this approach all of the physical models are solved implicitly and simultaneously in a single nonlinear system. This includes the coolant flow, nonlinear heat conduction, neutron kinetics, and thermal radiation, etc. Including modern physical models and accurate space and time discretizations will allow the simulation capability to be second order accurate in space and in time. This paper presents the current status of the development efforts as

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

    International Nuclear Information System (INIS)

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

  8. Benchmark calculations by the nuclear criticality safety analysis code system JACS(MGCL, KENO-IV)

    International Nuclear Information System (INIS)

    Since 1980, as many as 1394 cases of benchmark calculations on criticality problems have been performed by the KENO-IV Monte Carlo calculation code with the MGCL cross section data library. The code system is a part of the criticality safety evaluation code system JACS developed at JAERI. The code validation results have been published in a series of JAERI-M reports and others. This report summarizes these results and the reliability of the code system systematically. The number of the calculated cases briefly described in this report together with their experimental systems and data are 502 for 17 kinds of homogeneous single-unit systems, 331 for 8 kinds of homogeneous multi-unit systems and 561 for 16 kinds of heterogeneous systems. Discussions and interpretations are made on the calculated keff's (neutron multiplication factors) with their bias errors. The factors related to the bias errors are confirmed together with their causes and trends. (author)

  9. 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.

  10. Fuel performance analysis code 'FAIR'

    International Nuclear Information System (INIS)

    For modelling nuclear reactor fuel rod behaviour of water cooled reactors under severe power maneuvering and high burnups, a mechanistic fuel performance analysis code FAIR has been developed. The code incorporates finite element based thermomechanical module, physically based fission gas release module and relevant models for modelling fuel related phenomena, such as, pellet cracking, densification and swelling, radial flux redistribution across the pellet due to the build up of plutonium near the pellet surface, pellet clad mechanical interaction/stress corrosion cracking (PCMI/SSC) failure of sheath etc. The code follows the established principles of fuel rod analysis programmes, such as coupling of thermal and mechanical solutions along with the fission gas release calculations, analysing different axial segments of fuel rod simultaneously, providing means for performing local analysis such as clad ridging analysis etc. The modular nature of the code offers flexibility in affecting modifications easily to the code for modelling MOX fuels and thorium based fuels. For performing analysis of fuel rods subjected to very long power histories within a reasonable amount of time, the code has been parallelised and is commissioned on the ANUPAM parallel processing system developed at Bhabha Atomic Research Centre (BARC). (author). 37 refs

  11. SWAT, Step-Wise Burnup Analysis Code System to Combine SRAC-95 Cell Calculation Code and ORIGEN2

    International Nuclear Information System (INIS)

    1 - Description of program or function: SWAT evaluates isotopic composition of spent nuclear fuel, especially for burnup credit issues by driving codes SRAC95 and ORIGEN2.1 or ORIGEN2. SWAT is an automated driver code system. At the initial development phase, it was constructed by combining source programs of SRAC and ORIGEN2. To overcome the problem associated with code updates, SWAT chose to use system function of UNIX operating system to execute SRAC95 and ORIGEN2. So that, SWAT is independent of development and modification of SRAC95 and ORIGEN2.1. In SWAT, ORIGEN2(82) or ORIGEN2.1 is used for burnup calculations using the matrix exponential method. An updated decay library is included in the distribution. SWAT uses SRAC95 for neutron spectrum and effective cross section calculation in 107 groups, using the collision probability method for given geometry and isotopic composition. One or two dimensional cell geometries are supported in SRAC95. NEA-1698/02: The main purpose of new package is to run SWAT on several machines not supported in previous package (IA64 under Linux, Windows with cygwin and Sun,...) and several commercial FORTRAN compiler (Intel, PGI, Fujitsu). 2 - Methods: In calculating the problem-dependent cross section in SWAT, the total burnup history is divided into 'burnup steps'. Power, boric acid concentration, temperature of each region, and void ratio of coolant are given as history data. For each burnup step, the neutron spectrum and effective cross section are evaluated by SRAC95 using the information given in previous burnup calculation and cell geometry information. The user can select geometry options for the collision probability method in SRAC95. 3 - Restrictions on the complexity of the problem: Resonance absorption calculation with ultra-fine group cross section can not be directly applicable for 2D geometry

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

    Science.gov (United States)

    Hartenstein, Richard G., Jr.

    1985-08-01

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

  13. Performance Analysis of Dual Unipolar/Bipolar Spectral Code in Optical CDMA Systems

    Directory of Open Access Journals (Sweden)

    C.T. Yen

    2013-03-01

    Full Text Available This study analyzes and calculates dual unipolar and bipolar coded configurations of spectral-amplitude-coding opticalcode division multiple access (SAC-OCDMA systems by using simulation methods. The important feature of theSAC-OCDMA systems is that multiple access interference (MAI can be eliminated by code sequences of a fixed inphasecross-correlation value. This property can be effectively canceled multiple access interference by using balancedetection schemes. This study uses Walsh-Hadamard codes as signature codes for the unipolar and bipolar schemes.The coder and decoder structures are based on optical filters of fiber Bragg gratings (FBGs. The simulation results ofunipolar/bipolar coding structures are first presented by commercial simulation obtained using OptiSystem software.The simulation results show that the bit error rate (BER through use of the bipolar coding method is superior to theunipolar scheme, especially when the received effect power is large. When the system needs good performance totransmit multimedia data, we can use bipolar scheme in the network. If the users only transmit voice data, the unipolarmethod can be employed. The eye diagram also shows that the bipolar encoding structure exhibits a wider openingthan the unipolar encoding structure. The flexible implementation of codewords assigns and integratable hardwaredesigns for the scheme with FBGs to realize dual coding OCDMA system is proposed.

  14. SPEAR-BETA fuel-performance code system. COSTF: cost implications analysis postprocessor. Final report

    International Nuclear Information System (INIS)

    The SPEAR-BETA fuel performance code system computes fuel reliability as a function of fuel precharacterization and operating history. This report describes an application of the SPEAR-BETA post-processor, COSTF, to the analysis of fuel failure cost implications for Oyster Creek Cycle 8. COSTF is a model which analyzes the cost implications of fuel failure and failure avoidance operations as a function of fuel precharacterization, operating history and cost parameters. The analysis reports baseline costs, representing those actually incurred during Cycle 8, and effects on these cost items of variations. Included are effects of varying 50 different cost parameters, as well as effects of varying power levels and ramp rates

  15. 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.

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

    Directory of Open Access Journals (Sweden)

    A. M. Beznarytnyy

    2013-09-01

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

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

    OpenAIRE

    Zummo SalamA

    2008-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-01-15

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

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

    International Nuclear Information System (INIS)

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

  20. RETRAN code analysis of Tsuruga-2 plant chemical volume control system (CVCS) reactor coolant leakage incident

    International Nuclear Information System (INIS)

    In the Chemical Volume Control System (CVCS) reactor primary coolant leakage incident, which occurred in Tsuruga-2 (4-loop PWR, 3,423 MWt, 1,160 MWe) on July 12, 1999, it took about 14 hours before the leakage isolation. The delayed leakage isolation and a large amount of leakage have become a social concern. Effective procedure modification was studied. Three betterments were proposed based on a qualitative analysis to reduce the pressure and temperature of the primary loop as fast as possible by the current plant facilities while maintaining enough subcooling of the primary loop. I analyzed the incident with RETRAN code in order to quantitatively evaluate the leakage reduction when these betterments are adopted. This paper is very new because it created a typical analysis method for PWR plant behavior during plant shutdown procedure which conventional RETRAN transient analyses rarely dealt with. Also the event time is very long. To carry out this analysis successfully, I devised new models such as an Residual Heat Removal System (RHR) model etc. and simplified parts of the conventional model. Based on the analysis results, I confirmed that leakage can be reduced by about 30% by adopting these betterments. Then the Japan Atomic Power Company (JAPC) modified the operational procedure for reactor primary coolant leakage events adopting these betterments. (author)

  1. Validation of CBZ code system for post-irradiation examination analysis and sensitivity analysis of (n,γ) branching ratio

    International Nuclear Information System (INIS)

    A code system CBZ is being developed in Hokkaido University. In order to validate it, PIE data, which are nuclide composition data of a spent fuel, have been analyzed with CBZ. The validity is evaluated as ratios of the calculation values to the experimental ones (C/E ratios). Differences between experimental values and calculation ones are smaller than 20% except some nuclides. Thus this code system is validated. Additionally, we evaluate influence of change of (n,γ) branching ratio on inventories of fission products and actinides. As a result, branching ratios of Sb-121, Pm-147, and Am-241 influence inventories of several nuclides. We perform PIE analysis using different (n,γ) branching ratio data from the ORIGEN-2 library, JNDC-Ver.2, and JEFF-3.1A, and find that differences in (n,γ) branching ratios between different nuclear libraries have a non-negligible influence on inventories of several nuclides. (author)

  2. ESCADRE and ICARE code systems

    International Nuclear Information System (INIS)

    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

  3. FORTRAN code-evaluation system

    Science.gov (United States)

    Capps, J. D.; Kleir, R.

    1977-01-01

    Automated code evaluation system can be used to detect coding errors and unsound coding practices in any ANSI FORTRAN IV source code before they can cause execution-time malfunctions. System concentrates on acceptable FORTRAN code features which are likely to produce undesirable results.

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

    Energy Technology Data Exchange (ETDEWEB)

    Vyskocil, Ladislav, E-mail: Ladislav.Vyskocil@ujv.cz; Macek, Jiri

    2014-11-15

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

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

    International Nuclear Information System (INIS)

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

  6. MARS 1.3 system analysis code coupling with CONTEMPT4/MOD5/PCCS containment analysis code using dynamic link library

    International Nuclear Information System (INIS)

    The two independent codes, MARS 1.3 and CONTEMPT4/MOD5/PCCS, have been coupled using the method of dynamic-link-library (DLL) technique. Overall configuration of the code system is designed so that MARS will be a main driver program which use CONTEMPT as associated routines. Using Digital Visual Fortran compiler, DLL was generated from the CONTEMPT source code with the interfacing routine names and arguments. Coupling of MARS with CONTEMPT was realized by calling the DLL routines at the appropriate step in the MARS code. Verification of coupling was carried out for LBLOCA transient of a typical plant design. It was found that the DLL technique is much more convenient than the UNIX process control techniques and effective for Window operating system. Since DLL can be used by more than one application and an application program can use many DLLs simultaneously, this technique would enable the existing codes to use more broadly with linking others

  7. Analysis of transients and accidents with the system code ATHLET for the Krsko Nuclear Power Plant

    International Nuclear Information System (INIS)

    Main aspects of the cooperation between the republic of Croatia and the F.R. of Germany in the field of NPP safety research are overviewed in the paper. The GRS system code ATHLET developed for the analysis of anticipated and abnormal plant transients, small and intermediate leaks as well as large breaks in light water reactors is now being available at the University of Zagreb. A very comprehensive ATHLET standard input data set for the NPP Krsko has been established. This data set was validated by calculation of the event 'Main Steam Isolation Valve Closure' that occurred at the PWR NPP Krsko in 1995 and comparing the resulting characteristic parameters with the corresponding measured data. (author)

  8. 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.

  9. ETR/ITER systems code

    International Nuclear Information System (INIS)

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

  10. Light-water reactor safety analysis codes

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-04-01

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

  12. Accumulative Landings System Code Tables

    Data.gov (United States)

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

  13. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Zummo SalamA

    2008-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Salam A. Zummo

    2009-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-06-01

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

  18. A ''SuperCode'' for performing systems analysis of tokamak experiments and reactors

    International Nuclear Information System (INIS)

    A new code, named the ''SUPERCODE,'' has been developed to fill the gap between currently available zero dimensional systems codes and highly sophisticated, multidimensional plasma performance codes. The former are comprehensive in content, fast to execute, but rather simple in terms of the accuracy of the physics and engineering models. The latter contain state-of-the-art plasma physics modelling but are limited in engineering content and time consuming to run. The SUPERCODE upgrades the reliability and accuracy of systems codes by calculating the self consistent 1 1/2 dimensional MHD-transport plasma evolution in a realistic engineering environment. By a combination of variational techniques and careful formation, there is only a modest increase in CPU time over O-D runs, thereby making the SUPERCODE suitable for use as a systems studies tool. In addition, considerable effort has been expended to make the code user- and programming-friendly, as well as operationally flexible, with the hope of encouraging wide usage throughout the fusion community

  19. 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

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

    Institute of Scientific and Technical Information of China (English)

    Qiao Xiaoqiang; Zhao Hangsheng; Cai Yueming

    2008-01-01

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

  1. ATHLET/BIPR-VVER - an advanced coupled code system for VVER safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Langenbuch, S.; Velkov, K. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Forschungsinstitute, D-85748 Garching (Germany); Lizorkin, M.; Nikonov, S. [RRC Kurchatov Institute, 123182 Moscow (Russian Federation)

    2008-07-01

    The coupled thermal-hydraulics and neutron-kinetics code ATHLET/BIPR-VVER was developed within a co-operation between the RRC Kurchatov Institute (KI) and GRS. The modelling capability of this coupled code as well as the status of validation by benchmark activities and comparisons with plant measurements are described. The paper is focused on the recent model developments, validation and coupled code application for the safety justification of VVER plants. A special region of interest which requires advanced modelling is the core outlet where the thermocouples are located. This is of high importance for the validation of the coupled system code ATHLET/BIPR-VVER on local parameters. With the new advanced modelling is analysed again the benchmark problem of Phase 1 of the CEA/NEA/OECD VVER-1000 Coolant Transient Benchmark. Based on data comparison with the experimental measurements the mixing phenomena at assembly head is estimated and mixing coefficients are introduced in the thermal-hydraulic core outlet models of the coupled system code ATHLET/BIPR-VVER. (authors)

  2. ATHLET/BIPR-VVER - an advanced coupled code system for VVER safety analysis

    International Nuclear Information System (INIS)

    The coupled thermal-hydraulics and neutron-kinetics code ATHLET/BIPR-VVER was developed within a co-operation between the RRC Kurchatov Institute (KI) and GRS. The modelling capability of this coupled code as well as the status of validation by benchmark activities and comparisons with plant measurements are described. The paper is focused on the recent model developments, validation and coupled code application for the safety justification of VVER plants. A special region of interest which requires advanced modelling is the core outlet where the thermocouples are located. This is of high importance for the validation of the coupled system code ATHLET/BIPR-VVER on local parameters. With the new advanced modelling is analysed again the benchmark problem of Phase 1 of the CEA/NEA/OECD VVER-1000 Coolant Transient Benchmark. Based on data comparison with the experimental measurements the mixing phenomena at assembly head is estimated and mixing coefficients are introduced in the thermal-hydraulic core outlet models of the coupled system code ATHLET/BIPR-VVER. (authors)

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

    International Nuclear Information System (INIS)

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

  4. Development of a personal computer code for fire protection analysis of DOE facility air-cleaning systems

    International Nuclear Information System (INIS)

    The United States Department of Energy (DOE) has sponsored development of a computer code to aid analysts performing fire hazards analyses for DOE facilities. The code selected for this application was the FIRAC code developed by the Los Alamos National Laboratory for the Nuclear Regulatory Commission. The original code has been modified by the Westinghouse-Hanford Company. The FIRAC code simulates fire accidents in nuclear facilities and predicts effects of a hypothetical fire within a compartment and its effect throughout the rest of the facility, particularly the air-cleaning systems. The FIRAC code was designed to run on Cray supercomputers. The input format is difficult to use. For this code to be useful to the DOE fire protection community, it had to be converted to run on an IBM PC and couple with a menu-driven pre-processor that would make preparation of the input easy to use for fire protection engineers. In addition, a graphical display of the analysis results was required. In this paper the authors describe the pre-processor, the PC version of FIRAC, and the post-processor graphics package. In the presentation, a demonstration of how to set up a problem and use the code is made. 4 figs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-03-01

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

  10. Post-test calculation and uncertainty analysis of the experiment Quench-07 with the system code Athlet-CD

    International Nuclear Information System (INIS)

    In the frame of developmental assessment and code validation, a post-test calculation of the test QUENCH-07 was performed with ATHLET-CD. The system code ATHLET-CD is being developed for best-estimate simulation of accidents with core degradation and for evaluation of accident management procedures. It applies the detailed models of the thermal-hydraulic code ATHLET in an efficient coupling with dedicated models for core degradation and fission products behaviour. The first step of the work was the simulation of the test QUENCH-07 applying the modelling options recommended in the code User's Manual (reference calculation). The global results of this calculation showed a good agreement with the measured data. This calculation was complemented by a sensitivity analysis in order to investigate the influence of a combined variation of code input parameters on the simulation of the main phenomena observed experimentally. Results of this sensitivity analysis indicate that the main experimental measurements lay within the uncertainty range of the corresponding calculated values. Among the main contributors to the uncertainty of code results are the heat transfer coefficient due to forced convection to superheated steam-argon mixture, the thermal conductivity of the shroud isolation and the external heater rod resistance. Uncertainties on modelling of B4C oxidation do not affect significantly the total calculated hydrogen release rates. (authors)

  11. Post-test calculation and uncertainty analysis of the experiment QUENCH-07 with the system code ATHLET-CD

    Energy Technology Data Exchange (ETDEWEB)

    Austregesilo, Henrique [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Forschungsinstitute, 85748 Garching (Germany)]. E-mail: Henrique.Austregesilo@grs.de; Bals, Christine [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Forschungsinstitute, 85748 Garching (Germany); Trambauer, Klaus [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Forschungsinstitute, 85748 Garching (Germany)

    2007-09-15

    In the frame of developmental assessment and code validation, a post-test calculation of the test QUENCH-07 was performed with ATHLET-CD. The system code ATHLET-CD is being developed for best-estimate simulation of accidents with core degradation and for evaluation of accident management procedures. It applies the detailed models of the thermal-hydraulic code ATHLET in an efficient coupling with dedicated models for core degradation and fission products behaviour. The first step of the work was the simulation of the test QUENCH-07 applying the modelling options recommended in the code User's Manual (reference calculation). The global results of this calculation showed a good agreement with the measured data. This calculation was complemented by a sensitivity analysis in order to investigate the influence of a combined variation of code input parameters on the simulation of the main phenomena observed experimentally. Results of this sensitivity analysis indicate that the main experimental measurements lay within the uncertainty range of the corresponding calculated values. Among the main contributors to the uncertainty of code results are the heat transfer coefficient due to forced convection to superheated steam-argon mixture, the thermal conductivity of the shroud isolation and the external heater rod resistance. Uncertainties on modelling of B{sub 4}C oxidation do not affect significantly the total calculated hydrogen release rates.

