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Sample records for system analysis code

  1. SASSYS LMFBR systems analysis code

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  2. Code system for fast reactor neutronics analysis

    International Nuclear Information System (INIS)

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

    1983-04-01

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

  3. Systemization of burnup sensitivity analysis code

    International Nuclear Information System (INIS)

    Tatsumi, Masahiro; Hyoudou, Hideaki

    2004-02-01

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

  4. Systemization of burnup sensitivity analysis code. 2

    International Nuclear Information System (INIS)

    Tatsumi, Masahiro; Hyoudou, Hideaki

    2005-02-01

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

  5. Use of computer codes for system reliability analysis

    International Nuclear Information System (INIS)

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

    1988-01-01

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

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

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

    International Nuclear Information System (INIS)

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

    2011-03-01

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

  8. 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)

  9. Use of computer codes for system reliability analysis

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  10. Use of computer codes for system reliability analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    1989-01-01

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

  11. Development of realistic thermal hydraulic system analysis code

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-05-01

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

  12. Development of realistic thermal hydraulic system analysis code

    International Nuclear Information System (INIS)

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

    2002-05-01

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

  13. Development of the integrated system reliability analysis code MODULE

    International Nuclear Information System (INIS)

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

    1987-01-01

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

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

    International Nuclear Information System (INIS)

    Sakai, Takaaki

    2002-11-01

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  17. Development of a nuclear power plant system analysis code

    International Nuclear Information System (INIS)

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

    1997-07-01

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

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

    International Nuclear Information System (INIS)

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

    2015-07-01

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

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

    International Nuclear Information System (INIS)

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

    1985-03-01

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

  20. 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)

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-02-01

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

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

    Science.gov (United States)

    2010-12-01

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

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

    International Nuclear Information System (INIS)

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

    1997-09-01

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

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

    International Nuclear Information System (INIS)

    Tatsumi, Masahiro; Hyoudou, Hideaki

    2008-08-01

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

  6. THYDE-NEU: Nuclear reactor system analysis code

    International Nuclear Information System (INIS)

    Asahi, Yoshiro

    2002-03-01

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

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

    International Nuclear Information System (INIS)

    Honma, Toshimitsu; Sasahara, Takashi.

    1993-10-01

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

  8. 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.

  9. Experimental transport analysis code system in JT-60

    International Nuclear Information System (INIS)

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

    1988-03-01

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

  10. 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.

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    International Nuclear Information System (INIS)

    Berry, G.; Geyer, H.

    1983-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-05-25

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

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

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1989-02-01

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

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

    International Nuclear Information System (INIS)

    Konstantin Mikityuk; Sandro Pelloni; Paul Coddington

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Asahi, Yoshiro; Hirano, Masashi; Sato, Kazuo

    1986-12-01

    THYDE-P2, being characterized by the new thermal-hydraulic network model, is applicable to analysis of RCS behaviors in response to various disturbances including LB (large break)-LOCA(loss-of-coolant accident). In LB-LOCA analysis, THYDE-P2 is capable of through calculation from its initiation to complete reflooding of the core without an artificial change in the methods and models. The first half of the report is the description of the methods and models for use in the THYDE-P2 code, i.e., (1) the thermal-hydraulic network model, (2) the various RCS components models, (3) the heat sources in fuel, (4) the heat transfer correlations, (5) the mechanical behavior of clad and fuel, and (6) the steady state adjustment. The second half of the report is the user's mannual for the THYDE-P2 code (version SV04L08A) containing items; (1) the program control (2) the input requirements, (3) the execution of THYDE-P2 job, (4) the output specifications and (5) the sample problem to demonstrate capability of the thermal-hydraulic network model, among other things. (author)

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

    International Nuclear Information System (INIS)

    Shi, Chengbin; Cheng, Maosong; Liu, Guimin

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-01-01

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

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  3. 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.

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

    International Nuclear Information System (INIS)

    2006-03-01

    In order to develop the system to affect more advanced and rational regulations of nuclear fuels and earlier introduction of new technologies in nuclear power plants, domestic and overseas safety regulation systems and state of their implementation for water cooled reactor fuel and safety analysis code had been investigated and new regulation system to affect nuclear fuel type authorization and analysis code certification was proposed. Topical report system for common parts related with nuclear fuel type authorization and analysis code certification was firstly proposed for knowledge base. Maintaining consistent safety examination supported by experts, introduction of domestic efficient system for lead irradiation test fuel, and analysis code certification and quality assurance were also proposed. (T. Tanaka)

  5. SASSYS LMFBR systems code

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  6. An Adaptation of the HELIOS/MASTER Code System to the Analysis of VHTR Cores

    International Nuclear Information System (INIS)

    Noh, Jae Man; Lee, Hyun Chul; Kim, Kang Seog; Kim, Yong Hee

    2006-01-01

    KAERI is developing a new computer code system for an analysis of VHTR cores based on the existing HELIOS/MASTER code system which was originally developed for a LWR core analysis. In the VHTR reactor physics, there are several unique neutronic characteristics that cannot be handled easily by the conventional computer code system applied for the LWR core analysis. Typical examples of such characteristics are a double heterogeneity problem due to the particulate fuels, the effects of a spectrum shift and a thermal up-scattering due to the graphite moderator, and a strong fuel/reflector interaction, etc. In order to facilitate an easy treatment of such characteristics, we developed some methodologies for the HELIOS/MASTER code system and tested their applicability to the VHTR core analysis

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

    International Nuclear Information System (INIS)

    Tsuchihashi, Keichiro

    1985-01-01

    SRAC, JAERI thermal reactor standard code system for reactor design and analysis, developed in Japan Atomic Energy Research Institute, is for all types of thermal neutron nuclear design and analysis. The code system has undergone extensive verifications to confirm its functions, and has been used in core modification of the research reactor, detailed design of the multi-purpose high temperature gas reactor and analysis of the experiment with a critical assembly. In nuclear calculation with the code system, multi-group lattice calculation is first made with the libraries. Then, with the resultant homogeneous equivalent group constants, reactor core calculation is made. Described are the following: purpose and development of the code system, functions of the SRAC system, bench mark tests and usage state and future development. (Mori, K.)

  8. 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

  9. 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.

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

    International Nuclear Information System (INIS)

    Hwang, Dae Hyun; Seo, K. W.

    2006-01-01

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

  11. Perspectives on the development of next generation reactor systems safety analysis codes

    International Nuclear Information System (INIS)

    Zhang, H.

    2015-01-01

    'Full text:' Existing reactor system analysis codes, such as RELAP5-3D and TRAC, have gained worldwide success in supporting reactor safety analyses, as well as design and licensing of new reactors. These codes are important assets to the nuclear engineering research community, as well as to the nuclear industry. However, most of these codes were originally developed during the 1970s', and it becomes necessary to develop next-generation reactor system analysis codes for several reasons. Firstly, as new reactor designs emerge, there are new challenges emerging in numerical simulations of reactor systems such as long lasting transients and multi-physics phenomena. These new requirements are beyond the range of applicability of the existing system analysis codes. Advanced modeling and numerical methods must be taken into consideration to improve the existing capabilities. Secondly, by developing next-generation reactor system analysis codes, the knowledge (know how) in two phase flow modeling and the highly complex constitutive models will be transferred to the young generation of nuclear engineers. And thirdly, all computer codes have limited shelf life. It becomes less and less cost-effective to maintain a legacy code, due to the fast change of computer hardware and software environment. There are several critical perspectives in terms of developing next-generation reactor system analysis codes: 1) The success of the next-generation codes must be built upon the success of the existing codes. The knowledge of the existing codes, not just simply the manuals and codes, but knowing why and how, must be transferred to the next-generation codes. The next-generation codes should encompass the capability of the existing codes. The shortcomings of existing codes should be identified, understood, and properly categorized, for example into model deficiencies or numerical method deficiencies. 2) State-of-the-art models and numerical methods must be considered to