  12. Post-test calculation and uncertainty analysis of the experiment QUENCH-07 with the system code ATHLET-CD

    International Nuclear Information System (INIS)

    In the frame of developmental assessment and code validation, a post-test calculation of the test QUENCH-07 was performed with ATHLET-CD. The system code ATHLET-CD is being developed for best-estimate simulation of accidents with core degradation and for evaluation of accident management procedures. It applies the detailed models of the thermal-hydraulic code ATHLET in an efficient coupling with dedicated models for core degradation and fission products behaviour. The first step of the work was the simulation of the test QUENCH-07 applying the modelling options recommended in the code User's Manual (reference calculation). The global results of this calculation showed a good agreement with the measured data. This calculation was complemented by a sensitivity analysis in order to investigate the influence of a combined variation of code input parameters on the simulation of the main phenomena observed experimentally. Results of this sensitivity analysis indicate that the main experimental measurements lay within the uncertainty range of the corresponding calculated values. Among the main contributors to the uncertainty of code results are the heat transfer coefficient due to forced convection to superheated steam-argon mixture, the thermal conductivity of the shroud isolation and the external heater rod resistance. Uncertainties on modelling of B4C oxidation do not affect significantly the total calculated hydrogen release rates

  13. Tandem Mirror Reactor Systems Code (Version I)

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-01-15

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

  16. The Validation and Verification of Gas System Analysis Code GAMMA+ with S-CO2 Compressor Test Data

    International Nuclear Information System (INIS)

    In this study, the validation and verification of the GAMMA+ code was carried out by utilizing the experimental data from the SCO2PE. For a power conversion system analysis of a High Temperature Gas cooled Reactor (HTGR), the GAMMA code was developed. The code has been continuously updated to become GAMMA+. For this study, the GAMMA+ code was modified to connect with the NIST database. A novel compressor module was added to calculate the compressor map for the GAMMA+ code. Using the updated GAMMA+ code, the results were obtained for Table I inlet conditions and calculated results are shown in Table II and Figs. 6, 7 and 8 for two cases. As shown in the results, the GAMMA+ code has shown reasonable results in comparison with the SCO2PE experiment data except for the compressor outlet temperature. In this study, the process to calculate the compressor outlet temperature was added while considering the isentropic turbomachinery efficiency from the measured performance line. However, since the pressure ratio of the SCO2PE compressor is very low, the uncertainty of measurement is quite high near the critical point, even with the NIST database. Therefore, in calculating the isentropic compressor efficiency of SCO2PE, the electric power supplied for the compressor is utilized as a denominator for the compressor efficiency formula. The isentropic compressor efficiency formulas applied for the experiment and GAMMA+ code are respectively shown in Eq. (1) and (2). Therefore, in case of the compressor outlet temperature, there is quite a difference between the experiment and GAMMA+ data. η Experiment= m(hs-hin)/w (1). η GAMMA= m(hs-hin)/m(hout-hin) (2) This study is a preliminary study to utilize the GAMMA+ code for predicting the SCO2PE data. Thus only the steady state results are obtained so far. Further studies of transient data comparison will be performed in the near future

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

    Science.gov (United States)

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

    2012-04-01

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

  18. 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 codes is compared to the earlier reported two-dimensional (2-D)/3-D codes for asynchronous 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. PMID:20220892

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-03-01

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

  20. Parallel processing of structural integrity analysis codes

    International Nuclear Information System (INIS)

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

  1. The thermalhydraulic code ATHLET for analysis of PWR and BWR systems

    International Nuclear Information System (INIS)

    For the analyses of severe accidents with core degradation an extended version ATHLET-SA (Severe Accidents) is being developed. In this version the thermal-hydraulic models of ATHLET will be supplemented by core meltdown and fission product models. ATHLET/Mod 1.0 contains the four equation thermodynamic non-equilibrium model from the former DRUFAN code. The relative velocity between phases is determined by a new full-range drift flux model. This version also contains fast running models from the former ALMOD code including the GCSM module for the flexible description of control and balance-of-plant systems. As an example of recent assessment calculations a post-test calculation of LOBI Test A2-77A is presented. (orig.)

  2. The Tile-map Based Vulnerability Assessment Code of a Physical Protection System: SAPE (Systematic Analysis of Protection Effectiveness)

    International Nuclear Information System (INIS)

    Increasing threats on nuclear facilities demands stronger physical protection system (PPS) within the limited budget. For this reason we need an efficient physical protection system and before making an efficient PPS we need to evaluate it. This evaluation process should faithfully reflect real situation, reveal weak points and unnecessary protection elements, and give comparable quantitative values. Performance based analysis helps to build an efficient physical protection system. Instead of regulating the number of sensors and barriers, the performance based analysis evaluates a PPS fit to the situation of a facility. The analysis assesses delay (sensors) and detection (barriers) of a PPS against an intrusion, and judges whether a response force arrives before intruders complete their job. Performance based analysis needs complicated calculation and, hence, several assessment codes have been developed. A code called the estimation of adversary sequence interruption (EASI) was developed to analyze vulnerability along a single intrusion path. The systematic analysis of vulnerability to intrusion (SAVI) code investigates multi-paths to a valuable asset in an actual facility. SAVI uses adversary sequence diagram to describe multi-paths

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

  4. ENSDF ANALYSIS AND UTILITY CODES.

    Energy Technology Data Exchange (ETDEWEB)

    BURROWS, T.

    2005-04-04

    The ENSDF analysis and checking codes are briefly described, along with their uses with various types of ENSDF datasets. For more information on the programs see ''Read Me'' entries and other documentation associated with each code.

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

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

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

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

    Science.gov (United States)

    Jones, Scott M.

    2007-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  9. BWR stability and bifurcation analysis using reduced order models and system codes: Identification of a subcritical Hopf bifurcation using RAMONA

    International Nuclear Information System (INIS)

    The system code RAMONA, as well as a recently developed BWR reduced order model (ROM), are employed for the stability analysis of a specific operational point of the Leibstadt nuclear power plant. This has been done in order to assess the ROM's applicability and limitations in a quantitative manner. In the context of a detailed local bifurcation analysis carried out using RAMONA in the neighbourhood of the chosen Leibstadt operational point, a bridge is built between the ROM and the system code. This has been achieved through interpreting RAMONA solutions on the basis of the physical mechanisms identified in the course of applying the ROM. This leads, for the first time, to the identification of a subcritical Poincare-Andronov-Hopf (PAH) bifurcation using a system code. As a consequence, the possibility of the so-called correspondence hypothesis is suggested to underline the relationship between a stable (unstable) limit cycle solution and the occurrence of a supercritical (subcritical) PAH bifurcation in the modeling of boiling water reactor stability behaviour

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-12-01

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

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

    International Nuclear Information System (INIS)

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

  12. Coupling the RELAP5-3d advanced systems analysis code with commercial and advanced CFD software

    International Nuclear Information System (INIS)

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

  13. Post-test calculation and uncertainty analysis of the experiment QUENCH-07 with the system code ATHLET-CD

    International Nuclear Information System (INIS)

    Full text of publication follows: The QUENCH fuel bundle experiments, performed at the Forschungszentrum Karlsruhe in Germany, aim to investigate the hydrogen source term and the bundle degradation during reflood of an overheated reactor core. The test QUENCH-07, in which the bundle was cooled from high temperatures by steam injected from the bottom, was the first experiment in this test series with a boron carbide absorber rod in the bundle. One major objective of this test was to provide information on the B4C/SS/Zry interactions, on the formation of gaseous reaction products during B4C oxidation and control rod degradation, and on the impact of control rod degradation on surrounding rods. In the general frame of developmental assessment and code validation, a post-test calculation of test QUENCH-07 complemented by an uncertainty analysis was performed with the code ATHLET-CD. The system code ATHLET-CD is being developed for realistic simulation of accidents with core degradation and for evaluation of accident management procedures. It applies the detailed and validated models of the thermal-hydraulic code ATHLET in an efficient coupling with dedicated models for core degradation, including mechanical rod behaviour, zirconium and B4C oxidation, melting and relocation of metallic and ceramic components, and for the release and transport of fission products and aerosols. The first step of the work was the simulation of the QUENCH-07 experiment, applying the modeling options recommended in the code User's Manual (reference calculation). The global results of this calculation, mainly with respect to the hydrogen release rate and to the time evolution of bundle temperatures in different elevations, showed a good agreement with the measured data. This calculation was complemented by a sensitivity analysis in order to investigate the influence of a combined variation of code input parameters on the simulation of the main phenomena observed in the experiment. For this

  14. The Validation and Verification of Gas System Analysis Code GAMMA+ with S-CO{sub 2} Compressor Test Data

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Seong Jun; Ahn, Yoonhan; Lee, Jeong Ik [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-10-15

    In this study, the validation and verification of the GAMMA+ code was carried out by utilizing the experimental data from the SCO2PE. For a power conversion system analysis of a High Temperature Gas cooled Reactor (HTGR), the GAMMA code was developed. The code has been continuously updated to become GAMMA+. For this study, the GAMMA+ code was modified to connect with the NIST database. A novel compressor module was added to calculate the compressor map for the GAMMA+ code. Using the updated GAMMA+ code, the results were obtained for Table I inlet conditions and calculated results are shown in Table II and Figs. 6, 7 and 8 for two cases. As shown in the results, the GAMMA+ code has shown reasonable results in comparison with the SCO2PE experiment data except for the compressor outlet temperature. In this study, the process to calculate the compressor outlet temperature was added while considering the isentropic turbomachinery efficiency from the measured performance line. However, since the pressure ratio of the SCO2PE compressor is very low, the uncertainty of measurement is quite high near the critical point, even with the NIST database. Therefore, in calculating the isentropic compressor efficiency of SCO2PE, the electric power supplied for the compressor is utilized as a denominator for the compressor efficiency formula. The isentropic compressor efficiency formulas applied for the experiment and GAMMA+ code are respectively shown in Eq. (1) and (2). Therefore, in case of the compressor outlet temperature, there is quite a difference between the experiment and GAMMA+ data. η Experiment= m(h{sub s}-h{sub in})/w (1). η GAMMA= m(h{sub s}-h{sub in})/m(h{sub out}-h{sub in}) (2) This study is a preliminary study to utilize the GAMMA+ code for predicting the SCO2PE data. Thus only the steady state results are obtained so far. Further studies of transient data comparison will be performed in the near future.

  15. VVER-1000 main steam line break analysis using the coupled code system DYN3D/ATHLET

    International Nuclear Information System (INIS)

    Calculations using the coupled code system DYN3D/ATHLET were performed in the frame of the OECD/NEA MSLB benchmark for a VVER-1000 reactor. The coolant mixing inside the reactor pressure vessel was treated using a validated empirical mixing model implemented into the DYN3D/ATHLET code. Using very conservative boundary conditions (reduced scram worth, two stuck rods, running MCP throughout the whole transient) a return-to-power was predicted. For the assessment of the empirical mixing model a time dependent calculation using the computational fluid dynamics code CFX-10 was performed. For that analysis, a detailed model of the reactor pressure vessel consisting of the inlets nozzles, downcomer, lower plenum and a part of the core and having 4.67 million unstructured tetra cell elements was used. For the considered case with running main coolant pumps, this calculation shows a sector formation at the core inlet with a certain amount of mixing at the edges of the sector. A core calculation using these CFX results as boundary conditions predicted also a return-to-power with a maximum value being about 200 MW lower than in the coupled code calculation. This variation calculation confirms the applicability of the empirical mixing model. The comparison shows also, that in this way results with a reasonable degree of conservatism can be obtained. (authors)

  16. Beta Testing of CFD Code for the Analysis of Combustion Systems

    Science.gov (United States)

    Yee, Emma; Wey, Thomas

    2015-01-01

    A preliminary version of OpenNCC was tested to assess its accuracy in generating steady-state temperature fields for combustion systems at atmospheric conditions using three-dimensional tetrahedral meshes. Meshes were generated from a CAD model of a single-element lean-direct injection combustor, and the latest version of OpenNCC was used to calculate combustor temperature fields. OpenNCC was shown to be capable of generating sustainable reacting flames using a tetrahedral mesh, and the subsequent results were compared to experimental results. While nonreacting flow results closely matched experimental results, a significant discrepancy was present between the code's reacting flow results and experimental results. When wide air circulation regions with high velocities were present in the model, this appeared to create inaccurately high temperature fields. Conversely, low recirculation velocities caused low temperature profiles. These observations will aid in future modification of OpenNCC reacting flow input parameters to improve the accuracy of calculated temperature fields.

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

    CERN Document Server

    Corsetti, Fabiano

    2014-01-01

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

  18. Flow Analysis of Code Customizations

    DEFF Research Database (Denmark)

    Hessellund, Anders; Sestoft, Peter

    2008-01-01

    Inconsistency between metadata and code customizations is a major concern in modern, configurable enterprise systems. The increasing reliance on metadata, in the form of XML files, and code customizations, in the form of Java files, has led to a hybrid development platform. The expected consistency...... requirements between metadata and code should be checked but often are not, so current tools offer surprisingly poor development support. In this paper, we adapt classical data flow analyses to detect inconsistencies and provide better static guarantees. We provide a formalization of the consistency...... significant number of previously undetected consistency errors and have received very positive feedback from the developer community in the case study....

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

    Science.gov (United States)

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

    2001-06-01

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

  20. Stability analysis by ERATO code

    International Nuclear Information System (INIS)

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

  1. RETRAN code analysis of Tsuruga-2 plant chemical volume control system (CVCS) reactor coolant leakage incident

    International Nuclear Information System (INIS)

    JAPC purchased RETRAN, a program for transient thermal hydraulic analysis of complex fluid flow system, from the U.S. Electric Power Research Institute in 1992. Since then, JAPC has been utilizing RETRAN to evaluate safety margins of actual plant operation, in coping with troubles (investigating trouble causes and establishing countermeasures), and supporting reactor operation (reviewing operational procedures etc.). In this paper, a result of plant analysis performed on a CVCS reactor primary coolant leakage incident which occurred at JAPC's Tsuruga-2 plant (4-loop PWR, 3423 MWt, 1160 MW) on July 12 of 1999 and, based on the result, we made a plan to modify our operational procedure for reactor primary coolant leakage events in order to make earlier plant shutdown and this reduced primary coolant leakage. (author)

  2. HAMMER code system

    International Nuclear Information System (INIS)

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

  3. Performance Analysis for Bit Error Rate of DS- CDMA Sensor Network Systems with Source Coding

    Directory of Open Access Journals (Sweden)

    Haider M. AlSabbagh

    2012-03-01

    Full Text Available The minimum energy (ME coding combined with DS-CDMA wireless sensor network is analyzed in order to reduce energy consumed and multiple access interference (MAI with related to number of user(receiver. Also, the minimum energy coding which exploits redundant bits for saving power with utilizing RF link and On-Off-Keying modulation. The relations are presented and discussed for several levels of errors expected in the employed channel via amount of bit error rates and amount of the SNR for number of users (receivers.

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

    International Nuclear Information System (INIS)

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

  5. An Analysis of Syndrome Coding

    Science.gov (United States)

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

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

  6. Development of molecular dynamics simulation code linked to movie system and its application for analysis on dust particles in plasma

    International Nuclear Information System (INIS)

    In order to analyze structure and dynamic behavior of dust particles trapped in plasma-dc sheath boundary, we have developed object oriented 3-dimensional simulation code linked to movie system powered by JAVA. This report provides an overview of the developed simulation code, which will be opened as a public domain code. By using code, we have reproduced a Coulomb dust cloud observed experimentally with a funnel shape composed of multiple layers and simulate longitudinal waves' propagation to get their dispersion relation. (author)

  7. Analytical qualification of system identification (modal analysis) codes for use in the dynamic testing of nuclear power plant structures

    International Nuclear Information System (INIS)

    The analytical evaluation of two particular system identification codes used at Lawrence Livermore Laboratory is presented. Both codes are eigenparameter identification codes; however, one uses a time domain approach while the other a frequency domain approach. The evaluation was accomplished by analytically generating several time history signals in which the true modal parameters were known. These time histories ranged from widely spaced modes with spacing factors of 100 percent to closely spaced modes with spacing factors of 6 percent. These signals were then polluted with various levels of simulated measurement noise and the ability of our computer codes to extract the parameters from this noisy data was evaluated

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  10. Development of thermal-hydraulic system analysis code SSC-K for pool-type liquid metal reactor

    International Nuclear Information System (INIS)

    The Supper System Code of KAERI (SSC-K) is a best-estimate system code for analyzing an variety of off-normal or accident of a pool type design. It is developed at KAERI on the basis of SSC-L developed at BNL to analyze pool-type LMR transients. Because of inherent difference between th pool and loop design, the major modefications of SSC-L is required for the safety analysis of KALIMER. The major difference between KALIMER and general loop type LMRs exists in the primary heat transport system. In KALIMER, all of the essential components consisted of the primary heat transport system are located within the reactor vessel. This is contrast to the loop type LMRs, in which all the primary components are connected via piping to form loops attached externally to the reactor vessel. KALIMER has only one cover gas space. This eliminates the need for separate cover gas systems over liquid level in pump tanks and upper plenum. Since the sodium in hot pool is separated from cold pool by insulated barrier in KALIMER, The liquid level in hot pool is different from that in the cold pool mainly due to hydraulic losses and pump suction heads occuring during flow through the circulation pathes. In some accident conditions the liquid in the hot pool is flooded into cold pool and forms the natural circulation flow path. During the loss of heat sink transients, this will provided as a major heat rejection mechanism with the passive decay heat removal system. Since the pipes in the primary system exist only between pump discharge and core inlet plenum and are submerged in cold pool, a pipe rupture accident becomes less severe due to a constant back pressure exerted against the coolant flow from break. The intermediate and steam generator systems of both are generally identical. To adapt SSC-K to KALIMER design, the major modification of SSC-L has been made for the safety analysis of KALIMER. Test runs have been performed for the qualitative verification of the developed models. The

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-08-01

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

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

    International Nuclear Information System (INIS)

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

  14. System Design Description for the TMAD Code

    International Nuclear Information System (INIS)

    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

  15. Performance analysis of 2D asynchronous hard-limiting optical code-division multiple access system through atmospheric scattering channel

    Science.gov (United States)

    Zhao, Yaqin; Zhong, Xin; Wu, Di; Zhang, Ye; Ren, Guanghui; Wu, Zhilu

    2013-09-01

    Optical code-division multiple access (OCDMA) systems usually allocate orthogonal or quasi-orthogonal codes to the active users. When transmitting through atmospheric scattering channel, the coding pulses are broadened and the orthogonality of the codes is worsened. In truly asynchronous case, namely both the chips and the bits are asynchronous among each active user, the pulse broadening affects the system performance a lot. In this paper, we evaluate the performance of a 2D asynchronous hard-limiting wireless OCDMA system through atmospheric scattering channel. The probability density function of multiple access interference in truly asynchronous case is given. The bit error rate decreases as the ratio of the chip period to the root mean square delay spread increases and the channel limits the bit rate to different levels when the chip period varies.

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

    Directory of Open Access Journals (Sweden)

    Mitul Prajapati

    2012-03-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

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

  20. Fire-accident analysis code (FIRAC) verification

    International Nuclear Information System (INIS)

    The FIRAC computer code predicts fire-induced transients in nuclear fuel cycle facility ventilation systems. FIRAC calculates simultaneously the gas-dynamic, material transport, and heat transport transients that occur in any arbitrarily connected network system subjected to a fire. The network system may include ventilation components such as filters, dampers, ducts, and blowers. These components are connected to rooms and corridors to complete the network for moving air through the facility. An experimental ventilation system has been constructed to verify FIRAC and other accident analysis codes. The design emphasizes network system characteristics and includes multiple chambers, ducts, blowers, dampers, and filters. A larger industrial heater and a commercial dust feeder are used to inject thermal energy and aerosol mass. The facility is instrumented to measure volumetric flow rate, temperature, pressure, and aerosol concentration throughout the system. Aerosol release rates and mass accumulation on filters also are measured. We have performed a series of experiments in which a known rate of thermal energy is injected into the system. We then simulated this experiment with the FIRAC code. This paper compares and discusses the gas-dynamic and heat transport data obtained from the ventilation system experiments with those predicted by the FIRAC code. The numerically predicted data generally are within 10% of the experimental data

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

    Science.gov (United States)

    Poole, Eugene L.; Overman, Andrea L.

    1988-01-01

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

  2. MARAS - a computer code for semi-Markov reliability analysis of alternating systems

    International Nuclear Information System (INIS)

    It is now recognized that current testing and maintenance requirements invoke too many inadvertent reactor trips and that operating staff must devote significant amount of time and effort to comply with the requirements. With this recognition, the value and the impact of the proposed changes in the allowed outage time (AOT) and surveillance test interval(STI) are evaluated for the alternating system. Because of the testing and AOT requirements, the alternating system exhibits semi-Markovian characteristics which change states in accordance with a Markov chain but take a nonexponentially distributed amount of time between changes. It is observed from the results that there is an optimal point that gives lowest core damage probability and that the optimal point depends on input parameters. With these results, we can conclude that the methodology developed in this study can be applied to the existing alternating systems to evaluate accurately the various alternatives in the technical specifications

  3. Uncertainty analysis of the SWEPP drum assay system for graphite content Code 300

    International Nuclear Information System (INIS)

    The Idaho National Engineering Laboratory is being used as a temporary storage facility for transuranic waste generated by the U.S. Nuclear Weapons program at the Rocky Flats Plant (RFP) in Golden, Colorado. Currently, there is a large effort in progress to prepare to ship this waste to the Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico. In order to meet the TRU Waste Characterization Quality Assurance Program Plan nondestructive assay compliance requirements and quality assurance objectives, it is necessary to determine the total uncertainty of the radioassay results produced by the Stored Waste Examination Pilot Plant (SWEPP) Passive Active Neutron (PAN) radioassay system. This paper discusses a modified statistical sampling and verification approach used to determine the total uncertainty of a SWEPP/PAN measurement. In this modified approach the total performance of the SWEPP/PAN nondestructive assay system is simulated using computer models of the assay system and the resultant output is compared with the known input to assess the total uncertainty

  4. Sensitivity analysis of U238 cross sections in fast nuclear systems-SENSEAV-R computer code

    International Nuclear Information System (INIS)

    For many performance parameters of reactors the tabulated ratio calculation/experiment indicate that some potential problems may exist either in the cross section data or in the calculation models used to investigate the critical experimental data. A first step toward drawing a more definite conclusion is to perform a selective analysis of sensitivity profiles and covariance data files for the cross section data used in the calculation. Many works in the current literature show that some of these uncertainties come from uncertainties in 238U(n,γ), 238U(n,f) 239Pu(n,f). Perturbation methods were developed to analyze the effects of finite changes in a large number of cross sections and summarize the investigation by a group dependent sensitivity coefficient. As an application, the results of this investigation indicates that improvements should be done only on the medium and low energy ranges of 238U(n,γ) based on an analysis of cost and economic benefits. (Author)

  5. Uncertainty analysis of the SWEPP drum assay system for graphite content code 300

    International Nuclear Information System (INIS)

    The Idaho National Engineering Laboratory is being used as a temporary storage facility for transuranic waste generated by the U.S. Nuclear Weapons program at the Rocky Flats Plant (RFP) in Golden, Colorado. Currently, there is a large effort in progress to prepare to ship this waste to the Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico. In order to meet the TRU Waste Characterization Quality Assurance Program Plan nondestructive assay compliance requirements and quality assurance objectives, it is necessary to determine the total uncertainty of the radioassay results produced by the Stored Waste Examination Pilot Plant (SWEPP) Passive Active Neutron (PAN) radioassay system. This paper discusses a modified statistical sampling and verification approach used to determine the total uncertainty of a SWEPP/PAN measurement. In this modified approach the total performance of the SWEPP/PAN nondestructive assay system is simulated using computer models of the assay system and the resultant output is compared with the known input to assess the total uncertainty. The first waste form to be tested using this approach is weapons grade plutonium-contaminated graphite molds contained in 207 liter drums. The validity of the simulation approach is verified by comparing simulated output against results from calibration measurements using known plutonium sources and a graphite waste form calibration drum. For actual graphite waste form conditions, a set of 50 cases covering a statistical sampling of the conditions exhibited in graphite wastes was compiled using a Latin hypercube statistical sampling approach. The distributions from which Latin hypercube sample was drawn was derived from reviews of approximately 100 real-time radiography video tapes of RFP graphite waste drums, results from previous SWEPP/PAN measurements on graphite waste drums, and shipping data from RFP where the graphite waste was generated

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-12-01

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

  7. Uncertainty analysis for control rod ejection accidents simulated by KIKO3D/TRABCO code system

    International Nuclear Information System (INIS)

    Recently, considerable conservatism must be applied in the traditional safety analyses for taking into account the uncertainties originating from the input parameters, approximations in the models, due to the safety reserves, etc. The extreme values for all of the input parameters are supposed in the traditional safety analysis at the same time. Additionally it must be mentioned that the selection of the input parameter values leading to conservative results often is not easy. The main goal of this paper is to present a more realistic methodology for the case of control rod ejection accidents. The applied consistent statistical approach leads to conservative results also, but avoids the unnecessary cumulative conservatism. A method based on a mathematical model ('Two-Sided Statistical Tolerance Intervals', [1-2]) was chosen for the realization of uncertainty analyses of Reactivity Initiated Accidents (RIA). (author)

  8. Understanding Code Patterns - Analysis, Interpretation & Measurement

    CERN Document Server

    Dundas, Jitesh

    2011-01-01

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

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

    Institute of Scientific and Technical Information of China (English)

    王晋; 张东辉; 胡文军

    2016-01-01

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

  10. 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.

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

    International Nuclear Information System (INIS)

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

  12. A benchmark-problem specification and calculation using SENSIBL, a one- and two-dimensional sensitivity and uncertainty analysis code of the AARE system

    International Nuclear Information System (INIS)

    The lack of suitable benchmark problems makes it difficult to test sensitivity codes with a covariance library. A benchmark problem has therefore been defined for one- and two-dimensional sensitivity and uncertainity analysis codes and code systems. The problem, representative of a fusion reactor blanket, has a simple, three-zone /tau/-z geometry containing a D-T fusion neutron source distributed in a central void region surrounded by a thick 6LiH annulus. The response of interest is the 6Li tritium production per source neutron, T6. The calculation has been performed with SENSIBL using other codes from the AARE code system as a test of both SENSIBL and the linked, modular system. The caluclation was performed using the code system in the standard manner with a covariance data library in the COVFILS-2 format but modified to contain specifically tailored covariance data for H and 6Li (Path A). The calculation was also performed by a second method which uses specially perturbed H and Li cross sections (Path B). This method bypasses SENSIBL and allows a hand calculation of the benchmark T6 uncertainties. The results of Path A and Path B were total uncertainties in T6 of 0.21% and 0.19%, respectively. The closeness of the results for this challenging test gives confidence that SENSIBL and the AARE system will perform well for realistic sensitivity and uncertainty analyses

  13. Sandia National Laboratories analysis code data base

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, C.W.

    1994-11-01

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

  14. Web interface for plasma analysis codes

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-04-15

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

  15. Development of the methodology of the safety analysis performed by the coupled Kiko-3D/Athlet code system in VVER-440 type NPP

    International Nuclear Information System (INIS)

    In the deterministic safety analysis, codes are required in order to provide evaluations of potential nuclear power plant accidents. In the fields of the core transient behaviour, the computer codes have achieved a high degree of realistic modelling. Nevertheless, some further tools for the investigations of the wide range of physical phenomena in the whole plant transient, such as modeling the ex-core detector signals and the malfunctioning of the emergency control system are unavoidable, too. The main objective of this work is to show the status of the code package based on the best estimate coupled code system Athlet/Kiko-3D and to present a safety analysis. The programs and methods used in KFKI-AEKI for safety analysis of VVER-440 NPP are presented. The accident analysis methodology for a boron dilution scenario, in which an inactive coolant loop is started, is shown. The cooling and the strong dilution increase the reactivity resulting in increasing power level especially in the affected sector. Due to the use of time dependent signals of the ex-core detectors the SCRAM (emergency shutdown) is delayed. Investigating the DNBR (burnout ratio) value by the TRABCO code, no dangerous hot spot was found

  16. Understanding Code Patterns - Analysis, Interpretation & Measurement

    OpenAIRE

    Dundas, Jitesh

    2011-01-01

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

  17. Fire-accident analysis code (FIRAC) verification

    International Nuclear Information System (INIS)

    The FIRAC computer code predicts fire-induced transients in nuclear fuel cycle facility ventilation systems. FIRAC calculates simultaneously the gas-dynamic, material transport, and heat transport transients that occur in any arbitrarily connected network system subjected to a fire. The network system may include ventilation components such as filters, dampers, ducts, and blowers. These components are connected to rooms and corridors to complete the network for moving air through the facility. An experimental ventilation system has been constructed to verify FIRAC and other accident analysis codes. The design emphasizes network system characteristics and includes multiple chambers, ducts, blowers, dampers, and filters. A large industrial heater and a commercial dust feeder are used to inject thermal energy and aerosol mass. The facility is instrumented to measure volumetric flow rate, temperature, pressure, and aerosol concentration throughout the system. Aerosol release rates and mass accumulation on filters also are measured. This paper compares and discusses the gas-dynamic and heat transport data obtained from the ventilation system experiments with those predicted by the FIRAC code. The numerically predicted data generally are within 10% of the experimental data

  18. Analysis of thermohydraulic limits during WWER-1000 nuclear power plants heat-up using RELAP5 system code

    International Nuclear Information System (INIS)

    Plant heat-up is a process which all operating systems such as primary coolant circuit, pressurizer, primary and secondary sides of the steam generators and etc. are transferred from a cold shutdown to a hot standby status. During plant heat-up, some thermohydraulic limitations such as maximum and minimum allowable pressure and maximum rate of increase in pressure and temperature which are recommended by plant commissioning program and NPP safety related documents should be considered. Maximum allowable pressure prevents brittle fracture in reactor vessel, Minimum allowable pressure in the inlet of the reactor coolant pumps (RCPs) prevent pump cavitations and maximum allowable rate of increase in temperature and pressure respectively prevent thermal and mechanical shocks. Thus, tuning pressure and temperature increasing rates during plant heat-up is important from plant safety point of view. The RELAP5 system code was used to model and analysis the behavior of WWER-1000 plants during heat-up. In plant heat-up, at first the primary circuit pressure increases by injection of N2 gas into pressurizer in order to provide minimum required NPSH (net positive suction head) for operation of the RCPs. After short time RCPs are turned on to operate which increases the primary coolant circuit temperature through friction losses. At a time which is specified by heat-up procedure the pressurizer heaters are turning on to increase the primary circuit pressure. Heat transfer from primary to secondary side in the steam generators causes increasing of the secondary side temperature and pressure. Temperature and pressure of primary and secondary circuits increase until plant reaches to hot standby condition. The results show that the thermohydraulic parameters during plant heat-up are in an acceptable range and have a good agreement with available data in technical documents. (authors)

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

    Directory of Open Access Journals (Sweden)

    S. Sivanesskumar

    2011-01-01

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

  20. User's manual of the MKENO-DAR code system

    International Nuclear Information System (INIS)

    The computer code manual of MKENO-DAR which is a direct angular representation Monte Carlo code for criticality safety analysis is already issued as JAERI-M report, however, complex pre-stage data handlings and calculations by auxiliary programs are required before the execution of MKENO-DAR. The MKENO-DAR CODE SYSTEM widely spans a whole code system including MKENO-DAR and other pre-stage auxiliary programs. This report discusses the systematic treatment of the MKENO-DAR CODE SYSTEM and shows the simplified calculation technique from the user side of view. (author)

  1. The EGS5 code system

    International Nuclear Information System (INIS)

    The Electron-Gamma Shower (EGS) code system is a general purpose package for the Monte Carlo simulation of the coupled transport of electrons and photons in an arbitrary geometry for particles with energies above a few keV up to several hundred GeV (depending on the atomic numbers of the target materials). This report introduces a new, enhanced version called EGS5. In addition to explaining and documenting the various enhancements and changes to the previous version (EGS4), this document includes several introductory and advanced tutorials on the use of EGS5, and also contains the EGS5 User Manual. Our intention has been to make this document wholly self-contained so that the user need not refer to the original EGS4 manual (SLAC-265) in order to use the code. To this end, we have taken the liberty of incorporating into Chapter 2 of this report those portions of Chapter 2 of SLAC-265 which describe physics models of EGS4 retained by EGS5, thereby documenting all the physics contained in EGS5. (author)

  2. Validation of the KARATE-440 code system by the analysis of recriticality measurement of the Novovoronezh NPP

    International Nuclear Information System (INIS)

    In this paper the results of KARATE-440 calculations on Novovoronezh NPP Recriticality Experiment are presented, the corresponding parameters are analyzed. The simulation of the processes and the comparison of the results with the measurements are of particular interest as these efforts make our code to be validated in a higher level. Even if only some well defined states of the transient were simulated, satisfactory agreement was found between measured and calculated data. The results present evidence that the KARATE-440 code package can adequately model the reactor states in a wide range of performance parameters and the special core type referred in the experiment so it is acceptable for neutronic analysis of all the VVER-440 NPPs. (Authors)

  3. Expansion of the CHR bone code system

    International Nuclear Information System (INIS)

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

  4. Analysis of the Pressurized Water Reactor Main Steam Line Break Benchmark by the Coupled Code System ATHLET-QUABOX/CUBBOX

    International Nuclear Information System (INIS)

    The Organization for Economic Cooperation and Development (OECD) Pressurized Water Reactor Main Steam Line Break (MSLB) Benchmark has been calculated for all three exercises by the coupled code system ATHLET-QUABOX/CUBBOX developed by Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS). The results obtained are presented, and a detailed comparison with other solutions of the benchmark is discussed. An attempt is made to explain the differences observed in the solutions by the different modeling of physical processes in the codes. The sensitivity of results on modeling features is also investigated. In addition, the effect of different mapping schemes between fuel assemblies of the core loading and the thermal-fluid dynamics on the accuracy of three-dimensional (3-D) neutronics solutions is studied. The results for the MSLB transient are also evaluated to compare 3-D neutronics and point-kinetics solutions in view of integral and local parameters. Thus, the experiences with the coupled code system ATHLET-QUABOX/CUBBOX during the MSLB benchmark activity are summarized

  5. Integrated burnup calculation code system SWAT

    International Nuclear Information System (INIS)

    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)

  6. Manometer Behavior Analysis using CATHENA, RELAP and GOTHIC Codes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yang Hoon; Han, Kee Soo; Moon, Bok Ja; Jang, Misuk [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

    2014-05-15

    In this presentation, simple thermal hydraulic behavior is analyzed using three codes to show the possibility of using alternative codes. We established three models of simple u-tube manometer using three different codes. CATHENA (Canadian Algorithm for Thermal hydraulic Network Analysis), RELAP (Reactor Excursion and Leak Analysis Program), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are used for this analysis. CATHENA and RELAP are widely used codes for the analysis of system behavior of CANDU and PWR. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. In this paper, the internal behavior of u-tube manometer was analyzed using 3 codes, CATHENA, RELAP and GOTHIC. The general transient behavior is similar among 3 codes. However, the behavior simulated using GOTHIC shows some different trend compared with the results from the other 2 codes at the end of the transient. It would be resulted from the use of different physical model in GOTHIC, which is specialized for the multi-phase thermal hydraulic behavior analysis of containment system unlike the other two codes.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-03-01

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

  9. The CZTU uranium concentration analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Clark, D., LLNL

    1998-07-17

    A {sup 235}U analysis code, CZTU, has been written that can non- destructively evaluate the percentage of {sup 235}U in a uranium sample from the analysis of the emitted gamma rays. This code utilizes gamma spectra measured from room temperature Cadmium Zinc Telluride (CdZnTe or CZT) detectors. It has an accuracy midway between that obtained with sodium iodide and germanium crystal detectors. This report describes how to use the code, some results, limitations and design considerations.

  10. elestres: nuclear fuel analysis code

    International Nuclear Information System (INIS)

    The computer code ELESTRES models the thermal and mechanical behaviour of an individual fuel element, during its irradiation life under normal operating conditions. The finite element code ELESTRES models the two-dimensional axisymmetric behaviour of a CANDU fuel element during normal operation.The main focus of the code is to estimate temperatures, fission gas release and axial variations of deformation and stresses in the pellet and in the sheath. Thus the code is able to predict details like stresses/strains at circumferential. This paper describes the current version of ELESTRES. The emphasis is on a recent addition: multiaxial stresses in the sheath near circumferential ridges. For accuracy in the critical region, a fine mesh used near the ridge. To keep computing costs low, a coarse mesh is used near the midplane of the pellet

  11. Analysis and Compensation of Transmitter and Receiver I/Q Imbalances in Space-Time Coded Multiantenna OFDM Systems

    Directory of Open Access Journals (Sweden)

    Yaning Zou

    2007-12-01

    Full Text Available The combination of orthogonal frequency division multiplexing (OFDM and multiple-input multiple-output (MIMO techniques has been widely considered as the most promising approach for building future wireless transmission systems. The use of multiple antennas poses then big restrictions on the size and cost of individual radio transmitters and receivers, to keep the overall transceiver implementation feasible. This results in various imperfections in the analog radio front ends. One good example is the so-called I/Q imbalance problem related to the amplitude and phase matching of the transceiver I and Q chains. This paper studies the performance of space-time coded (STC multiantenna OFDM systems under I/Q imbalance, covering both the transmitter and the receiver sides of the link. The challenging case of frequency-selective I/Q imbalances is assumed, being an essential ingredient in future wideband wireless systems. As a practical example, the Alamouti space-time coded OFDM system with two transmit and M receive antennas is examined in detail and a closed-form solution for the resulting signal-to-interference ratio (SIR at the detector input due to I/Q imbalance is derived. This offers a valuable analytical tool for assessing the I/Q imbalance effects in any STC-OFDM system, without lengthy data or system simulations. In addition, the impact of I/Q imbalances on the channel estimation in the STC-OFDM context is also analyzed analytically. Furthermore, based on the derived signal models, a practical pilot-based I/Q imbalance compensation scheme is also proposed, being able to jointly mitigate the effects of frequency-selective I/Q imbalances as well as channel estimation errors. The performance of the compensator is analyzed using extensive computer simulations, and it is shown to virtually reach the perfectly matched reference system performance with low pilot overhead.

  12. Analysis and Compensation of Transmitter and Receiver I/Q Imbalances in Space-Time Coded Multiantenna OFDM Systems

    Directory of Open Access Journals (Sweden)

    Zou Yaning

    2008-01-01

    Full Text Available Abstract The combination of orthogonal frequency division multiplexing (OFDM and multiple-input multiple-output (MIMO techniques has been widely considered as the most promising approach for building future wireless transmission systems. The use of multiple antennas poses then big restrictions on the size and cost of individual radio transmitters and receivers, to keep the overall transceiver implementation feasible. This results in various imperfections in the analog radio front ends. One good example is the so-called I/Q imbalance problem related to the amplitude and phase matching of the transceiver I and Q chains. This paper studies the performance of space-time coded (STC multiantenna OFDM systems under I/Q imbalance, covering both the transmitter and the receiver sides of the link. The challenging case of frequency-selective I/Q imbalances is assumed, being an essential ingredient in future wideband wireless systems. As a practical example, the Alamouti space-time coded OFDM system with two transmit and M receive antennas is examined in detail and a closed-form solution for the resulting signal-to-interference ratio (SIR at the detector input due to I/Q imbalance is derived. This offers a valuable analytical tool for assessing the I/Q imbalance effects in any STC-OFDM system, without lengthy data or system simulations. In addition, the impact of I/Q imbalances on the channel estimation in the STC-OFDM context is also analyzed analytically. Furthermore, based on the derived signal models, a practical pilot-based I/Q imbalance compensation scheme is also proposed, being able to jointly mitigate the effects of frequency-selective I/Q imbalances as well as channel estimation errors. The performance of the compensator is analyzed using extensive computer simulations, and it is shown to virtually reach the perfectly matched reference system performance with low pilot overhead.

  13. Comparison of Interfacial and Wall Friction Models in Thermal-Hydraulic System Analysis Codes (Rev1.0)

    International Nuclear Information System (INIS)

    This reports is a literature survey on models and correlations for interfacial and wall friction models that are used to simulate thermal-hydraulics in nuclear reactors. The interfacial and wall frictions are needed to solve the momentum equations of gas, continuous liquid and droplet. Not only existing system codes, such as RELAP5-3D, TRAC-M, MARS, TRACE, CATHARE) but also up-to-date researches were reviewed. This report is a revised version of the previous technical report(KAERI/TR-3437/2007)

  14. FORTRAN Automated Code Evaluation System (FACES) user's manual, version 2

    Science.gov (United States)

    1975-01-01

    A system which provides analysis services for FORTRAN based software systems not normally available from system software is presented. The system is not a compiler, and compiler syntax diagnostics are not duplicated. For maximum adaptation to FORTRAN dialects, the code presented to the system is assumed to be compiler acceptable. The system concentrates on acceptable FORTRAN code features which are likely to produce undesirable results and identifies potential trouble areas before they become execution time malfunctions.