  12. Perspectives on the development of next generation reactor systems safety analysis codes

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, H., E-mail: Hongbin.Zhang@inl.gov [Idaho National Laboratory, Idaho Falls, ID (United States)

    2015-07-01

    'Full text:' Existing reactor system analysis codes, such as RELAP5-3D and TRAC, have gained worldwide success in supporting reactor safety analyses, as well as design and licensing of new reactors. These codes are important assets to the nuclear engineering research community, as well as to the nuclear industry. However, most of these codes were originally developed during the 1970s', and it becomes necessary to develop next-generation reactor system analysis codes for several reasons. Firstly, as new reactor designs emerge, there are new challenges emerging in numerical simulations of reactor systems such as long lasting transients and multi-physics phenomena. These new requirements are beyond the range of applicability of the existing system analysis codes. Advanced modeling and numerical methods must be taken into consideration to improve the existing capabilities. Secondly, by developing next-generation reactor system analysis codes, the knowledge (know how) in two phase flow modeling and the highly complex constitutive models will be transferred to the young generation of nuclear engineers. And thirdly, all computer codes have limited shelf life. It becomes less and less cost-effective to maintain a legacy code, due to the fast change of computer hardware and software environment. There are several critical perspectives in terms of developing next-generation reactor system analysis codes: 1) The success of the next-generation codes must be built upon the success of the existing codes. The knowledge of the existing codes, not just simply the manuals and codes, but knowing why and how, must be transferred to the next-generation codes. The next-generation codes should encompass the capability of the existing codes. The shortcomings of existing codes should be identified, understood, and properly categorized, for example into model deficiencies or numerical method deficiencies. 2) State-of-the-art models and numerical methods must be considered to

  13. Restructuring of burnup sensitivity analysis code system by using an object-oriented design approach

    International Nuclear Information System (INIS)

    Kenji, Yokoyama; Makoto, Ishikawa; Masahiro, Tatsumi; Hideaki, Hyoudou

    2005-01-01

    A new burnup sensitivity analysis code system was developed with help from the object-oriented technique and written in Python language. It was confirmed that they are powerful to support complex numerical calculation procedure such as reactor burnup sensitivity analysis. The new burnup sensitivity analysis code system PSAGEP was restructured from a complicated old code system and reborn as a user-friendly code system which can calculate the sensitivity coefficients of the nuclear characteristics considering multicycle burnup effect based on the generalized perturbation theory (GPT). A new encapsulation framework for conventional codes written in Fortran was developed. This framework supported to restructure the software architecture of the old code system by hiding implementation details and allowed users of the new code system to easily calculate the burnup sensitivity coefficients. The framework can be applied to the other development projects since it is carefully designed to be independent from PSAGEP. Numerical results of the burnup sensitivity coefficient of a typical fast breeder reactor were given with components based on GPT and the multicycle burnup effects on the sensitivity coefficient were discussed. (authors)

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

    International Nuclear Information System (INIS)

    Neufert, A.; Van de Velde, A.

    1999-01-01

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

  15. Feasibility study on the rod ejection accident analysis with RETRAN-MASTER code system

    International Nuclear Information System (INIS)

    Kim, Y. H.; Lee, C. S.

    2003-01-01

    KEPRI has been developed the in-house methodology for non-LOCA safety analyses based on the codes and methodologies of vendors and EPRI. Using the methodology, the rod ejection accident, which is classified into the generic accident analysis category of reactivity insertion accident in primary system, has been analyzed with RETRAN-MASTER code system. And the feasibility of the coupled code system has been verified by the review of the results. Furthermore, to assess the important parameters to the accident, the sensitivity analyses have been carried out over some parameters

  16. PROMETHEUS - a code system for dynamic 3-D analysis of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Khotylev, V.A.; Hoogenboom, J.E.; Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.