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

    International Nuclear Information System (INIS)

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

  16. EURDYN-1D: a computer code for the one-dimensional non-linear dynamic analysis of structural systems. Description and users' manual (release 1)

    International Nuclear Information System (INIS)

    The goal of the present report is to provide for a comprehensive users' manual describing the capabilities of the computer code EURDYN-1D. It includes information and examples about the type of problems which can be solved with the code and explanation on how to prepare input data and, how to interpret output results. The field of applications of EURDYN-1D is the one dimensional dynamic analysis of general structural systems and the code is particularly suited for fast transient events involving propagation of longitudinal mechanical waves (subsonic) in structures. Both geometrical and physical non-linearities can be taken into account. Typical examples are impact problems, fast dynamic loading due the explosions or sudden release for initial loads due to failures, etc. To these classes belong many problems encountered in the reactor safety field as well as in more common and general technological applications

  17. Research on application of system of neutron, thermohydraulic and safety analysis codes in order to simulation of the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Requirements of neutron, thermohydraulic and safety analysis calculation are very important because of issuing new version of SAR for DNRR, research on construction of new research reactor and nuclear power plant. Research on application of system of neutron, thermohydraulic and safety analysis codes in order to simulation of the Dalat Nuclear Research Reactor has been done in the frame work of research theme in the year 2002-2003. The purposes of the research are maintaining safety operation of the DNRR and enhancement of man power and calculation and safety analysis tool potential. (author)

  18. The GRS thermalhydraulic system code ATHLET

    International Nuclear Information System (INIS)

    The thermalhydraulic system code ATHLET is being developed by the Gesellschaft fuer Reaktorsicherheit (GRS) for the safety analysis of LWRs. The field of application comprises the whole spectrum of operational and abnormal transients, small and intermediate leaks as well as large breaks. An extended version (ATHLET-SA) that is being developed in cooperation with the Institut fuer Kernenergetik und Energiesysteme (IKE) at Stuttgart will allow the simulation of severe accidents with core degradation. A remarkable feature of ATHLET is the optional use of different fluiddynamic models. Depending on the type of problem to be analyzed fast running or more detailed models can be selected from essentially the same input deck. Emphasis is given to a methodical assessment procedure which provides the basis for the quantification of code uncertainties. (author). 12 refs

  19. SRAC2006: A comprehensive neutronics calculation code system

    International Nuclear Information System (INIS)

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

  20. Development and applicability analysis of ATHLET-SC code

    International Nuclear Information System (INIS)

    Research activities of supercritical water reactor (SCWR) have been carried out worldwide,aiming at cost reduction by system simplification and higher thermal efficiency. One of the most important tasks for the design and assessment of SCWR performance is to develop system analysis codes which are applicable under supercritical conditions. The paper presents the development of new system analysis code ATHLET-SC based on ATHLET 2.1A. Thermo-physical properties package valid for supercritical conditions are implemented into the existing ATHLET code to extend its application to safety analysis of SCWR. In to evaluate the applicability of the modified code, a core calculation model of mixed SCWR(SCWR-M) was proposed and analyzed, and transients of core power were simulated. Moreover, a simplified supercritical water cooled loop was proposed and its stability behaviors were analyzed. The results achieved up to now indicate a good applicability of the modified ATHLET code (ATHLET-SC) in simulation of SCWR. (authors)

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

    Science.gov (United States)

    Salim, Mohd Faiz; Roslan, Ridha; Ibrahim, Mohd Rizal Mamat @

    2014-02-01

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

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

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    Ahmed Nabih Zaki Rashed

    2013-04-01

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

  4. Development of criticality accident analysis code AGNES

    International Nuclear Information System (INIS)

    A one-point kinetics code, AGNES2, has been developed for the evaluation of the criticality accident of nuclear solution fuel system. The code has been evaluated through the simulation of TRACY experiments and used for the study of the condition of the JCO criticality accident. A code, AGNES-P, for the criticality accident of nuclear powder system has been developed based on AGNES2. It is expected that these codes be useful for the evaluation of criticality safety for fuel reprocessing and fabrication plants. (author)

  5. Current lead thermal analysis code 'CURRENT'

    International Nuclear Information System (INIS)

    Large gas-cooled current lead with the capacity more than 30 kA and 22 kV is required for superconducting toroidal and poloidal coils for fusion application. The current lead is used to carry electrical current from the power supply system at room temperature to the superconducting coil at 4 K. Accordingly, the thermal performance of the current lead is significantly important to determine the heat load requirements of the coil system at 4 K. Japan Atomic Energy Research Institute (JAERI) has being developed the large gas-cooled current leads with the optimum condition in which the heat load is around 1 W per 1 kA at 4 K. In order to design the current lead with the optimum thermal performances, JAERI developed thermal analysis code named as ''CURRENT'' which can theoretically calculate the optimum geometric shape and cooling conditions of the current lead. The basic equations and the instruction manual of the analysis code are described in this report. (author)

  6. Thermal-hydraulics analysis of thermal mock-up test for JCO precipitation vessel using α-FLOW code system

    International Nuclear Information System (INIS)

    In the criticality accident occurred in the JCO Tokai plant, a significant power level in the plateau part covering about 20-hour duration continued following the initial-burst power. In order to reproduce the thermal characteristics of the plateau part, a mock-up device was made. A series of power-tracking tests had been performed to estimate the amount of water evaporation and the solution temperature change by changing the power of electric heaters. Based on the experimental data of the power-tracking test on the JCO precipitation vessel, a series of thermal-hydraulics analyses in the plateau part was performed using α-FLOW code system. Assuming a heat transfer coefficient at outer surface of the vessel, two-dimensional calculations were performed, and the calculated solution temperatures and the outlet temperature of the cooling-water were reproduced the measured value within 3-5degC. (author)

  7. Application of the FAST code system to the static analysis of the low-void core of Gen-IV sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    The neutronic, thermal-hydraulic, and thermal-mechanic models of the low void core (CFV), a core design similar to the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) core design, represent the static core at End-of-Cycle and at nominal conditions, using the core specifications provide by the ASTRID core designers. The models are implemented in the FAST code system, a code package that establishes the coupling between 3D core neutronic, thermal-hydraulic and thermal-mechanic simulations for steady state and transient analysis. The static neutronic analysis, performed by means of the SERPENT 2 Monte-Carlo code, provides the core excess reactivity, power distributions, kinetic parameters, reactivity coefficients, and control rod reactivity worth as the main outcomes. In addition, an extensive study is carried out concerning the coolant void worth by analyzing specific sodium voiding scenarios. It can be confirmed that the ASTRID peculiar core design features a negative global coolant worth. Within the static thermal-hydraulic study, carried out by employing the TRACE code, the core assemblies division into cooling groups is performed. The respective coolant flow rates and temperatures at the core outlet are obtained. Finally, the static thermal-mechanic analysis, which was accomplished by means of the FRED code, gives general insights about the fuel temperature distribution within the fuel pins, gas gap conductance, and fission gas release. As the major outcome, the necessary static parameters of the core to proceed to the planned analyses of core transients are obtained. This document is made up of an abstract and the slides of the presentation. (author)

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

    International Nuclear Information System (INIS)

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

  9. Integral effect test and code analysis on the cooling performance of the PAFS (passive auxiliary feedwater system) during an FLB (feedwater line break) accident

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Byoung-Uhn, E-mail: bubae@kaeri.re.kr; Kim, Seok; Park, Yu-Sun; Kang, Kyoung-Ho

    2014-08-15

    Highlights: • This study focuses on the experimental validation of the operational performance of the PAFS (passive auxiliary feedwater system). • A transient simulation of the FLB (feedwater line break) in the integral effect test facility, ATLAS-PAFS, was performed to investigate thermal hydraulic behavior during the PAFS actuation. • The test result confirmed that the APR+ has the capability of coping with the FLB scenario by adopting the PAFS and proper set-points for its operation. • The experimental result was utilized to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. - Abstract: APR+ (Advanced Power Reactor Plus), which is a GEN-III+ nuclear power plant developed in Korea, adopts PAFS (passive auxiliary feedwater system) as an advanced safety feature. The PAFS can completely replace an active auxiliary feedwater system by cooling down the secondary side of steam generators with a natural convection mechanism. This study focuses on experimental and analytical investigation for cooling and operational performance of the PAFS during an FLB (feedwater line break) transient with an integral effect test facility, ATLAS-PAFS. To realistically simulate the FLB accident of the APR+, the three-level scaling methodology was taken into account to design the test facility and determine the test condition. From the test result, the PAFS was actuated to successfully cool down the decay heat of the reactor core by the condensation heat transfer at the PCHX (passive condensation heat exchanger), and thus it could be confirmed that the APR+ has the capability of coping with a FLB scenario by adopting the PAFS and proper set-points for its operation. This integral effect test data were used to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. The code analysis result proved that it could reasonably predict the FLB transient including the actuation of the PAFS and the natural convection.

  10. Integral effect test and code analysis on the cooling performance of the PAFS (passive auxiliary feedwater system) during an FLB (feedwater line break) accident

    International Nuclear Information System (INIS)

    Highlights: • This study focuses on the experimental validation of the operational performance of the PAFS (passive auxiliary feedwater system). • A transient simulation of the FLB (feedwater line break) in the integral effect test facility, ATLAS-PAFS, was performed to investigate thermal hydraulic behavior during the PAFS actuation. • The test result confirmed that the APR+ has the capability of coping with the FLB scenario by adopting the PAFS and proper set-points for its operation. • The experimental result was utilized to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. - Abstract: APR+ (Advanced Power Reactor Plus), which is a GEN-III+ nuclear power plant developed in Korea, adopts PAFS (passive auxiliary feedwater system) as an advanced safety feature. The PAFS can completely replace an active auxiliary feedwater system by cooling down the secondary side of steam generators with a natural convection mechanism. This study focuses on experimental and analytical investigation for cooling and operational performance of the PAFS during an FLB (feedwater line break) transient with an integral effect test facility, ATLAS-PAFS. To realistically simulate the FLB accident of the APR+, the three-level scaling methodology was taken into account to design the test facility and determine the test condition. From the test result, the PAFS was actuated to successfully cool down the decay heat of the reactor core by the condensation heat transfer at the PCHX (passive condensation heat exchanger), and thus it could be confirmed that the APR+ has the capability of coping with a FLB scenario by adopting the PAFS and proper set-points for its operation. This integral effect test data were used to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. The code analysis result proved that it could reasonably predict the FLB transient including the actuation of the PAFS and the natural convection

  11. 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

  12. Analysis fom the OECD/NEA PWR Main Steam Line Break (MSLB) Benchmark exercise 3 with the Coupled Code System RELAP5/PANBOX

    International Nuclear Information System (INIS)

    The main purpose of the computational OECD/NEA PWR MSLB-Benchmark is the evaluation of the prediction capability of advanced code systems by means of a code-to-code comparison. The postulated MSLB-transient is characterized by a strong non-symmetrical core thermal behaviour due to the feedback between neutron kinetics and plant thermal hydraulics. The analysis of such transients with pronounced spatial power distortion represents a considerable challenge for advanced code systems. It is initiated by a break of one main steam line when the reactor TMI-1 is operated at nominal power. High heat removal through the break leads to a strong cooldown rate of the broken loop compared to the intact one. Under such conditions a power increase and a re-criticality of the core despite scram can not be excluded due to the negative reactivity coefficients. The MSLB-Benchmark enfolds three exercises as follows: Exercise 1: integral plant simulation with best-estimate codes using the point kinetics, Exercise 2: multidimensional simulation of the core for given initial and boundary conditions, and Exercise 3: integral plant simulation with coupled, best-estimate codes using 3D-neutron kinetics models. Forschungszentrum Karlsruhe (FZK) and Framatome advanced nuclear power (ANP) Erlangen participated on the MSLB-Benchmark with the code system RELAP5/PANBOX for the Exercise 3: Based on the plant and core models elaborated for Exercise 1 and 2, an integral TMI-1 plant model was elaborated for Exercise 3. Special emphasis was put on the development of a multidimensional core model for the space-time kinetics. Two scenarios, the best-estimate (BE) and the return-to-power (RP) scenario, were investigated. Additional investigations aimed to investigate the influence of the coolant mixing on re-criticality and power increase. Results of these investigations are presented and discussed in this report. It has been demonstrated that RELAP5/PANBOX is capable to simulate complex transient in a

  13. MARADD - a computer code for semi-Markov reliability analysis of test/repair policies for standby safety systems

    International Nuclear Information System (INIS)

    In this paper, three test/repair policies are considered on the other redundant parts during the allowed outage time of the first failed component: (1) Policy 1 - a prompt additional test, (2) Policy 2 - no additional test, and (3) Policy 3 - an additional test after repairing the first failed one. For the analysis of the three test/repair policies, a computer program MARADD (Markov Reliability Analysis for the Additional Test Requirement) was developed in the present study, based on semi-Markov reliability analysis. The methodology was applied to the diesel generator system on which Policy 1 is currently performed. For the three test/repair policies, the system unavailability, the core damage frequency, and the plant shutdown frequency in one year of plant operation were calculated for several AOTs and STIs using the nominal input data and also using various input data for sensitivity studies

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

    International Nuclear Information System (INIS)

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

  15. Analysis of MELCOR code structure

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Ha; Park, Sun Hee

    2000-04-01

    MELCOR executes in two parts. The first is a MELGEN program, in which most of the input is specified, processed, and checked. The second part of MELCOR is the MELCOR program itself, which advances the program through time based on the database generated by MELGEN and any additional MELCOR input. In particular, MELCOR execution involves two steps: (1) a setup mode in MEXSET, during which the database is read from the restart file and any additional input is processed, and (2) a run mode in MEXRUN, which advances the simulation through time, updating the time-dependent portion of the database each cycle. MELGEN and MELCOR share a structured and modular architecture that facilitates the incorporation of additional or altenative phenomenological modes. This structure consists of four primary levels: executive level, database manager routine level, package level, and utility level. MELCOR is composed of 24 different packages, each of which models a different portion of the accident phenomenology or program control. To identify the relation of the MELCOR subroutines with the packages, first two or three letters of the package's name are duplicated in the name of the subroutines. The same rule applies to the naming of the common block. Data flows and the specific subroutines in the MELGEN and MELCOR are analyzed by their functions according to the hierarchy of four levels for model improvement and replacement during the integral code development project.

  16. Analysis of MELCOR code structure

    International Nuclear Information System (INIS)

    MELCOR executes in two parts. The first is a MELGEN program, in which most of the input is specified, processed, and checked. The second part of MELCOR is the MELCOR program itself, which advances the program through time based on the database generated by MELGEN and any additional MELCOR input. In particular, MELCOR execution involves two steps: (1) a setup mode in MEXSET, during which the database is read from the restart file and any additional input is processed, and (2) a run mode in MEXRUN, which advances the simulation through time, updating the time-dependent portion of the database each cycle. MELGEN and MELCOR share a structured and modular architecture that facilitates the incorporation of additional or altenative phenomenological modes. This structure consists of four primary levels: executive level, database manager routine level, package level, and utility level. MELCOR is composed of 24 different packages, each of which models a different portion of the accident phenomenology or program control. To identify the relation of the MELCOR subroutines with the packages, first two or three letters of the package's name are duplicated in the name of the subroutines. The same rule applies to the naming of the common block. Data flows and the specific subroutines in the MELGEN and MELCOR are analyzed by their functions according to the hierarchy of four levels for model improvement and replacement during the integral code development project

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

    Energy Technology Data Exchange (ETDEWEB)

    Ho, Clifford Kuofei

    2008-12-01

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

  18. Development of system analysis code for pyrochemical process using molten salt electrorefining part 2. Cathode processor calculation code with distillation process and parameter surveys using developed analytical model for cooling system of pyrochemical process cell

    International Nuclear Information System (INIS)

    This report describes accomplishment of development of a cathode processor calculation code to simulate the mass and heat transfer phenomena with the distillation process and parameter survey using developed analytical model for cooling behavior of the pyrochemical process cell on personal computers. The pyrochemical process using molten salt electrorefining would introduce new technologies for new fuels of particle oxide, particle nitride and metallic fuels. Evaporation calculations using cathode processor calculation code with distillation process, which was developed in 2000, were evaluated. By selecting proper input data (time step, mesh size etc.), the results showed that the present code agreed well for the evaporation rate of cadmium., and the capability of the distillation process design and simulation with the code has been confirmed. Parameter surveys using developed analytical model were performed for the purpose of reflection of cooling system design of the pyrochemical process cell. 4 cases of cooling flow patterns were surveyed at the normal and low flow rate conditions. From the result of parameter surveys, it was shown that the cooling pattern with direct cooling for heating facilities in the lower cell and balk cooling for upper cell is desirable. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

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

    International Nuclear Information System (INIS)

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

  1. Code system to compute radiation dose in human phantoms

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Westerberg, Arthur

    1978-12-04

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

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

    International Nuclear Information System (INIS)

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

  4. ETF system code: composition and applications

    International Nuclear Information System (INIS)

    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

  5. Estimating dosimetric quantities of radon progeny using human CT scan data and small tissue volume analysis with Geant4 code system

    Directory of Open Access Journals (Sweden)

    Van Den Akker Evelynn

    2015-01-01

    Full Text Available Estimating the health effects of radon exposure is of great interest because radon is considered the second leading cause of lung cancer after smoking. The dose-response curve is not well understood at low-dose levels where radon exposure is estimated. Therefore, the health mechanisms of radiation due to radon progeny at the cellular and molecular levels are of interest for providing an indication of a possible threshold value above which the exposure may indicate cancer formation. In this paper we present a macroscopic and cellular level numerical analysis of the radon-induced dose estimates based on the Geant4 code system. Macroscopic estimates are assessed based on patient-specific computer tomography scans that provide geometries easily applicable to modeling radiation effects of the radon progeny sources. A small tissue volumes analysis based on the Geant4 code system is developed so as to provide information about the interactions and particle track structures at the microscopic (cellular levels producing the dosimetric effects of radon short-lived progenies. The results presented in this paper also call attention to the capabilities of Geant4 to provide radon-related dosimetric parameters of large and small-scale biological systems.

  6. 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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  9. BWR plant dynamic analysis code BWRDYN user's manual

    International Nuclear Information System (INIS)

    Computer code BWRDYN has been developed for thermal-hydraulic analysis of a BWR plant. It can analyze the various types of transient caused by not only small but also large disturbances such as operating mode changes and/or system malfunctions. The verification of main analytical models of the BWRDYN code has been performed with measured data of actual BWR plant. Furthermore, the installation of BOP (Balance of Plant) model has made it possible to analyze the effect of BOP on reactor system. This report describes on analytical models and instructions for user of the BWRDYN code. (author)

  10. Code Coupling for Multi-Dimensional Core Transient Analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  11. Analysis of ZR-6 experiments using the KARATE-440 code system upgraded by ENDF/B-VI data

    International Nuclear Information System (INIS)

    Extensive validation of nuclear libraries was performed against lattice experiments with the help of the MULTICELL lattice calculation code and the COREMICRO 2D fine diffusion code. New cross-section library set was developed for these codes based on the ENDF/B-VI nuclear database. These codes belong to the KARATE-440 code package which is used for the calculation of VVER-440 reactor cores. The ZR-6 zero power critical facility was used as experimental database, where a wide range of hexagonal lattices were investigated. In the first step single-pin-cell calculations were performed for uniform critical ZR-6 lattices, where the axial and radial leakage were taken into account by the measured material buckling. In this step beside of criticality some spectral indices were investigated too. In the second step 2D calculations were applied for all the experimental configurations, including different temperatures. The ∂ρ/∂Η and ∂ρ/∂Τ coefficients were evaluated, too.(author)

  12. Code Formal Verification of Operation System

    Directory of Open Access Journals (Sweden)

    Yu Zhang

    2010-12-01

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

  13. Comparative study of Thermal Hydraulic Analysis Codes for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Various codes are used for the thermal hydraulic analysis of nuclear reactors. The use of some codes among these is limited by user and some codes are not even open to general person. Thus, the use of alternative code is considered for some analysis. In this study, simple thermal hydraulic behaviors are analyzed using three codes to show that alternative codes are possible for the analysis of nuclear reactors. We established three models of the simple u-tube manometer using three different codes. RELAP5 (Reactor Excursion and Leak Analysis Program), SPACE (Safety and Performance Analysis CodE for nuclear power Plants), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are selected for this analysis. RELAP5 is widely used codes for the analysis of system behavior of PWRs. SPACE has been developed based on RELAP5 for the analysis of system behavior of PWRs and licensing of the code is in progress. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. The internal behavior of u-tube manometer was analyzed by RELAP5, SPACE and GOTHIC codes. The general transient behavior was similar among 3 codes. However, the stabilized status of the transient period analyzed by REPAP5 was different from the other codes. It would be resulted from the different physical models used in the other codes, which is specialized for the multi-phase thermal hydraulic behavior analysis

  14. Comparative study of Thermal Hydraulic Analysis Codes for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yang Hoon; Jang, Mi Suk; Han, Kee Soo [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

    2015-05-15

    Various codes are used for the thermal hydraulic analysis of nuclear reactors. The use of some codes among these is limited by user and some codes are not even open to general person. Thus, the use of alternative code is considered for some analysis. In this study, simple thermal hydraulic behaviors are analyzed using three codes to show that alternative codes are possible for the analysis of nuclear reactors. We established three models of the simple u-tube manometer using three different codes. RELAP5 (Reactor Excursion and Leak Analysis Program), SPACE (Safety and Performance Analysis CodE for nuclear power Plants), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are selected for this analysis. RELAP5 is widely used codes for the analysis of system behavior of PWRs. SPACE has been developed based on RELAP5 for the analysis of system behavior of PWRs and licensing of the code is in progress. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. The internal behavior of u-tube manometer was analyzed by RELAP5, SPACE and GOTHIC codes. The general transient behavior was similar among 3 codes. However, the stabilized status of the transient period analyzed by REPAP5 was different from the other codes. It would be resulted from the different physical models used in the other codes, which is specialized for the multi-phase thermal hydraulic behavior analysis.

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

    International Nuclear Information System (INIS)

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

  16. Study of nuclear computer code maintenance and management system

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

    Taewan Kim; Victor Petrov; Annalisa Manera; Simon Lo

    2012-01-01

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

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

    International Nuclear Information System (INIS)

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

  20. Adjoint sensitivity analysis of the RELAPS/MOD3.2 two-fluid thermal-hydraulic code system

    International Nuclear Information System (INIS)

    This work presents the implementation of the Adjoint Sensitivity Analysis Procedure (ASAP) for the non-equilibrium, non-homogeneous two-fluid model, including boron concentration and non-condensable gases, of the RELAP5/MOD3.2 code. The end-product of this implementation is the Adjoint Sensitivity Model (ASM-REL/TF), which is derived for both the differential and discretized equations underlying the two-fluid model with non-condensable(s). The consistency requirements between these two representations are also highlighted. The validation of the ASM-REL/TF has been carried out by using sample problems involving: (i) liquid-phase only, (ii) gas-phase only, and (iii) two-phase mixture (of water and steam). Thus the 'Two-Loops with Pumps' sample problem supplied with RELAP5/MOD3.2 has been used to verify the accuracy and stability of the numerical solution of the ASM-REL/TF when only the liquid-phase is present. Furthermore, the 'Edwards Pipe' sample problem, also supplied with RELAP5/MOD3.2, has been used to verify the accuracy and stability of the numerical solution of the ASM-REL/TF when both (i.e., liquid and gas) phases are present. In addition, the accuracy and stability of the numerical solution of the ASM-REL/TF have been verified when only the gas-phase is present by using modified 'Two-Loops with Pumps' and the 'Edwards Pipe' sample problems in which the liquid and two-phase fluids, respectively, were replaced by pure steam. The results obtained for these sample problems depict typical sensitivities of junction velocities and volume-averaged pressures to perturbations in initial conditions, and indicate that the numerical solution of the ASM-REL/TF is as robust, stable, and accurate as the original RELAP5/MOD3.2 calculations. In addition, the solution of the ASM-REL/TF has been used to calculate sample sensitivities of volume-averaged pressures to variations in the pump head. (orig.)

  1. Centrifugal Compressor Aeroelastic Analysis Code

    Science.gov (United States)

    Keith, Theo G., Jr.; Srivastava, Rakesh

    2002-01-01

    Centrifugal compressors are very widely used in the turbomachine industry where low mass flow rates are required. Gas turbine engines for tanks, rotorcraft and small jets rely extensively on centrifugal compressors for rugged and compact design. These compressors experience problems related with unsteadiness of flowfields, such as stall flutter, separation at the trailing edge over diffuser guide vanes, tip vortex unsteadiness, etc., leading to rotating stall and surge. Considerable interest exists in small gas turbine engine manufacturers to understand and eventually eliminate the problems related to centrifugal compressors. The geometric complexity of centrifugal compressor blades and the twisting of the blade passages makes the linear methods inapplicable. Advanced computational fluid dynamics (CFD) methods are needed for accurate unsteady aerodynamic and aeroelastic analysis of centrifugal compressors. Most of the current day industrial turbomachines and small aircraft engines are designed with a centrifugal compressor. With such a large customer base and NASA Glenn Research Center being, the lead center for turbomachines, it is important that adequate emphasis be placed on this area as well. Currently, this activity is not supported under any project at NASA Glenn.

  2. CATHENA 4. A thermalhydraulics network analysis code

    International Nuclear Information System (INIS)

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

  3. MELCOR Accident Consequence Code System (MACCS)

    International Nuclear Information System (INIS)

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose 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

  4. MELCOR Accident Consequence Code System (MACCS)

    International Nuclear Information System (INIS)

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

  5. 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.

  6. Code Formal Verification of Operation System

    OpenAIRE

    Yu Zhang; Yunwei Dong; Huo Hong; Fan Zhang

    2010-01-01

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

  7. The development of LOCA analysis codes for nuclear power plant

    International Nuclear Information System (INIS)

    This research aims at assessment of the best-estimate codes, so as to develop a reliable analysis method for their actual applications. There are two additional purposes in the study: The first is the development of methodology for sizing the safety systems for advanced reactor design using the best-estimate codes, and the other is the development of our own best-estimate methodology, referring to USNRC approval of the acceptance criteria for ECCS based on the best-estimate method. The use of the best-estimate codes as those assumed in FSAR by only input arrangement, has resulted in achievement of at least 250 K safety margin. The fact that the predicted PCT in LBLOCA analysis is well bounded within the acceptance criteria using the best-estimates codes, should be verified by the quantification of the code uncertainty in the future. In the case of computer code improvement, the reflood models have been improved and satisfactory results have been obtained. In the case of uncertainty evaluation, the calculational matrices based on the assessment of experiments with the improved RELAP5 code for the quantification of the code uncertainty have been formulated separately for blowdown and reflood phases. (Author)

  8. The code system COROUT: Radioactive inventory calculations

    International Nuclear Information System (INIS)

    The code system COROUT is devoted to the evaluation of nuclear reactor out-of-core radioactive inventory for the sake of the nuclear power plant decommissioning problem. The code includes calculations of the neutron flux distributions and activation kinetics in the consistent way. Only thermal neutrons are taken into consideration in the present code version. Code is divided into three steps. The first step prepares the necessary data file containing data on reactor geometry, core flux, reactor operational history and data on elements in the out-of-core zones. The main part of calculations are performed during the second step. Here the thermal neutron flux distribution in the out-of-core area is calculated for two-dimensional cylindrical geometry and the system of gain-loss equations and the activation kinetics is solved for the elements in the different out-of-core shells. The Vladimirov's method of iterations on the spatial grid is used for the neutron flux calculations. The kinetic equations are solved by the operational method. The change of neutron field due to activation during reactor campaign is taken into account. The third part of COROUT code system allows to prepare plots of flux and activity distribution for different shells. All steps could be initiated independently using the results stored at the previous steps. The code is destined for the personal computers and has been written on the base of 32-bit FORTRAN language for IBM PC. 4 refs, 6 figs, 1 tab

  9. Data processing code system for foil experiments

    International Nuclear Information System (INIS)

    A code system has been developed for an efficient measurement of reaction rates in foil irradiation experiments. The code system consists of four codes, namely of, (i) setting up experimental parameters and collecting γ-ray spectrum data, (ii) analysing γ-ray spectrum, (iii) calculating reaction rate distributions, and (iv) furnishing utility programs. This code system provides a useful tool of data processing of irradiated foil to obtain the γ-ray spectrum and the reaction rate distribution. These procedures can be executed automatically. The routine for processing foil counting data covers the following functions : the data smoothing, the peak searching by means of the first and second derivative methods, and the determination of the photo peak area and its error with use of a functional fitted by a non-linear least squares method. The code for reaction rate calculation has the following functions : the determination of decay constants of each isotope by using decay data of foil counting and the calculation of reaction rates after correcting irradiation time and weight of a foil. These codes are written by FORTRAN-77 for mini-computer PDP-11/44 (DEC), of which the maximum program memory size is limited to 32k bytes. (author)

  10. Static Code Analysis with Gitlab-CI

    CERN Document Server

    Datko, Szymon Tomasz

    2016-01-01

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

  11. MHD stability analysis code ERATO-J

    International Nuclear Information System (INIS)

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

  12. Principles of the reactor code system RHEIN

    International Nuclear Information System (INIS)

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

  13. Results and experiences from the analysis of the OECD PWR MSLB and BWR TT benchmarks by the coupled code system ATHLET-QUABOX/CUBBOX

    International Nuclear Information System (INIS)

    The GRS results of the two QECD/NRC benchmarks for coupled codes - PWR Main Steam Line Break (MSLB) and BWR Turbine Trip (TT) are presented in the paper. Both benchmarks have been calculated for all exercises by the coupled code system ATHLET - QUABOX/CUBBOX developed by GRS. The OECD PWR Main Steam Line Break Benchmark has been defined to validate the coupled code systems with 3D neutronics by comparing solutions of different codes (code-to-code comparison). The BWR TT Benchmark is based on measurements. Both benchmarks proved to be a valuable source for coupled code validation. The sensitivity of results on modelling features is also discussed in the paper. In addition, the effect of different mapping schemes between fuel assemblies of the core loading and the thermal-fluid dynamics on the accuracy of 3D neutronics solutions is shown. The results for the MSLB transient are also evaluated to compare 3D neutronics and point-kinetics solutions in view of integral and local parameters. Thus, the experiences with the coupled code system ATHLET - QUABOX/CUBBOX during the MSLB and TT benchmarks activities are summarised. (author)

  14. TBL analysis by best estimate codes

    International Nuclear Information System (INIS)

    TRAC-BD1 (Version 12) is a three-dimensional thermal-hydraulic code for analyzing boiling water reactor (BWR) loss of coolant accidents (LOCAs). The code was developed by EG and G Idaho, Inc. and General Electric Co. (GE) under the sponsorship of the US Nuclear Regulatory Commission. SAFER, which was developed under the cooperative efforts of GE, Hitachi and Toshiba as a licensing code, is a one-dimensional thermal-hydraulic code for analyzing long term coolant inventory of the reactor system in BWR-LOCAs. Analyses for large and small break tests conducted with the Two-Bundle Loop (TBL) have been performed to assess the capabilities of these codes. The TBL was constructed to simulate thermal-hydraulic behaviors during LOCAs in a BWR/5 plant. It is the only integral test facility with two full size electrically heated bundles and consists of two full length jet pumps, vessel internals, two recirculation loops, two blowdown lines and emergency core cooling systems

  15. Development of Tritium Permeation Analysis Code (TPAC)

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-10-01

    Idaho National Laboratory developed the Tritium Permeation Analysis Code (TPAC) for tritium permeation in the Very High Temperature Gas Cooled Reactor (VHTR). All the component models in the VHTR were developed and were embedded into the MATHLAB SIMULINK package with a Graphic User Interface. The governing equations of the nuclear ternary reaction and thermal neutron capture reactions from impurities in helium and graphite core, reflector, and control rods were implemented. The TPAC code was verified using analytical solutions for the tritium birth rate from the ternary fission, the birth rate from 3He, and the birth rate from 10B. This paper also provides comparisons of the TPAC with the existing other codes. A VHTR reference design was selected for tritium permeation study from the reference design to the nuclear-assisted hydrogen production plant and some sensitivity study results are presented based on the HTGR outlet temperature of 750 degrees C.

  16. PWRDYN: a computer code for PWR plant dynamic analysis

    International Nuclear Information System (INIS)

    This report describes analytical models and calculated results of a PWR plant dynamic analysis code PWRDYN. The code has been developed in order to analyze and evaluate transient responses for small disturbance such as operating mode change and control system characteristic analysis. The features included in PWRDYN are 1) One loop approximation of primary loops, 2) Praimary coolant is always subcooled, 3) At the secondary side of steam generator is used one dimensional model and natural circulation is calculated assuming constant by positive driving head. 4) Main control systems are incorporated. In the transient responses caused by small perturbation, the calculated results by PWRDYN are in good agreement with the RETRAN calculations. Furthermore, computing time is very short so as about one seventh of real time, hence the code is convenient and useful for dynamic analysis of PWR plants. (author)

  17. Finite-Length Analysis of BATS Codes

    OpenAIRE

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

    2013-01-01

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

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

    International Nuclear Information System (INIS)

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

  19. Development of the SCS performance analysis and capacity evaluation code

    International Nuclear Information System (INIS)

    The Shutdown Cooling System (SCS) removes core decay heat during the planned plant shutdown or after the accident. A computer code such as DESCENT used by Combustion Engineering or RHRCOOL used by Westinghouse, is utilized to analyze the capacity and performance of the SCS for the system design of new plant and the replacement/repair of SCS heat exchanger of the operating reactors. These codes include approximated correlations for heat exchangers for the tube side flow ratio, total heat transfer coefficient, and the balance of the resistance constant calculated by the heat exchanger design codes, such as HTRI or HTFS. HTRI or HTFS does not have the capability to simulate the transient conditions of SCS. In this study, the SCS performance analysis and capacity evaluation (SPACE) code is developed to evaluate the total heat transfer coefficient for the heat exchanger as well as to analyze the SCS cooldown performance

  20. Performance analysis for a chaos-based code-division multiple access system in wide-band channel

    Directory of Open Access Journals (Sweden)

    Ciprian Doru Giurcăneanu

    2015-08-01

    Full Text Available Code-division multiple access technology is widely used in telecommunications and its performance has been extensively investigated in the past. Theoretical results for the case of wide-band transmission channel were not available until recently. The novel formulae which have been published in 2014 can have an important impact on the future of wireless multiuser communications, but limitations come from the Gaussian approximations used in their derivation. In this Letter, the authors obtain more accurate expressions of the bit error rate (BER for the case when the model of the wide-band channel is two-ray, with Rayleigh fading. In the authors’ approach, the spreading sequences are assumed to be generated by logistic map given by Chebyshev polynomial function of order two. Their theoretical and experimental results show clearly that the previous results on BER, which rely on the crude Gaussian approximation, are over-pessimistic.

  1. 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...

  2. Verification and uncertainty analysis of fuel codes using distributed computing

    International Nuclear Information System (INIS)

    Of late, nuclear safety analysis computer codes have been held to increasingly high standards of quality assurance. As well, best estimate with uncertainty analysis is taking a more prominent role, displacing to some extent the idea of a limit consequence analysis. In turn, these activities have placed ever-increasing burdens on available computing resources. A recent project at Ontario Hydro has been the development of the capability of using the workstations on our Windows NT LAN as a distributed batch queue. The application developed is called SheepDog. This paper reports on the challenges and opportunities met in this project, as well as the experience gained in applying this method to verification and uncertainty analysis of fuel codes. SheepDog has been applied to performing uncertainty analysis, in a basically CSAU like method, of fuel behaviour during postulated accident scenarios at a nuclear power station. For each scenario, several hundred cases were selected according to a Latin Hypercube scheme, and used to construct a response surface surrogate for the codes. Residual disparities between code predictions and response surfaces led to the suspicion that there were discontinuities in the predictions of the analysis codes. This led to the development of 'stress testing' procedures. This refers to two procedures: coarsely scanning through several input parameters in combination, and finely scanning individual input parameters. For either procedure, the number of code runs required is several hundred. In order to be able to perform stress testing in a reasonable time, SheepDog was applied. The results are examined for such considerations as continuity, smoothness, and physical reasonableness of trends and interactions. In several cases, this analysis uncovered previously unknown errors in analysis codes, and allowed pinpointing the part of the codes that needed to be modified. The challenges involved include the following: the usual choices of development

  3. Preliminary study of coupling CFD code FLUENT and system code RELAP5

    International Nuclear Information System (INIS)

    Highlights: • System code RELAP5/MOD3.1 is coupled with CFD code FLUENT through DLL and UDF. • Transient water flow in a simple straight tube is tested using the coupled tool. • Simulation of Edwards’ pipe blowdown experiment using the coupled tool is conducted. • Coupled analysis of a more comprehensive thermal–hydraulic system is performed. - Abstract: The present paper discusses a coupling strategy of the 3D (three-dimensional) computational fluid dynamics (CFD) code ANSYS-FLUENT with the best estimate 1D (one-dimensional) thermal–hydraulic system code RELAP5/MOD3.1. Preliminarily, by using DLL (Dynamic Link Library) technology and FLUENT UDF (User Defined Functions), an explicit coupling method expected to be able to support the analysis of multi-purpose thermal–hydraulic phenomena in nuclear reactor systems has been developed. Calculations for two test cases using the coupled FLUENT/RELAP5 code have been carried out to test and demonstrate the coupling capability: (i) the first one consisting of single-phase water transient flow in a square straight tube with well controlled mass flow rates; (ii) the second one illustrating the process of single-phase water flow in a system including two closed loops and one vessel, on which loss of loop water flow due to pump trip and increase of loop water temperature are studied. Both reasonable 1D systematic behaviors and 3D distribution information are naturally obtained for the test cases. Besides, a study of a highly transient experiment problem, i.e. Edwards–O’Brien pipe blowdown problem, has been performed by using the coupled FLUENT/RELAP5 code. The results are compared with standalone RELAP5 calculation and available experimental data, which shows the coupled FLUENT/RELAP5 code’s acceptable potential for the capability of analyzing either simple single-phase or complex two-phase flow problem

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

    International Nuclear Information System (INIS)

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

  5. Probabilistic structural analysis computer code (NESSUS)

    Science.gov (United States)

    Shiao, Michael C.

    1988-01-01

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

  6. Code comparison for accelerator design and analysis

    International Nuclear Information System (INIS)

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

  7. Code comparison for accelerator design and analysis

    International Nuclear Information System (INIS)

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

  8. Coding and transformations in the olfactory system.

    Science.gov (United States)

    Uchida, Naoshige; Poo, Cindy; Haddad, Rafi

    2014-01-01

    How is sensory information represented in the brain? A long-standing debate in neural coding is whether and how timing of spikes conveys information to downstream neurons. Although we know that neurons in the olfactory bulb (OB) exhibit rich temporal dynamics, the functional relevance of temporal coding remains hotly debated. Recent recording experiments in awake behaving animals have elucidated highly organized temporal structures of activity in the OB. In addition, the analysis of neural circuits in the piriform cortex (PC) demonstrated the importance of not only OB afferent inputs but also intrinsic PC neural circuits in shaping odor responses. Furthermore, new experiments involving stimulation of the OB with specific temporal patterns allowed for testing the relevance of temporal codes. Together, these studies suggest that the relative timing of neuronal activity in the OB conveys odor information and that neural circuits in the PC possess various mechanisms to decode temporal patterns of OB input. PMID:24905594

  9. Development and validation of I-activation analysis code

    International Nuclear Information System (INIS)

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

  10. MCNP-POLIMI v1.0, Monte Carlo N-Particle Transport Code System To Simulate Time-Analysis Quantities

    International Nuclear Information System (INIS)

    1 - Description of program or function: MCNP is a general-purpose, continuous-energy, generalized geometry, time-dependent, coupled neutron-photon-electron Monte Carlo transport code system. Based on the Los Alamos National Laboratory code MCNP4C (formerly distributed as CCC-700), MCNP-PoliMi was developed to simulate time-analysis quantities. In particular, the code includes the correlation between neutron interaction and the corresponding photon production. Conversely to the technique adopted by standard MCNP, MCNP PoliMi samples secondary photons according to the neutron collision type. A post-processing code, i.e. the Matlab script 'postmain', is included and can be tailored to model specific detector characteristics. These features make MCNP-PoliMi a versatile tool to simulate particle interactions and detection processes. 2 - Methods: MCNP treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces. For neutrons, all reactions in a particular cross-section evaluation are accounted for. Both free gas and S(alpha, beta) thermal treatments are used. Criticality sources as well as fixed and surface sources are available. For photons, the code takes account of incoherent and coherent scattering with and without electron binding effects, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. A very general source and tally structure is available. The tallies have extensive statistical analysis of convergence. Rapid convergence is enabled by a wide variety of variance reduction methods. Energy ranges are 0-60 MeV for neutrons (data generally only available up to 20 MeV) and 1 keV - 1 GeV for photons and electrons. The MCNP-PoliMi code was developed to simulate each neutron-nucleus interaction as closely as possible. In particular, neutron interaction and

  11. MELCOR Accident Consequence Code System (MACCS)

    International Nuclear Information System (INIS)

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

  12. 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.