    1996-09-01

    The paper presents a multidimensional, general-purpose neutronics code system. It solves a number of steady-state and/or transient problems with coupled thermal hydraulics in one-, two-, or three-dimensional geometry. Due to a number of specialized features such as cavity treatment, automated convergence control, burnup treatment using the full isotopic transition matrix, the code system can be applied for the analysis of fast and slow transients in small, large, and innovative reactor cores. (author)

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

    International Nuclear Information System (INIS)

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

    2007-07-01

    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

  18. Revised SRAC code system

    International Nuclear Information System (INIS)

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

    1986-09-01

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

  19. Coupling of RELAP5-3D and GAMMA codes for Nuclear Hydrogen System Analysis

    International Nuclear Information System (INIS)

    Jin, Hyung Gon

    2007-02-01

    RELAP5-3D is one of the most important system analysis codes in nuclear field, which has been developed for best-estimate transient simulation of light water reactor coolant systems during postulated accidents. The GAMMA code is a multi-dimensional multi-component mixture analysis code with the complete set of chemical reaction models which is developed for safety analysis of HTGR (High Temperature Gas Cooled Reactor) air-ingress. The two codes, RELAP5-3D and GAMMA, are coupled to be used for nuclear-hydrogen system analysis, which requires the capability of the analysis of multi-component gas mixture and two-phase flow. In order to couple the two codes, 4 steps are needed. Before coupling, the GAMMA code was transformed into DLL (dynamic link liberally) from executive type and RELAP5-3D was recompiled into Compaq Visual Fortran environments for our debugging purpose. As the second step, two programs - RELAP5-3D and GAMMA codes - must be synchronized in terms of time and time step. Based on that time coupling, the coupled code can calculate simultaneously. Time-step coupling had been accomplished successfully and it is tested by using a simple test input. As a next step, source-term coupling was done and it was also tested in two different test inputs. The fist case is a simple test condition, which has no chemical reaction. And the other test set is the chemical reaction model, including four non-condensable gas species, which are He, O2, CO, CO2. Finally, in order to analyze combined cycle system, heat-flux coupling has been made and a simple heat exchanger model was demonstrated

  20. Development of platform to compare different wall heat transfer packages for system analysis codes

    International Nuclear Information System (INIS)

    Kim, Min-Gil; Lee, Won Woong; Lee, Jeong Ik; Shin, Sung Gil

    2016-01-01

    System thermal hydraulic (STH) analysis code is used for analyzing and evaluating the safety of a designed nuclear system. The system thermal hydraulic analysis code typically solves mass, momentum and energy conservation equations for multiple phases with sets of selected empirical constitutive equations to close the problem. Several STH codes are utilized in academia, industry and regulators, such as MARS-KS, SPACE, RELAP5, COBRA-TF, TRACE, and so on. Each system thermal hydraulic code consists of different sets of governing equations and correlations. However, the packages and sets of correlations of each code are not compared quantitatively yet. Wall heat transfer mode transition maps of SPACE and MARS-KS have a little difference for the transition from wall nucleate heat transfer mode to wall film heat transfer mode. Both codes have the same heat transfer packages and correlations in most region except for wall film heat transfer mode. Most of heat transfer coefficients calculated for the range of selected variables of SPACE are the same with those of MARS-KS. For the intervals between 500K and 540K of wall temperature, MARS-KS selects the wall film heat transfer mode and Bromley correlation but SPACE select the wall nucleate heat transfer mode and Chen correlation. This is because the transition from nucleate boiling to film boiling of MARS-KS is earlier than SPACE. More detailed analysis of the heat transfer package and flow regime package will be followed in the near future

  1. Verification and implications of the multiple pin treatment in the SASSYS-1 LMR systems analysis code

    International Nuclear Information System (INIS)

    Dunn, F.E.

    1994-01-01

    As part of a program to obtain realistic, as opposed to excessively conservative, analysis of reactor transients, a multiple pin treatment for the analysis of intra-subassembly thermal hydraulics has been included in the SASSYS-1 liquid metal reactor systems analysis code. This new treatment has made possible a whole new level of verification for the code. The code can now predict the steady-state and transient responses of individual thermocouples within instrumented subassemlies in a reactor, rather than just predicting average temperatures for a subassembly. Very good agreement has been achieved between code predictions and the experimental measurements of steady-state and transient temperatures and flow rates in the Shutdown Heat Removal Tests in the EBR-II Reactor. Detailed multiple pin calculations for blanket subassemblies in the EBR-II reactor demonstrate that the actual steady-state and transient peak temperatures in these subassemblies are significantly lower than those that would be calculated by simpler models

  2. 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.