  13. SIMULATE-3K linkage with reactor systems codes

    International Nuclear Information System (INIS)

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

  14. Best Estimate Thermal-Hydraulic System Analysis using the MARS Code for the Steam Generator Tube Rupture Accident in the APR1400

    International Nuclear Information System (INIS)

    A postulated SGTR (Steam Generator Tube Rupture) accident of the APR1400 was analysed using the best estimate safety analysis code, MARS (Multidimensional Analysis of Reactor Safety). The SGTR accident is one of the design basis accidents, which has a unique feature of the penetration of the barrier between the reactor coolant system (RCS) and the secondary system resulting from the failure of a steam generator U-tube. The SGTR has an importance in safety due to a concern of a containment bypass of radioactive inventory. In the course of the SGTR, the radioactivity leaking from a broken steam generator Utube mixes with the shell-side water in an affected steam generator. Leak flow from ruptured U-tubes can increase a water level and a pressure of the affected steam generator. Following a reactor trip and a turbine trip, the main steam safety valves (MSSVs) can be open to mitigate an increase in the secondary system pressure. Meanwhile, the SGTR can provide a direct flow path from the primary to the secondary system resulting in the release of fission products into the atmosphere. As one of the most limiting SGTR accidents, a leak flow equivalent to a double-ended rupture of five Utubes was analysed in this study. The main objective of this study is not only to provide physical insight into the system response of the APR1400 reactor during a SGTR but also to investigate the effect of reactor trip type of the HSGL (High Steam Generator Level) trip and the LPP (Low Pressurizer Pressure) trip on the thermal-hydraulic system response

  15. Bar-code automated waste tracking system

    International Nuclear Information System (INIS)

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

  16. Bar-code automated waste tracking system

    Energy Technology Data Exchange (ETDEWEB)

    Hull, T.E.

    1994-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-01

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

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

    International Nuclear Information System (INIS)

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

  19. Channel coding and modulation based on chaotic systems

    OpenAIRE

    Kozic, Slobodan; Hasler, Martin

    2007-01-01

    In this thesis, a new class of codes on graphs based on chaotic dynamical systems are proposed. In particular, trellis coded modulation and iteratively decodable codes on graphs are studied. The codes are designed by controlling symbolic dynamics of chaotic systems and using linear convolutional codes. The relation between symbolic dynamics of chaotic systems and trellis aspects to minimum distance properties of coded modulations is explained. Our arguments are supported by computer simulatio...

  20. Network Coding for Distributed Storage Systems

    CERN Document Server

    Dimakis, Alexandros G; Wu, Yunnan; Wainwright, Martin J; Ramchandran, Kannan

    2008-01-01

    Distributed storage systems provide reliable access to data through redundancy spread over individually unreliable nodes. Application scenarios include data centers, peer-to-peer storage systems, and storage in wireless networks. Storing data using an erasure code, in fragments spread across nodes, requires less redundancy than simple replication for the same level of reliability. However, since fragments must be periodically replaced as nodes fail, a key question is how to generate encoded fragments in a distributed way while transferring as little data as possible across the network. For an erasure coded system, a common practice to repair from a node failure is for a new node to download subsets of data stored at a number of surviving nodes, reconstruct a lost coded block using the downloaded data, and store it at the new node. We show that this procedure is sub-optimal. We introduce the notion of regenerating codes, which allow a new node to download \\emph{functions} of the stored data from the surviving ...

  1. Network Coded Multicast over Multibeam Satellite Systems

    OpenAIRE

    R. Alegre-Godoy; Vazquez-Castro, M. A.

    2015-01-01

    We propose a multicast scheme for multibeam satellite systems exploiting both the multiuser and spatial diversity inherent in this type of systems while taking into account realistic physical distributions of User Terminals (UTs) over the coverage. Our proposed scheme makes use of the well-known Adaptive Coding and Modulation (ACM) feature in Digital Video Broadcasting over Satellite, 2nd Generation (DVB-S2) and Extension (DVB-S2X) standards but also incorporates a set of innovative features....

  2. Two-phase flow characteristics analysis code: MINCS

    International Nuclear Information System (INIS)

    Two-phase flow characteristics analysis code: MINCS (Modularized and INtegrated Code System) has been developed to provide a computational tool for analyzing two-phase flow phenomena in one-dimensional ducts. In MINCS, nine types of two-phase flow models-from a basic two-fluid nonequilibrium (2V2T) model to a simple homogeneous equilibrium (1V1T) model-can be used under the same numerical solution method. The numerical technique is based on the implicit finite difference method to enhance the numerical stability. The code structure is highly modularized, so that new constitutive relations and correlations can be easily implemented into the code and hence evaluated. A flow pattern can be fixed regardless of flow conditions, and state equations or steam tables can be selected. It is, therefore, easy to calculate physical or numerical benchmark problems. (author)

  3. Performance improvement of spectral amplitude coding-optical code division multiple access systems using NAND detection with enhanced double weight code

    Science.gov (United States)

    Ahmed, Nasim; Aljunid, Syed Alwee; Ahmad, R. Badlishah; Fadhil, Hilal A.; Rashid, Mohd Abdur

    2012-01-01

    The bit-error rate (BER) performance of the spectral amplitude coding-optical code division multiple access (SACOCDMA) system has been investigated by using NAND subtraction detection technique with enhanced double weight (EDW) code. The EDW code is the enhanced version of double weight (DW) code family where the code weight is any odd number and greater than one with ideal cross-correlation. In order to evaluate the performance of the system, we used mathematical analysis extensively along with the simulation experiment. The evaluation results obtained using the NAND subtraction detection technique was compared with those obtained using the complementary detection technique for the same number of active users. The comparison results revealed that the BER performance of the system using NAND subtraction detection technique has greatly been improved as compared to the complementary technique.

  4. Matlab Code for Structural Decomposition Analysis

    OpenAIRE

    Juan Tomas Sayago-Gomez

    2014-01-01

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

  5. Operational reactor physics analysis codes (ORPAC)

    International Nuclear Information System (INIS)

    For efficient, smooth and safe operation of a nuclear research reactor, many reactor physics evaluations are regularly required. As part of reactor core management the important activities are maintaining core reactivity status, core power distribution, xenon estimations, safety evaluation of in-pile irradiation samples and experimental assemblies and assessment of nuclear safety in fuel handling/storage. In-pile irradiation of samples requires a prior estimation of the reactivity load due to the sample, the heating rate and the activity developed in it during irradiation. For the safety of personnel handling irradiated samples the dose rate at the surface of shielded flask housing the irradiated sample should be less than 200 mR/Hr.Therefore, a proper shielding and radioactive cooling of the irradiated sample are required to meet the said requirement. Knowledge of xenon load variation with time (Startup-curve) helps in estimating Xenon override time. Monitoring of power in individual fuel channels during reactor operation is essential to know any abnormal power distribution to avoid unsafe situations. Complexities in the estimation of above mentioned reactor parameters and their frequent requirement compel one to use computer codes to avoid possible human errors. For efficient and quick evaluation of parameters related to reactor operations such as xenon load, critical moderator height and nuclear heating and reactivity load of isotope samples/experimental assembly, a computer code ORPAC (Operational Reactor Physics Analysis Codes) has been developed. This code is being used for regular assessment of reactor physics parameters in Dhruva and Cirus. The code ORPAC written in Visual Basic 6.0 environment incorporates several important operational reactor physics aspects on a single platform with graphical user interfaces (GUI) to make it more user-friendly and presentable. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-09-14

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

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

    International Nuclear Information System (INIS)

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

  8. The development of a severe accident analysis code

    International Nuclear Information System (INIS)

    For prevention and mitigation of the containment failure during severe accident, the study is focused on the severe accident phenomena, especially, the ones occurring inside the cavity in an effect to improve existing models and develop analytical tools for the assessment of severe accidents. For hydrogen control, the analysis of hydrogen concentration in the containment and visualization for the concentration in the cell were performed. The computer code to predict combustion flame characteristic was also developed. the analytical model for the expansion phase of vapor explosion was developed and verified with the experimental results. The corium release fraction model from the cavity with the capture volume was developed and applied to the power plants. Pre-test calculation was performed for molten corium concrete interaction study and the crust formation process, heat transfer characteristics of the crust, and the sensitivity study using MELCOR code was carried out. A stress analysis code using finite element method for the reactor vessel lower head failure analysis was developed and the effect by gap formation between molten corium and vessel was analyzed. Through the international program of PHEBUS-FP and participation in the software development, the study on fission products release and transportation in the software development, the study on fission products release and transportation and aerosol deposition were performed. The system for severe accident analysis codes, CONTAIN and MELCOR codes etc., under the cooperation with USNRC were also established by installing in workstation and applying to experimental results and real plants. (author). 116 refs., 31 tabs., 59 figs

  9. A Students Attendance System Using QR Code

    Directory of Open Access Journals (Sweden)

    Fadi Masalha

    2014-01-01

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

  10. Revised SWAT. The integrated burnup calculation code system

    International Nuclear Information System (INIS)

    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)

  11. 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)

  12. BCH codes for large IC random-access memory systems

    Science.gov (United States)

    Lin, S.; Costello, D. J., Jr.

    1983-01-01

    In this report some shortened BCH codes for possible applications to large IC random-access memory systems are presented. These codes are given by their parity-check matrices. Encoding and decoding of these codes are discussed.

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

    OpenAIRE

    Evgeny, G.

    2008-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Taewan Kim

    2012-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  17. HELIAS module development for systems codes

    International Nuclear Information System (INIS)

    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

  18. 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.

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

    Institute of Scientific and Technical Information of China (English)

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

    2008-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  2. A mean field theory of coded CDMA systems

    International Nuclear Information System (INIS)

    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

  3. 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.

  4. Simulation of some plant transients by the coupled code system ATHLET/KIKO3D

    International Nuclear Information System (INIS)

    The assessment of coupled reactor physics and thermal-hydraulics computations with the coupled KIKO3D-ATHLET code system is provided, from two stand-alone codes. The details of data flow in the coupling are reviewed and some selected results of the validation are described. The validated coupled system code is used in the safety analysis for VVER reactors. (author)

  5. Analysis of cell-survival fractions for heavy-ion irradiations based on microdosimetric kinetic model implemented in the particle and heavy ion transport code system

    International Nuclear Information System (INIS)

    It is considered that the linear energy transfer (LET) may not be the ideal index for expressing the relative biological effectiveness (RBE) of cell killing for heavy-ion irradiation, as the ion-species dependencies have clearly been observed in the relation between LET and RBE derived from cell-survival fraction data. The previously measured survival fractions of four cell lines irradiated by various ion species, employing the saturation-corrected dose-mean lineal energy, γ*, instead of LET as the index of the RBE were therefore re-analysed. In the analysis, the initial slopes of the survival fractions, the so-called α-parameter in the linear-quadratic model, were plotted as a function of γ*, which was calculated by the microdosimetric kinetic (MK) model implemented in the Particle and Heavy Ion Transport code System. It was found from the analysis that the ion-species dependencies observed in the relations between α and LET disappeared from those between α and γ*, and their relations can be well reproduced by a simple equation derived from the MK model. These results clearly indicate the suitability of γ* to be used in the estimation of the RBE of cell killing for heavy-ion irradiations, which is of great importance in the treatment planning of charged-particle therapy. (authors)

  6. A Students Attendance System Using QR Code

    OpenAIRE

    Fadi Masalha; Nael Hirzallah

    2014-01-01

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

  7. Burnup calculation code system COMRAD96

    International Nuclear Information System (INIS)

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

  8. Cooperative Regenerating Codes for Distributed Storage Systems

    OpenAIRE

    Shum, Kenneth W.

    2011-01-01

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

  9. 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).

  10. SRAC95; general purpose neutronics code system

    International Nuclear Information System (INIS)

    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)

  11. Suppression pool swell analysis using CFD code

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  13. Impact of modeling effects, initial and boundary conditions on performing ATWS analysis with the coupled system code ATHLET/BIPR-VVER

    International Nuclear Information System (INIS)

    The work demonstrates the successful application of coupled thermal-hydraulics/neutron-kinetics system codes by performing analyses of complex transients. Two simulation cases (Case no.1 and Case no.2) are compared for a NPP with VVER-1000 reactor (type V-320). The two cases differ in core layout, the initial and boundary conditions and in the nodalization schemas of the reactor pressure vessel. The main objective is to identify the importance of modelling differences on main NPP parameter histories for an ATWS case with loss of main feed water. This comparison can contribute to further developments and optimizations by performing safety analyses with coupled codes. The analyses have been carried out with the coupled system code ATHLET/BIPR-VVER, developed to perform best estimate simulations of three-dimensional neutron-kinetics and thermal-hydraulics processes in VVER reactors. (author)

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

    Directory of Open Access Journals (Sweden)

    Ali Naseri

    2011-11-01

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

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

    OpenAIRE

    Sari, Dewi Maya

    2015-01-01

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

  16. 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.

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

    International Nuclear Information System (INIS)

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

  18. Development and Assessment of Best Estimate Integrated Safety Analysis Code

    International Nuclear Information System (INIS)

    The integrated safety analysis code MARS3.0 has been developed and assessed through v and v procedure. Integrated safety analysis system has been established through coupling with severe accident code and utilizing MARS subchannel capability. The coupled containment module has been also improved. Development of indigenous thermal hydraulic models for MARS3.0 has been done through the implementation of multidimensional two phase flow model, APR1400, SMART safety issue models and new reactor models. Development of droplet field model has been also attempted and implemented to trial version. The full scope assessment has been carried out for the system analysis module and 3D vessel module. The code has been also assessed through participating international cooperation programs. The experimental data needed to code assessment has been collected and maintained through the WEB based data bank program. 3D GUI(graphic user interface) has been developed for MARS users. MARS users group has been organized, and currently it consists of 22 domestic organizations, including research, industrial, regulatory organizations and universities

  19. Digital Image Analysis for Detechip Code Determination

    Directory of Open Access Journals (Sweden)

    Marcus Lyon

    2012-09-01

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

  20. Digital Image Analysis for Detechip Code Determination

    Directory of Open Access Journals (Sweden)

    Marcus Lyon

    2012-08-01

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

  1. Analysis of SMA hybrid composite structures using commercial codes

    Science.gov (United States)

    Turner, Travis L.; Patel, Hemant D.

    2004-07-01

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

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

    Institute of Scientific and Technical Information of China (English)

    陈煌林

    2011-01-01

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

  3. Computer codes for safety analysis of Indian PHWRs

    International Nuclear Information System (INIS)

    Computer codes for safety analysis of PHWRs have been developed in India over the years. Some of the codes that have been developed in NPC are discussed in this paper. Computer code THYNAC and ATMIKA have been developed in NPC for the analysis of LOCA scenario. Both the codes are based on UVET model using three equations and slip correlations. Computer code ATMIKA is an improved version of code THYNAC with regard to numerics and flexibility in modelling. Apart from thermal hydraulic model these codes also include point neutron kinetics model. Codes COOLTMP and RCOMP are used to estimate heat-up of primary coolant and core components respectively under off-normal shutdown conditions as may be existing during special maintenance job or postulated failure. Code validations have been performed either against experiments or the published results of experiments performed elsewhere, or through International benchmark exercises sponsored by IAEA. The paper discusses these codes, their validations and salient applications

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

    DEFF Research Database (Denmark)

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

    2015-01-01

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

  5. Modeling of transients with the GRS system code ATHLET

    International Nuclear Information System (INIS)

    The code ATHLET is actually being developed at GRS. It is a fusion of the codes ALMOD, DRUFAN, and FLUT within a single, consistent code structure. The reasons for the ATHLET development are explained, and the code version to be used for analysis of anticipated and abnormal transients is described. The status of development and assessment is summarized

  6. Problems of optimal data coding in hodoscopic systems

    International Nuclear Information System (INIS)

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

  7. Comparison of Activation Analysis Codes between ORIGEN-S and FLUKA

    International Nuclear Information System (INIS)

    A status of activation on the target is analyzed through the activation code. New nuclides were generated by neutron irradiation on the target. The energy of the nuclides has an effect on other materials. Therefore information on the nuclides is required. The radioactivity intensity and the kind of nuclides are measured through an activation analysis. An activation analysis was conducted on the target materials. Generally, Origen-s code is used in the activation analysis. An activation analysis of the Tantalum target was performed using MCNPX and Origen-s combination code. The results were compared with those of the FLUKA code. The Origen-s and FLUKA code simulation results are provided for a comparison with the activation analysis code. A Slowing Down Time Spectrometer (SDTS) system is a highly efficient technique in a nuclear material analysis. An activation analysis on the shielding and target material was required for the SDTS system. The simulation results differed in terms of the radioactivity intensity of the nuclides. In addition, the nuclides generated were different. The Origen-s code showed a consistent trend with the half-life However, the FLUKA code was not satisfied. A study on the FLUKA code and Origen-s code is necessary to complement this problem. Simulation results are provided to select the activation analysis code

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

    Science.gov (United States)

    Fittje, James E.

    2007-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-06-01

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

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

    International Nuclear Information System (INIS)

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

  11. A novel method for performance improvement of optical CDMA system using alterable concatenated code

    Science.gov (United States)

    Qiu, Kun; Zhang, Chongfu

    2007-04-01

    A novel method using alterable concatenated code to pre-encode is proposed to reduce the impact of system impairment and multiple access interference (MAI) in optical code division multiple access (OCDMA) system, comprehensive comparisons between different concatenated code type and forward error correcting (FEC) scheme are studied by simulation. In the scheme, we apply concatenated coding to the embedded modulation scheme, and optical orthogonal code (OOC) is employed as address sequence code, an avalanche photodiode (APD) is selected as the system receiver. The bit error rate (BER) performance is derived taking into account the effect of some noises, dispersion power penalty and the MAI. From both theoretical analysis and numerical results, we can show that the proposed system has good performance at a BER of 10 -9 with a gain of 6.4 dB improvement achieved using the concatenated code as the pre-code, and this scheme permits implementation of a cost effective OCDMA system.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    White, J.R.

    1985-04-01

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

  14. MHTGR-350 Benchmark Analysis by MCS Code

    International Nuclear Information System (INIS)

    This benchmark contains various problems in three phases, which require the results for neutronics, thermal fluids solutions, transient calculation, and depletion calculation. The Phase-I exercise-1 problem was solved with MCS Monte Carlo (MC) code developed at UNIST. The global parameters and power distribution was compared with the results of McCARD MC code developed by SNU and a finite element method (FEM) - based diffusion code CAPP developed by KAERI. The MHTGR-350 benchmark Phase-I exercise 1 was solved with MCS. The results of MCS are compared with those of McCARD and CAPP. The results of MCS code showed good agreements with those of McCARD code while they showed considerable disagreements with those of CAPP code, which can be attributed to the fact that CAPP is a diffusion code while the others are MC transport codes

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-03-15

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

  17. Code system BCG for gamma-ray skyshine calculation

    International Nuclear Information System (INIS)

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

  18. Qualification of the nuclear reactor core model DYN3D coupled to the thermohydraulic system code ATHLET, applied as an advanced tool for accident analysis of VVER-type reactors. Final report

    International Nuclear Information System (INIS)

    The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.)

  19. Concatenated coding system with iterated sequential inner decoding

    DEFF Research Database (Denmark)

    Jensen, Ole Riis; Paaske, Erik

    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...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-10-15

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

  1. Health effects estimation code development for accident consequence analysis

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  3. MTR coded PRML systems for perpendicular magnetic recording

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-10-01

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

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

    International Nuclear Information System (INIS)

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

  5. Optimization and analysis of code-division multiplexed TES microcalorimeters

    CERN Document Server

    Fowler, J W; Hilton, G C; Irwin, K D; Schmidt, D R; Stiehl, G M; Swetz, D S; Ullom, J N; Vale., L R

    2011-01-01

    We are developing code-division multiplexing (CDM) systems for transition-edge sensor arrays with the goal of reaching multiplexing factors in the hundreds. We report on x-ray measurements made with a four-channel prototype CDM system that employs a flux-summing architecture, emphasizing data-analysis issues. We describe an empirical method to determine the demodulation matrix that minimizes cross-talk. This CDM system achieves energy resolutions of between 2.3 eV and 3.0 eV FWHM at 5.9 keV.

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

    Institute of Scientific and Technical Information of China (English)

    Hao Jie; Lu Yanbo; Liu Xinji; Xia Shutao

    2013-01-01

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

  7. Computer code for calculating reliability/availability of technical systems

    International Nuclear Information System (INIS)

    Three computer codes are reviewed, which can be applied to reliability analyses of technical systems. They are based on the fault tree and the laws of probability theory. The codes can be used for both non-repairable and repairable systems. The simulation code REMO 79 and the analytical code RELAV are based on the conception that a failure of system components is immediately detected and repaired. The model of the FUPRO2 code provides for failures to be detected and repaired only in periodic functional tests. Apart from code descriptions experience and far-reaching aspects resulting from modularization of the fault trees are summarized. (author)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-11-01

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

  10. Diffuser augmented wind turbine analysis code

    Science.gov (United States)

    Carroll, Jonathan

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

  11. MORSE Monte Carlo radiation transport code system

    International Nuclear Information System (INIS)

    This report is an addendum to the MORSE report, ORNL-4972, originally published in 1975. This addendum contains descriptions of several modifications to the MORSE Monte Carlo Code, replacement pages containing corrections, Part II of the report which was previously unpublished, and a new Table of Contents. The modifications include a Klein Nishina estimator for gamma rays. Use of such an estimator required changing the cross section routines to process pair production and Compton scattering cross sections directly from ENDF tapes and writing a new version of subroutine RELCOL. Another modification is the use of free form input for the SAMBO analysis data. This required changing subroutines SCORIN and adding new subroutine RFRE. References are updated, and errors in the original report have been corrected

  12. Validation of OPERA3D PCMI Analysis Code

    International Nuclear Information System (INIS)

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

  13. Analysis and Design of Tuned Turbo Codes

    CERN Document Server

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

    2010-01-01

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

  14. CORRELATING FEATURES AND CODE BY DYNAMIC AND SEMANTIC ANALYSIS

    Directory of Open Access Journals (Sweden)

    Ren Wu

    2015-10-01

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

  15. Analytical validation of the CACECO containment analysis code

    International Nuclear Information System (INIS)

    The CACECO containment analysis code was developed to predict the thermodynamic responses of LMFBR containment facilities to a variety of accidents. This report covers the verification of the CACECO code by problems that can be solved by hand calculations or by reference to textbook and literature examples. The verification concentrates on the accuracy of the material and energy balances maintained by the code and on the independence of the four cells analyzed by the code so that the user can be assured that the code analyses are numerically correct and independent of the organization of the input data submitted to the code

  16. A Content Analysis of Student Conduct Codes

    OpenAIRE

    Martin, Janice Earlene

    2004-01-01

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

  17. Computer codes and methods for simulating accelerator driven systems

    International Nuclear Information System (INIS)

    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. Layered Low-Density Generator Matrix Codes for Super High Definition Scalable Video Coding System

    Science.gov (United States)

    Tonomura, Yoshihide; Shirai, Daisuke; Nakachi, Takayuki; Fujii, Tatsuya; Kiya, Hitoshi

    In this paper, we introduce layered low-density generator matrix (Layered-LDGM) codes for super high definition (SHD) scalable video systems. The layered-LDGM codes maintain the correspondence relationship of each layer from the encoder side to the decoder side. This resulting structure supports partial decoding. Furthermore, the proposed layered-LDGM codes create highly efficient forward error correcting (FEC) data by considering the relationship between each scalable component. Therefore, the proposed layered-LDGM codes raise the probability of restoring the important components. Simulations show that the proposed layered-LDGM codes offer better error resiliency than the existing method which creates FEC data for each scalable component independently. The proposed layered-LDGM codes support partial decoding and raise the probability of restoring the base component. These characteristics are very suitable for scalable video coding systems.

  19. Calculated thermal-hydraulic response for Semiscale Mod-3 Test S-07-6 using RELAP5: a new LWR system analysis code

    International Nuclear Information System (INIS)

    The newly developed, advanced, light water reactor (LWR) simulation code, RELAP5, is used to analyze the response of Semiscale Mod-3 Test S-07-6. The objective of Test S-07-6 was to provide reference data to evaluate LWR integral blowdown, refill, and reflood behavior during a 200% cold leg break with emergency core coolant (ECC) injected into the intact loop cold leg. The calculated test results using RELAP5 illustrate many of the nonequilibrium and nonhomogeneous aspects of the ECC injection which are not directly observable in the test data. These results also demonstrate the capability of the RELAP5 code and compare well with the test data fo break flow, pressure, temperature, and density throughout the Semiscale Mod-3 system. The periodic depletion and replenishment of ECC water in the downcomer shown in the test data is also shown in the calculation

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

    International Nuclear Information System (INIS)

    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)

  1. 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.

  2. The analysis of thermal-hydraulic models in MELCOR code

    International Nuclear Information System (INIS)

    The objective of the present work is to verify the prediction and analysis capability of MELCOR code about the progression of severe accidents in light water reactor and also to evaluate appropriateness of thermal-hydraulic models used in MELCOR code. Comparing the results of experiment and calculation with MELCOR code is carried out to achieve the above objective. Specially, the comparison between the CORA-13 experiment and the MELCOR code calculation was performed

  3. 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.

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

    International Nuclear Information System (INIS)

    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. Simplified computer codes for cask impact analysis

    International Nuclear Information System (INIS)

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

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

    Institute of Scientific and Technical Information of China (English)

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

    2006-01-01

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

  7. Analysis of LWHCR-PROTEUS Phase 2 experiments performed using the AARE system with JEF-1 based data libraries, and comparison with other codes

    International Nuclear Information System (INIS)

    In this report the capability of the AARE modular code system and JEF-1 based nuclear data libraries to analyse LWHCR lattices is investigated by calculating the wet and dry cells of the PROTEUS-LWHCR Phase 2 experiment. The results are compared to those obtained using several cell codes, including WIMS-D, BOXER, MICROX-2, KARBUS, GRUCAH, and SPEKTRA. In particular, the main features of AARE, such as the self-shielding of resonance cross sections in the whole energy range of importance for structural materials and actinides (including the low energy resonances of heavy actinides), the shielding of oxygen resonances in the MeV range, the generation of adequate fission source spectra, the accurate calculation of migration areas, and the efficiency of the removal correction are investigated. It is shown that AARE can predict the k∞ void coefficient well with a 1 % deviation from experiment, even if a coarse 70 netron group library is used. KARBUS and the related 69 group KEDAK-4 library give as well a reliable estimate, but lead to less accurate prediction of reaction rates. The other codes give larger deviations. The JEF-1 evaluation for 242Pu gives systematically about 25 % too high capture rates in the fast energy range (above 1 keV). (author) 39 refs., 24 figs., 13 tabs

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

    International Nuclear Information System (INIS)

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

  9. Parallelization of Subchannel Analysis Code MATRA

    International Nuclear Information System (INIS)

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

  10. Parallelization of Subchannel Analysis Code MATRA

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  11. Performance Evaluation of Space-Time Turbo Code Concatenated With Block Code MC-CDMA Systems

    OpenAIRE

    Lokesh Kumar Bansal; Aditya Trivedi

    2011-01-01

    In this paper, performance of a space-time turbo code (STTuC) in concatenation with space-time block code (STBC) in multi-carrier code-division multiple-access (MCCDMA) system with multi-path fading channel is considered. The performance in terms of bit error rate (BER) is evaluated through simulations. The corresponding BER of the concatenated STTuC-STBC-MC-CDMA system is compared with STTuC-MC-CDMA system and STBC-MC-CDMA system. The simulation results show that the STTuC-MCCDMA system perf...

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  16. ANACROM - A computer code for chromatogram analysis

    International Nuclear Information System (INIS)

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

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

    Institute of Scientific and Technical Information of China (English)

    周侃; 金松坡

    2012-01-01

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

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

    Institute of Scientific and Technical Information of China (English)

    PANG Yu-ye; SUN Jun; WANG Jia

    2007-01-01

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

  19. Communication Systems Simulator with Error Correcting Codes Using MATLAB

    Science.gov (United States)

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

    2003-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  2. 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…

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

    Science.gov (United States)

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

    2006-01-01

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

  4. Development status of Severe Accident Analysis Code SAMPSON

    International Nuclear Information System (INIS)

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

  5. Development status of Severe Accident Analysis Code SAMPSON

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-11-01

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

  6. Universal optical line terminal encoding and decoding architecture in two-code keying for noncoherent spectral amplitude coding optical code division multiple access systems

    Science.gov (United States)

    Yeh, Bih-Chyun; Lin, Cheing-Hong; Yang, De-Nian

    2014-01-01

    We propose a new code family, called extended shifted prime codes, and the universal encoding architecture for spectral amplitude coding optical code division multiple access systems using a two-code keying scheme. The proposed system can eliminate multiuser interference and suppress phase-induced intensity noise. In addition, we design the ESP codes to be an encoding/decoding architecture based on the array waveguide grating architecture and reduce the power loss and the complexity of the optical line terminal. The numerical results demonstrate that the proposed system with ESP codes outperforms the existing one-dimensional shifted prime codes system.

  7. INTERF: the reaction rates and spectra editing code for analysis of fusion neutronics experiments

    International Nuclear Information System (INIS)

    The reaction rates and spectra editing code INTERF has been developed for transport calculation codes as a part of the analysis system for fusion neutronics experiments. This code can provide the ratio of calculation to experiment value for reaction rate (C/E), spectra, reaction rate distributions, contour distributions, etc. from results of transport calculation. The transport calculation codes that INTERF can process are the ANISN, DOT3.5, BERMUDA-2DN, MCNP and MORSE-DD codes. In this report, the concept, functions, input data, and input/output files in INTERF are described and the examples of input data for usage of INTERF are shown. (author)

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

    International Nuclear Information System (INIS)

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

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

    DEFF Research Database (Denmark)

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

    2016-01-01

    challenge in systems operating in the millimeter wave regime. This work takes the WCC one step beyond by performance evaluation in terms of bit error probability, over time-varying, frequency-selective multipath Rayleigh fading channels. The adopted propagation model follows the COST207 norm, the main......, such systems use diversity schemes in combination with digital signal processing (DSP) techniques to overcome effects such as fading and inter-symbol interference (ISI). Wavelet Channel Coding (WCC) has emerged as a technique to minimize the fading effects of wireless channels, which is a mayor...... international standard reference for GSM, UMTS, and EDGE applications. The results show how WCC can be efficient against ISI. To the best of our knowledge this is the first time WCC is considered on Radio-over-Fiber transmissions at mm-wave range....

  10. Automatic counterfeit protection system code classification

    Science.gov (United States)

    Van Beusekom, Joost; Schreyer, Marco; Breuel, Thomas M.

    2010-01-01

    Wide availability of cheap high-quality printing techniques make document forgery an easy task that can easily be done by most people using standard computer and printing hardware. To prevent the use of color laser printers or color copiers for counterfeiting e.g. money or other valuable documents, many of these machines print Counterfeit Protection System (CPS) codes on the page. These small yellow dots encode information about the specific printer and allow the questioned document examiner in cooperation with the manufacturers to track down the printer that was used to generate the document. However, the access to the methods to decode the tracking dots pattern is restricted. The exact decoding of a tracking pattern is often not necessary, as tracking the pattern down to the printer class may be enough. In this paper we present a method that detects what CPS pattern class was used in a given document. This can be used to specify the printer class that the document was printed on. Evaluation proved an accuracy of up to 91%.

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

    Science.gov (United States)

    Di Vito, Ben. L

    2016-01-01

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

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

    Institute of Scientific and Technical Information of China (English)

    钱鸿涛; 李政昕; 胡文军; 宫宇

    2015-01-01

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

  13. Nuclear modules of ITER tokamak systems code

    International Nuclear Information System (INIS)

    Nuclear modules were developed to model various reactor components in the ITER systems code. These modules include first wall, tritium breeding blanket (or shield), bulk shield, reactor vault, impurity control, and tritium system. The function of these modules is to define the performance parameters for each component as a function of the reactor operating conditions. Several design options and cost algorithms are included for each component. The first wall, blanket and shield modules calculate the beryllium zone thickness, the disruptions results, the nuclear responses in different components including the toroidal field coils. Tungsten shield/water coolant/steel structure and steel shield/water coolant are the shield options for the inboard and outboard sections of the reactor. Lithium nitrate dissolved in the water coolant with a variable beryllium zone thickness in the outboard section of the reactor provides the tritium breeding capability. The reactor vault module defines the thickness of the reactor wall and the roof based on the dose equivalent during operation including skyshine contribution. The impurity control module provides the design parameters for the divertor including plate design, heat load, erosion rate, tritium permeation through the plate material to the coolant, plasma contamination by sputtered impurities, and plate lifetime. Several materials: Be, C, V, Mo, and W can be used for the divertor plate to cover a range of plasma edge temperatures. The tritium module calculates tritium and deuterium flow rates for the reactor plant. The tritium inventory in the fuelers, neutral beams, vacuum pumps, impurity control, first wall, and blanket is calculated. Tritium requirements are provided for different operating conditions. The nuclear models are summarized in this paper including the different design options and key analyses of each module

  14. Pipe whip analysis using the TEDEL code

    International Nuclear Information System (INIS)

    In view of their abundance, piping systems are one of the main components in power industries and in particular in nuclear power plants. They must be designed for normal as well as faulted conditions, for safety requirements. The prediction of the dynamic behaviour of the free pipe requires accounting for several nonlinearities. For this purpose, a beam type finite element program (TEDEL) has been used. The aim of this paper is to enlight the main features of this program, when applied to pipe whip analysis. An example of application to a real case will also be presented

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-01-12

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

  16. Selected examples for safety analysis in VVER-440 type reactors simulated by the coupled ATHLET/KIKO3D code system

    International Nuclear Information System (INIS)

    Recently several projects have been initiated in Hungary aiming at the introduction of new fuel type, increased maximum allowed power and economic fuel cycle. The planned upgraded power and parallel application of new fuel type require the renewal of the relevant chapter of the Final Safety Analysis Report (FSAR). One of the main tools used for analyzing transient scenarios initiating by reactivity and power distribution anomalies was the ATHLET/KIKO3D coupled neutron kinetic / thermal-hydraulic code. This paper gives an overview of two analyses, which was prepared in the frame of the revision of Paks FSAR, namely the ''withdrawal of one control rod'' and ''initial phase of main steam line break'' events. (author)

  17. Morse Monte Carlo Radiation Transport Code System

    Energy Technology Data Exchange (ETDEWEB)

    Emmett, M.B.

    1975-02-01

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

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

    International Nuclear Information System (INIS)

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

  19. Vision aided inertial navigation system augmented with a coded aperture

    Science.gov (United States)

    Morrison, Jamie R.

    Navigation through a three-dimensional indoor environment is a formidable challenge for an autonomous micro air vehicle. A main obstacle to indoor navigation is maintaining a robust navigation solution (i.e. air vehicle position and attitude estimates) given the inadequate access to satellite positioning information. A MEMS (micro-electro-mechanical system) based inertial navigation system provides a small, power efficient means of maintaining a vehicle navigation solution; however, unmitigated error propagation from relatively noisy MEMS sensors results in the loss of a usable navigation solution over a short period of time. Several navigation systems use camera imagery to diminish error propagation by measuring the direction to features in the environment. Changes in feature direction provide information regarding direction for vehicle movement, but not the scale of movement. Movement scale information is contained in the depth to the features. Depth-from-defocus is a classic technique proposed to derive depth from a single image that involves analysis of the blur inherent in a scene with a narrow depth of field. A challenge to this method is distinguishing blurriness caused by the focal blur from blurriness inherent to the observed scene. In 2007, MIT's Computer Science and Artificial Intelligence Laboratory demonstrated replacing the traditional rounded aperture with a coded aperture to produce a complex blur pattern that is more easily distinguished from the scene. A key to measuring depth using a coded aperture then is to correctly match the blur pattern in a region of the scene with a previously determined set of blur patterns for known depths. As the depth increases from the focal plane of the camera, the observable change in the blur pattern for small changes in depth is generally reduced. Consequently, as the depth of a feature to be measured using a depth-from-defocus technique increases, the measurement performance decreases. However, a Fresnel zone

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

    International Nuclear Information System (INIS)

    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

  1. 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

  2. Implementation of the Resonance Analysis Code SAMMY

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

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

  3. Automated Analysis of Source Code Patches using Machine Learning Algorithms

    OpenAIRE

    Castro Lechtaler, Antonio; Liporace, Julio César; Cipriano, Marcelo; García, Edith; Maiorano, Ariel; Malvacio, Eduardo; Tapia, Néstor

    2015-01-01

    An updated version of a tool for automated analysis of source code patches and branch differences is presented. The upgrade involves the use of machine learning techniques on source code, comments, and messages. It aims to help analysts, code reviewers, or auditors perform repetitive tasks continuously. The environment designed encourages collaborative work. It systematizes certain tasks pertaining to reviewing or auditing processes. Currently, the scope of the automated test is limited. C...

  4. AVS 3D Video Coding Technology and System

    Institute of Scientific and Technical Information of China (English)

    Siwei Ma; Shiqi Wang; Wen Gao

    2012-01-01

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

  5. 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.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-11-15

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

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

    International Nuclear Information System (INIS)

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

  9. 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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  12. Pipe whip analysis using the Tedel code

    International Nuclear Information System (INIS)

    In view of their abundance, piping systems are one of the main components in power industries and in particular in nuclear power plants. They must be designed for normal as well as faulted conditions, for safety requirements. For example, in case of a sudden break, the pipe whip must be studied in order to determine if the free pipe may damage neighbouring structures like other pipes, concrete containments, etc... The prediction of the dynamic behaviour of the free pipe requires accounting for several non-linearities. For this purpose, a beam type finite element program (TEDEL) has been used. The aim of this paper is to outline the main features of this program, when applied to pipe whip analysis. An example of application to a real case (the behaviour of PWR primary piping under LOCA conditions) will also be presented. (author)

  13. Improved Flow Modeling in Transient Reactor Safety Analysis Computer Codes

    International Nuclear Information System (INIS)

    A method of accounting for fluid-to-fluid shear in between calculational cells over a wide range of flow conditions envisioned in reactor safety studies has been developed such that it may be easily implemented into a computer code such as COBRA-TF for more detailed subchannel analysis. At a given nodal height in the calculational model, equivalent hydraulic diameters are determined for each specific calculational cell using either laminar or turbulent velocity profiles. The velocity profile may be determined from a separate CFD (Computational Fluid Dynamics) analysis, experimental data, or existing semi-empirical relationships. The equivalent hydraulic diameter is then applied to the wall drag force calculation so as to determine the appropriate equivalent fluid-to-fluid shear caused by the wall for each cell based on the input velocity profile. This means of assigning the shear to a specific cell is independent of the actual wetted perimeter and flow area for the calculational cell. The use of this equivalent hydraulic diameter for each cell within a calculational subchannel results in a representative velocity profile which can further increase the accuracy and detail of heat transfer and fluid flow modeling within the subchannel when utilizing a thermal hydraulics systems analysis computer code such as COBRA-TF. Utilizing COBRA-TF with the flow modeling enhancement results in increased accuracy for a coarse-mesh model without the significantly greater computational and time requirements of a full-scale 3D (three-dimensional) transient CFD calculation. (authors)

  14. 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.