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

    International Nuclear Information System (INIS)

    Myong, Hyon Kook; Kim, Jong Eun; Ahn, Jong Ki; Yang, Seung Yong

    2007-03-01

    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

  4. SAMPO: a code system giving different orders of approximation for sensitivity and perturbation analysis

    International Nuclear Information System (INIS)

    Estiot, J.C.; Salvatores, M.; Palmiotti, G.

    1981-01-01

    We present the characteristics of SAMPO, a one dimension transport theory code system, which is used for the following types of calculation: sensitivity analysis for functional linear or bi-linear on the direct or adjoint flux and their ratios; classic perturbation analysis. First order calculations, as well higher order, can be presented

  5. Improvement of multi-dimensional realistic thermal-hydraulic system analysis code, MARS 1.3

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Chung, Bub Dong; Jeong, Jae Jun; Ha, Kwi Seok

    1998-09-01

    The MARS (Multi-dimensional Analysis of Reactor Safety) code is a multi-dimensional, best-estimate thermal-hydraulic system analysis code. This report describes the new features that have been improved in the MARS 1.3 code since the release of MARS 1.3 in July 1998. The new features include: - implementation of point kinetics model into the 3D module - unification of the heat structure model - extension of the control function to the 3D module variables - improvement of the 3D module input check function. Each of the items has been implemented in the developmental version of the MARS 1.3.1 code and, then, independently verified and assessed. The effectiveness of the new features is well verified and it is shown that these improvements greatly extend the code capability and enhance the user friendliness. Relevant input data changes are also described. In addition to the improvements, this report briefly summarizes the future code developmental activities that are being carried out or planned, such as coupling of MARS 1.3 with the containment code CONTEMPT and the three-dimensional reactor kinetics code MASTER 2.0. (author). 8 refs.

  6. Improvement of multi-dimensional realistic thermal-hydraulic system analysis code, MARS 1.3

    International Nuclear Information System (INIS)

    Lee, Won Jae; Chung, Bub Dong; Jeong, Jae Jun; Ha, Kwi Seok

    1998-09-01

    The MARS (Multi-dimensional Analysis of Reactor Safety) code is a multi-dimensional, best-estimate thermal-hydraulic system analysis code. This report describes the new features that have been improved in the MARS 1.3 code since the release of MARS 1.3 in July 1998. The new features include: - implementation of point kinetics model into the 3D module - unification of the heat structure model - extension of the control function to the 3D module variables - improvement of the 3D module input check function. Each of the items has been implemented in the developmental version of the MARS 1.3.1 code and, then, independently verified and assessed. The effectiveness of the new features is well verified and it is shown that these improvements greatly extend the code capability and enhance the user friendliness. Relevant input data changes are also described. In addition to the improvements, this report briefly summarizes the future code developmental activities that are being carried out or planned, such as coupling of MARS 1.3 with the containment code CONTEMPT and the three-dimensional reactor kinetics code MASTER 2.0. (author). 8 refs

  7. 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.

  8. EVNTRE, Code System for Event Progression Analysis for PRA

    International Nuclear Information System (INIS)