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

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

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

  16. Fast reactor nuclear physics parameters calculation code system 'EXPARAM'

    International Nuclear Information System (INIS)

    The calculation code system ''EXPARAM'' was designed to analyze the experimental results systematically measured at the fast critical assembly (FCA) in Tokai research establishment of JAERI. Some calculation codes developed independently in JAERI and in US research institutes were collected and arranged as the fast reactor physics calculation code system. The multi-group core calculation code and the perturbation calculation code based on the diffusion theory and transport theory calculate the physics parameters such as eigenvalue, reaction rate, Doppler reactivity worth and sodium void worth. The dynamic physics parameters such as prompt neutron lifetime and effective delayed neutron fraction are also calculated. Input and Output data of calculation codes are transferred to each other using a direct access file on UNIX computer system. (author)

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

    OpenAIRE

    Jia-Shing Sheu; Kai-Chung Teng

    2013-01-01

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

  18. Digital system detects binary code patterns containing errors

    Science.gov (United States)

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

    1966-01-01

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

  19. Throughput of Coded Optical CDMA Systems with AND Detectors

    Science.gov (United States)

    Memon, Kehkashan A.; Umrani, Fahim A.; Umrani, A. W.; Umrani, Naveed A.

    2012-09-01

    Conventional detection techniques used in optical code-division multiple access (OCDMA) systems are not optimal and result in poor bit error rate performance. This paper analyzes the coded performance of optical CDMA systems with AND detectors for enhanced throughput efficiencies and improved error rate performance. The results show that the use of AND detectors significantly improve the performance of an optical channel.

  20. Development of a qualified nodalization for small-break LOCA transient analysis in PSB-VVER integral test facility by RELAP5 system code

    International Nuclear Information System (INIS)

    This paper deals with development and qualification of a nodalization for modeling of the PSB-VVER integral test facility (ITF) by RELAP5/MOD3.2 code and prediction of its primary and secondary systems behaviors at steady state and transient conditions. The PSB-VVER is a full-height, 1/300 volume and power scale representation of a VVER-1000 NPP. A RELAP5 nodalization has been developed for PSB-VVER modeling and a nodalization qualification process has been applied for the developed nodalization at steady state and transient levels and a qualified nodalization has been proposed for modeling of the PSB ITF. The 11% small-break loss-of-coolant-accident (SBLOCA), i.e. rupture of one of the hydroaccumulators (HA) injection lines in the upper plenum (UP) region of reactor pressure vessel (RPV) below the hot legs (HL), inlets has been considered for nodalization qualification process. The influence of the different steam generator (SG) nodalizations on the RELAP5 results and on the nodalization qualification process has been examined. The 'steady state' qualification level includes checking the correctness of the initial and boundary conditions and geometrical fidelity. In the 'transient' qualification level, the time dependent results of the code calculation are compared with the experimental time trends from both the qualitative and quantitative point of view. For quantitative assessment of the results, a Fast Fourier Transform Based Method (FFTBM) has been used. The FFTBM was used to establish a range in which the steam generators nodalizations can vary.

  1. Recent transportation calculation code systems and their accuracy evaluation

    International Nuclear Information System (INIS)

    In the field of shielding design, many studies have been carried out for the development of radiation transportation calculation codes (transportation codes) including Monte Carlo codes. The present report outlines major transportation codes used in Japan for design of shielding. Major one-dimensional codes include ANISN (Sn), PALLAS-PL and SP-Br (direct integration) whili two-dimensional ones include DOT-3.5 and TWOTRAN-II. All these transportation codes have been developed on the basis of numerical solution to the Boltzmann's transportation equation. These codes are roughly divided into two groups: discrete ordinates type and Monte Carlo type. The former include Sn-type codes and direct integration type codes. Sn-type codes are currently used most widely. The accuracy and other features of a code should be tested before applysing it to practical shielding design. One of the techniques for this purpose is the benchmark method, which consists of benchmark tests and analysis of the test results. The possible overall error involved in calculations can be determined from the benchmark tests. (Nogami, K.)

  2. Cooperative Coding Using Cyclic Delay Diversity for OFDM Systems

    Science.gov (United States)

    Lee, Dongwoo; Jung, Young Seok; Lee, Jae Hong

    This paper proposes cooperative coding using cyclic delay diversity (CDD) for OFDM systems. The cooperative diversity is combined with channel coding while CDD is applied to the cooperative transmission of the multiple relays to improve the beneficial effects of the cooperating relays. Analyses of frame error probability (FEP) and the average channel power of the proposed scheme are shown. Simulation results show the frame error rate (FER) of the proposed scheme. The proposed scheme provides not only a simple code design and low system complexity compared to conventional space-time processing, but better FER and diversity gain compared to direct transmission and conventional cooperative coding without CDD.

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

    Energy Technology Data Exchange (ETDEWEB)

    Mulliken, J.S.

    1995-12-01

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

  4. Simulation of some plant transients by the coupled code system ATHLET/KIKO3D

    International Nuclear Information System (INIS)

    The assessment of coupled reactor physics and thermal-hydraulic computation with the KIKO3D-ATHLET code is provided. The details of data flow in the coupling are reviewed and some selected results of the validation are described. The validated coupled system code is used in the safety analysis for WWER reactor. (Authors)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-08-01

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

  7. Production of analysis code for 'JOYO' dosimetry experiment

    International Nuclear Information System (INIS)

    As part of the measurement and analysis plan for the Dosimetry Experiment at the ''JOYO'' experimental fast reactor, neutron flux spectra analysis is performed using the NEUPAC (Neutron Unfolding Code Package) computer program. The code calculates the neutron flux spectra and other integral quantities from the activation data of the dosimeter foils. The NEUPAC code is based on the J1-type unfolding method, and the estimated neutron flux spectra is obtained as its solution. The program is able to determine the integral quantities and their sensitivities, together with an error estimate of the unfolded spectra and integral quantities. The code also performs a chi-square test of the input/output data, and contains many options for the calculational routines. This report presents the analytic theory, the program algorithms, and a description of the functions and use of the NEUPAC code. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    1978-12-04

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

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

    OpenAIRE

    Probstner, R.; J. Turan

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    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

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

    Institute of Scientific and Technical Information of China (English)

    CHENG Yuxin; ZHANG Lei; YI Na; XIANG Haige

    2007-01-01

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

  12. Performance Analysis of 3-Dimensional Turbo Codes

    CERN Document Server

    Rosnes, Eirik

    2011-01-01

    In this work, we consider the minimum distance properties and convergence thresholds of 3-dimensional turbo codes (3D-TCs), recently introduced by Berrou et al.. Here, we consider binary 3D-TCs while the original work of Berrou et al. considered double-binary codes. In the first part of the paper, the minimum distance properties are analyzed from an ensemble perspective, both in the finite-length regime and in the asymptotic case of large block lengths. In particular, we analyze the asymptotic weight distribution of 3D-TCs and show numerically that their typical minimum distance dmin may, depending on the specific parameters, asymptotically grow linearly with the block length, i.e., the 3D-TC ensemble is asymptotically good for some parameters. In the second part of the paper, we derive some useful upper bounds on the dmin when using quadratic permutation polynomial (QPP) interleavers with a quadratic inverse. Furthermore, we give examples of interleaver lengths where an upper bound appears to be tight. The b...

  13. Global sensitivity analysis of the XUV-ABLATOR code

    Science.gov (United States)

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

    2013-05-01

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

  14. Analysis code for gamma spectra detected with Ge detectors

    International Nuclear Information System (INIS)

    The Atomic Energy Center of Miyagi Prefecture introduced a Ge detector γ-ray spectrometer system for the purpose of examining environmental radiation around Onagawa Nuclear Power Station in March, 1981, and also developed an automatic spectrum analyzing code ARACC (automatic radioactivity calculation code) for qualitatively and quantitatively determining nuclides from the data obtained. After that, the data processing method and output format were improved. The program includes 17 sub-programs under the main program ARACC, in addition, there are the programs for preparing tables, calibrating the energy channels and calculating the efficiency. The Ge system of the Center can obtain normal, anti-compton and coincident spectrum data in a single measurement, but the ARACC is provided for analyzing only the normal spectra among these. In this report, not only the start of program, parameter input, printing output and magnetic tape output, but also the conversion of spectrum data format, peak searching, peak area determination, the treatment of disturbing peaks and the limit of detection are described as the methods of data analysis. The limit of detection is defined as such quantity that the significance level of concluding essentially existing radioactivity as non-existent as a result of measurement is lower than 5 %. (Wakatsuki, Y.)

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

    Directory of Open Access Journals (Sweden)

    Jia-Shing Sheu

    2013-04-01

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

  16. Developments of fuel performance analysis codes in KEPCO NF

    Energy Technology Data Exchange (ETDEWEB)

    Han, H. T.; Choi, J. M.; Jung, C. D.; Yoo, J. S. [KEPCO Nuclear Fuel Co. Ltd, Daejeon (Korea, Republic of)

    2012-03-15

    The KEPCO NF has developed fuel performance analysis and design code named as ROPER, and utility codes of XGCOL and XDNB in order to perform fuel rod design evaluation for Korean nuclear power plants. The ROPER code intends to cover full range of fuel performance evaluation. The XGCOL code is for the clad flattening evaluation and the XDNB code is for the extensive DNB propagation evaluation. In addition to these, the KEPCO NF is now in the developing stage for 3-dimensional fuel performance analysis code, named as OPER3D, using 3-dimensional FEM for the nest generation within the joint project CANDU ENERGY in order to analyze PCMI behavior and fuel performance under load following operation. Of these, the ROPER code is now in the stage of licensing activities by Korean regulatory body and the other two are almost in the final developing stage. After finishing the developing, licensing activities are to be performed. These activities are intending to acquire competitiveness, originality, vendor-free ownership of fuel performance codes in the KEPCO NF.

  17. Numerical diffusion and the tracking of solute fields in system codes. Part III. Application to a boron dilution transient analysis in the AP600

    International Nuclear Information System (INIS)

    For pt.II see ibid., vol.179, p.321-44, 1998. A study of a pump restart scenario in the AP600 with an unborated coolant plug in two of the four cold legs is presented. It has been performed with TRAC-PF1/MOD2 coupled with a 3-dimensional core neutronics model based on the nodal expansion method (NEM), and high order boron tracking algorithms. These are based on ULTIMATE-QUICKEST for 1-dimensional components and a flux corrected method developed by Smolarckievicz in the 3-dimensional vessel in order to reduce the numerical diffusion inherent to the upwind method used by most system codes to solve the transport equations. No turbulent diffusion model was included in the calculation to produce more conservative results. The results show that reduction of the numerical diffusion yields predictions with a significantly reduced margin in the size of the unborated plugs allowed to form in the primary side piping. In addition, two pump restart strategies have been suggested by the results, which could substantially decrease the size of an unborated plug injected into the core, in case it was suspected to have formed in a primary loop. (orig.)

  18. 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...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-08-01

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

  20. 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.

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

    International Nuclear Information System (INIS)

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

  2. Development of computer code CRIFLAN for critical flow analysis in reactor safety

    International Nuclear Information System (INIS)

    The computer code Critical Flow Analysis (CRIFLAN) has been developed for critical flow analysis in reactor safety. The code employs a heuristic model to describe critical discharges of water for both sub cooled and saturated conditions. Unlike many large thermal-hydraulic codes (RELAP5/MOD3, ATHLETE, and ASTEC) currently in use, CRIFLAN is a stand alone code and is not system specific. The proposed CRIFLAN model is validated against correlation based models such as H-F, Moody, and HEM and against the published experimental data. The CRIFLAN predictions are in good agreement with H-F results. It is concluded that the CRIFLAN code offers a reliable, user-friendly analytical tool for critical flow analysis in the field of reactor safety. (author)

  3. TRANSURANUS: A fuel rod analysis code ready for use

    International Nuclear Information System (INIS)

    The basic concepts of fuel rod performance codes are discussed. The TRANSURANUS code developed at the Institute for Transuranium Elements, Karlsruhe (GE) is presented. It is a quasi two-dimensional (11/2-D) code designed for treatment of a whole fuel rod for any type of reactor and any situation. The fuel rods found in the majority of test- or power reactors can be analyzed for very different situations (normal, off-normal and accidental). The time scale of the problems to be treated may range from milliseconds to years. The TRANSURANUS code consists of a clearly defined mechanical/mathematical framework into which physical models can easily be incorporated. This framework has been extensively tested and the programming very clearly reflects this structure. The code is well structured and easy to understand. It has a comprehensive material data bank for different fuels, claddings, coolants and their properties. The code can be employed in a deterministic and a statistical version. It is written in standard FORTRAN 77. The code system includes: 2 preprocessor programs (MAKROH and AXORDER) for setting up new data cases; the post-processor URPLOT for plotting all important quantities as a function of the radius, the axial coordinate or the time; the post-processor URSTART evaluating statistical analyses. The TRANSURANUS code exhibits short running times. A new WINDOWS-based interactive interface is under development. The code is now in use in various European institutions and is available to all interested parties. 7 figs., 15 refs

  4. Arithmetic coding as a non-linear dynamical system

    Science.gov (United States)

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

    2009-04-01

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

  5. Architecture of the ETR [experimental test reactor] systems code

    International Nuclear Information System (INIS)

    TETRA, a tokamak systems code capable of modeling experimental test reactors (ETRs), was developed in a joint effort by participants of the fusion community. The first version of this code was constructed to model devices similar to the Tokamak Ignition/Burn Engineering Reactor (TIBER) in configuration and design. A major feature of this code is its ability to perform optimization studies. Future work will include broadening the scope of the code, particularly in the area of materials selection, to more accurately simulate tokamak configurations such as the Next European Torus (NET) and the Fusion Engineering Reactor (FER). 18 refs., 2 figs., 4 tabs

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

    International Nuclear Information System (INIS)

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

  7. Benchmarking Of Improved DPAC Transient Deflagration Analysis Code

    Energy Technology Data Exchange (ETDEWEB)

    Laurinat, James E.; Hensel, Steve J.

    2013-03-21

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

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

    International Nuclear Information System (INIS)

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

  9. A program to validate computer codes for container impact analysis

    International Nuclear Information System (INIS)

    The detailed analysis of containers during impacts to assess either margins to failure or the consequences of different design strategies, requires the use of sophisticated computer codes to model the interactions of the various structural components. The combination of plastic deformation, impact and sliding at interfaces and dynamic loading effects provides a severe test of both the skill of the analyst and the robustness of the computer codes. A program of experiments has been under way at Winfrith since 1987 using extensively instrumented models to provide data for the validation of such codes. Three finite element codes, DYNA3D, HONDO-II and ABAQUS, were selected as suitable tools to cover the range of conditions expected in typical impacts. The impact orientation, velocity and instrumentation locations for the experiments are specified by pre-test calculations using these codes. Post-test analyses using the actual impact orientation and velocities are carried out as necessary if significant discrepancies are found

  10. Verification of Sensitivity and Uncertainty Analysis Code with the GODIVA and VHTR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Han, Tae Young; Park, Ho Jin; Lee, Hyun Chul; Jo, Chang Keun; Noh, Jae Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    Korea Atomic Energy Research Institute has designed a UO{sub 2} fueled 200 MWth prismatic very high temperature reactor (PMR200) and DeCART/CAPP code systems with two-step procedure have been established for the core analysis. However, it is necessary to analyze the sensitivity and uncertainty analysis for the nuclear design and safety analysis of the new core and the code validation. For this reason, we are developing a sensitivity and uncertainty analysis code, MUSAD (Modules of Uncertainty and Sensitivity Analysis for DeCART), which has the specific functions of producing the sensitivity coefficients and uncertainty quantification to the system multiplication factor (k{sub eff}) using the covariance data and the group constants. In this paper, the methodology of the sensitivity and uncertainty analysis code was described and the calculation results on the GODIVA benchmark and the PMR200 pin cell problem were compared with the results by TSUNAMI of SCALE 6.1 and McCARD. In this paper, the methodology for the sensitivity and uncertainty analysis code, MUSAD, was described and the verification calculations on the GODIVA benchmark and the PMR200 pin cell problem were carried out. As a result, they are in a good agreement when compared with the results by TSUNAMI and McCARD. From this study, it is expected that MUSAD code can produce accurate results, if the cross sections and the core parameter for the S and U analysis could be completely obtained from DeCART.

  11. Verification of Sensitivity and Uncertainty Analysis Code with the GODIVA and VHTR fuel

    International Nuclear Information System (INIS)

    Korea Atomic Energy Research Institute has designed a UO2 fueled 200 MWth prismatic very high temperature reactor (PMR200) and DeCART/CAPP code systems with two-step procedure have been established for the core analysis. However, it is necessary to analyze the sensitivity and uncertainty analysis for the nuclear design and safety analysis of the new core and the code validation. For this reason, we are developing a sensitivity and uncertainty analysis code, MUSAD (Modules of Uncertainty and Sensitivity Analysis for DeCART), which has the specific functions of producing the sensitivity coefficients and uncertainty quantification to the system multiplication factor (keff) using the covariance data and the group constants. In this paper, the methodology of the sensitivity and uncertainty analysis code was described and the calculation results on the GODIVA benchmark and the PMR200 pin cell problem were compared with the results by TSUNAMI of SCALE 6.1 and McCARD. In this paper, the methodology for the sensitivity and uncertainty analysis code, MUSAD, was described and the verification calculations on the GODIVA benchmark and the PMR200 pin cell problem were carried out. As a result, they are in a good agreement when compared with the results by TSUNAMI and McCARD. From this study, it is expected that MUSAD code can produce accurate results, if the cross sections and the core parameter for the S and U analysis could be completely obtained from DeCART

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

    International Nuclear Information System (INIS)

    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)

  13. Modern Nuclear Data Evaluation with the TALYS Code System

    International Nuclear Information System (INIS)

    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.

  14. MORSE Monte Carlo radiation transport code system

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  17. Characteristic Analysis of Fire Modeling Codes

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yoon Hwan; Yang, Joon Eon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Jong Hoon [Kyeongmin College, Ujeongbu (Korea, Republic of)

    2004-04-15

    This report documents and compares key features of four zone models: CFAST, COMPBRN IIIE, MAGIC and the Fire Induced Vulnerability Evaluation (FIVE) methodology. CFAST and MAGIC handle multi-compartment, multi-fire problems, using many equations; COMPBRN and FIVE handle single compartment, single fire source problems, using simpler equation. The increased rigor of the formulation of CFAST and MAGIC does not mean that these codes are more accurate in every domain; for instance, the FIVE methodology uses a single zone approximation with a plume/ceiling jet sublayer, while the other models use a two-zone treatment without a plume/ceiling jet sublayer. Comparisons with enclosure fire data indicate that inclusion of plume/ceiling jet sublayer temperatures is more conservative, and generally more accurate than neglecting them. Adding a plume/ceiling jet sublayer to the two-zone models should be relatively straightforward, but it has not been done yet for any of the two-zone models. Such an improvement is in progress for MAGIC.

  18. Development of GUI systems for the MIDAS code

    International Nuclear Information System (INIS)

    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)

  19. Software Security Analysis : Managing source code audit

    OpenAIRE

    Persson, Daniel; Baca, Dejan

    2004-01-01

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

  20. Digital Image Analysis for Detechip Code Determination

    OpenAIRE

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

    2012-01-01

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