    2002-01-01

    1 - Description of program or function: EVNTRE is a generalized event tree processor that was developed for use in probabilistic risk analysis of severe accident progressions for nuclear power plants. The general nature of EVNTRE makes it applicable to a wide variety of analyses that involve the investigation of a progression of events which lead to a large number of sets of conditions or scenarios. EVNTRE efficiently processes large, complex event trees. It can assign probabilities to event tree branch points in several different ways, classify pathways or outcomes into user-specified groupings, and sample input distributions of probabilities and parameters. PSTEVNT, a post-processor program used to sort and reclassify the 'binned' data output from EVNTRE and generate summary tables, is included. 2 - Methods: EVNTRE processes event trees that are cast in the form of questions or events, with multiple choice answers for each question. Split fractions (probabilities or frequencies that sum to unity) are either supplied or calculated for the branches of each question in a path-dependent manner. EVNTRE traverses the tree, enumerating the leaves of the tree and calculating their probabilities or frequencies based upon the initial probability or frequency and the split fractions for the branches taken along the corresponding path to an individual leaf. The questions in the event tree are usually grouped to address specific phases of time regimes in the progression of the scenario or severe accident. Grouping or binning of each path through the event tree in terms of a small number of characteristics or attributes is allowed. Boolean expressions of the branches taken are used to select the appropriate values of the characteristics of interest for the given path. Typically, the user specifies a cutoff tolerance for the frequency of a pathway to terminate further exploration. Multiple sets of input to an event tree can be processed by using Monte Carlo sampling to generate

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

    International Nuclear Information System (INIS)

    McCarville, T.J.; Meier, W.R.; Carson, C.F.; Glasgow, B.B.

    1987-01-01

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

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

    International Nuclear Information System (INIS)

    Yu, Keuk Jong; You, Sung Chang; Kim, Han Gon

    2009-01-01

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

  11. Evaluation of hydrogen production system coupling with HTTR using dynamic analysis code

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Ohashi, Hirofumi; Inaba, Yoshitomo; Nishihara, Tetsuo; Hayashi, Koji; Inagaki, Yoshiyuki

    2006-01-01

    The Japan Atomic Energy Agency (JAEA) was entrusted 'Development of Nuclear Heat Utilization Technology' by Ministry of Education, Culture, Sports, Science and Technology. In this development, the JAEA investigated the system integration technology to couple the hydrogen production system by steam reforming with the High Temperature Engineering Test Reactor (HTTR). Prior to the construction of the hydrogen production system coupling with the HTTR, a dynamic analysis code had to be developed to evaluate the system transient behaviour of the hydrogen production system because there are no examples of chemical facilities coupled with nuclear reactor in the world. This report describes the evaluation of the hydrogen production system coupling with HTTR using analysis code, N-HYPAC, which can estimate transient behaviour of the hydrogen production system by steam reforming. The results of this investigation provide that the influence of the thermal disturbance caused by the hydrogen production system on the HTTR can be estimated well. (author)

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

    International Nuclear Information System (INIS)

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

    2001-01-01

    This paper presents the results of system thermalhydraulic code using the sub-channel analysis approach in predicting the Neptun boil off experiments. This approach will be suitable for further works in view of coupling the system code with a 3D neutron kinetic one. The boil off tests were conducted in order to simulate the consequences of loss of coolant inventory leading to uncovery and heat up of fuel elements of a nuclear reactor core. In this framework, the Neptun low pressure test No5002, which is a good repeat experiment, is considered. The calculations were carried out using the system transient analysis code Relap5/Mod3.2. A detailed nodalization of the Neptun test section was developed. A reference case was run, and the overall data comparison shows good agreement between calculated and experimental thermalhydraulic parameters. A series of sensitivity analyses were also performed in order to assess the code prediction capabilities. The obtained results were almost satisfactory, this demonstrates, as well, the reasonable success of the subchannel analysis approach adopted in the present context for a system thermalhydraulic code.(author)

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

    International Nuclear Information System (INIS)

    In, W. K.; Yoo, Y. J.; Hwang, T. H.; Ji, S. K.

    2001-01-01

    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

  14. User's manual of SECOM2: a computer code for seismic system reliability analysis

    International Nuclear Information System (INIS)

    Uchiyama, Tomoaki; Oikawa, Tetsukuni; Kondo, Masaaki; Tamura, Kazuo

    2002-03-01

    This report is the user's manual of seismic system reliability analysis code SECOM2 (Seismic Core Melt Frequency Evaluation Code Ver.2) developed at the Japan Atomic Energy Research Institute for systems reliability analysis, which is one of the tasks of seismic probabilistic safety assessment (PSA) of nuclear power plants (NPPs). The SECOM2 code has many functions such as: Calculation of component failure probabilities based on the response factor method, Extraction of minimal cut sets (MCSs), Calculation of conditional system failure probabilities for given seismic motion levels at the site of an NPP, Calculation of accident sequence frequencies and the core damage frequency (CDF) with use of the seismic hazard curve, Importance analysis using various indicators, Uncertainty analysis, Calculation of the CDF taking into account the effect of the correlations of responses and capacities of components, and Efficient sensitivity analysis by changing parameters on responses and capacities of components. These analyses require the fault tree (FT) representing the occurrence condition of the system failures and core damage, information about response and capacity of components and seismic hazard curve for the NPP site as inputs. This report presents the models and methods applied in the SECOM2 code and how to use those functions. (author)

  15. Enhancement of safety analysis reliability for a CANDU-6 reactor using RELAP-CANDU/SCAN coupled code system

    International Nuclear Information System (INIS)

    Kim, Man Woong; Choi, Yong Seog; Sin, Chul; Kim, Hyun Koon; Kim, Hho Jung; Hwang, Su Hyun; Hong, In Seob; Kim, Chang Hyo

    2005-01-01

    In LOCA analysis of the CANDU reactor, the system thermal-hydraulic code, RELAP-CANDU, alone cannot predict the transient behavior accurately. Therefore, the best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. To perform on-line calculation of safety analysis for CANDU reactor, a coupled thermal hydraulics-neutronics code system was developed in such a way that the best-estimate thermal-hydraulic system code for CANDU reactor, RELAP-CANDU, is coupled with the full three-dimensional reactor core kinetic code

  16. Objective Oriented Design of Architecture for TH System Safety Analysis Code and Verification

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong

    2008-03-15

    In this work, objective oriented design of generic system analysis code has been tried based on the previous works in KAERI for two phase three field Pilot code. It has been performed to implement of input and output design, TH solver, component model, special TH models, heat structure solver, general table, trip and control, and on-line graphics. All essential features for system analysis has been designed and implemented in the final product SYSTF code. The computer language C was used for implementation in the Visual studio 2008 IDE (Integrated Development Environment) since it has easier and lighter than C++ feature. The code has simple and essential features of models and correlation, special component, special TH model and heat structure model. However the input features is able to simulate the various scenarios, such as steady state, non LOCA transient and LOCA accident. The structure validity has been tested through the various verification tests and it has been shown that the developed code can treat the non LOCA and LOCA simulation. However more detailed design and implementation of models are required to get the physical validity of SYSTF code simulation.

  17. Objective Oriented Design of Architecture for TH System Safety Analysis Code and Verification

    International Nuclear Information System (INIS)

    Chung, Bub Dong

    2008-03-01

    In this work, objective oriented design of generic system analysis code has been tried based on the previous works in KAERI for two phase three field Pilot code. It has been performed to implement of input and output design, TH solver, component model, special TH models, heat structure solver, general table, trip and control, and on-line graphics. All essential features for system analysis has been designed and implemented in the final product SYSTF code. The computer language C was used for implementation in the Visual studio 2008 IDE (Integrated Development Environment) since it has easier and lighter than C++ feature. The code has simple and essential features of models and correlation, special component, special TH model and heat structure model. However the input features is able to simulate the various scenarios, such as steady state, non LOCA transient and LOCA accident. The structure validity has been tested through the various verification tests and it has been shown that the developed code can treat the non LOCA and LOCA simulation. However more detailed design and implementation of models are required to get the physical validity of SYSTF code simulation

  18. FERRET data analysis code

    International Nuclear Information System (INIS)

    Schmittroth, F.

    1979-09-01

    A documentation of the FERRET data analysis code is given. The code provides a way to combine related measurements and calculations in a consistent evaluation. Basically a very general least-squares code, it is oriented towards problems frequently encountered in nuclear data and reactor physics. A strong emphasis is on the proper treatment of uncertainties and correlations and in providing quantitative uncertainty estimates. Documentation includes a review of the method, structure of the code, input formats, and examples

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

    International Nuclear Information System (INIS)

    Ahmed, H.; Cowie, A.; Malek, R.A.; Rafer, A.; Ma, D.; Tebo, F.

    1978-10-01

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    Kunz, R.F.; Kasmala, G.F.; Murray, C.J.; Mahaffy, J.H.

    2000-01-01

    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)

  2. Verification and validation of the THYTAN code for the graphite oxidation analysis in the HTGR systems

    International Nuclear Information System (INIS)

    Shimazaki, Yosuke; Isaka, Kazuyoshi; Nomoto, Yasunobu; Seki, Tomokazu; Ohashi, Hirofumi

    2014-12-01

    The analytical models for the evaluation of graphite oxidation were implemented into the THYTAN code, which employs the mass balance and a node-link computational scheme to evaluate tritium behavior in the High Temperature Gas-cooled Reactor (HTGR) systems for hydrogen production, to analyze the graphite oxidation during the air or water ingress accidents in the HTGR systems. This report describes the analytical models of the THYTAN code in terms of the graphite oxidation analysis and its verification and validation (V and V) results. Mass transfer from the gas mixture in the coolant channel to the graphite surface, diffusion in the graphite, graphite oxidation by air or water, chemical reaction and release from the primary circuit to the containment vessel by a safety valve were modeled to calculate the mass balance in the graphite and the gas mixture in the coolant channel. The computed solutions using the THYTAN code for simple questions were compared to the analytical results by a hand calculation to verify the algorithms for each implemented analytical model. A representation of the graphite oxidation experimental was analyzed using the THYTAN code, and the results were compared to the experimental data and the computed solutions using the GRACE code, which was used for the safety analysis of the High Temperature Engineering Test Reactor (HTTR), in regard to corrosion depth of graphite and oxygen concentration at the outlet of the test section to validate the analytical models of the THYTAN code. The comparison of THYTAN code results with the analytical solutions, experimental data and the GRACE code results showed the good agreement. (author)

  3. Architectural and Algorithmic Requirements for a Next-Generation System Analysis Code

    Energy Technology Data Exchange (ETDEWEB)

    V.A. Mousseau

    2010-05-01

    This document presents high-level architectural and system requirements for a next-generation system analysis code (NGSAC) to support reactor safety decision-making by plant operators and others, especially in the context of light water reactor plant life extension. The capabilities of NGSAC will be different from those of current-generation codes, not only because computers have evolved significantly in the generations since the current paradigm was first implemented, but because the decision-making processes that need the support of next-generation codes are very different from the decision-making processes that drove the licensing and design of the current fleet of commercial nuclear power reactors. The implications of these newer decision-making processes for NGSAC requirements are discussed, and resulting top-level goals for the NGSAC are formulated. From these goals, the general architectural and system requirements for the NGSAC are derived.

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

    International Nuclear Information System (INIS)

    Tsuchihashi, Keichiro; Takano, Hideki; Horikami, Kunihiko; Ishiguro, Yukio; Kaneko, Kunio; Hara, Toshiharu.

    1983-01-01

    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)

  5. The Analysis and the Performance Simulation of the Capacity of Bit-interleaved Coded Modulation System

    Directory of Open Access Journals (Sweden)

    Hongwei ZHAO

    2014-09-01

    Full Text Available In this paper, the capacity of the BICM system over AWGN channels is first analyzed; the curves of BICM capacity versus SNR are also got by the Monte-Carlo simulations===?=== and compared with the curves of the CM capacity. Based on the analysis results, we simulate the error performances of BICM system with LDPC codes. Simulation results show that the capacity of BICM system with LDPC codes is enormously influenced by the mapping methods. Given a certain modulation method, the BICM system can obtain about 2-3 dB gain with Gray mapping compared with Non-Gray mapping. Meanwhile, the simulation results also demonstrate the correctness of the theory analysis.

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

    International Nuclear Information System (INIS)

    Tozawa, K.; Matsumoto, T.; Kakehi, I.

    2000-04-01

    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

  7. Tokamak Systems Code

    International Nuclear Information System (INIS)

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

    1985-03-01

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

  8. CASKETSS-2: a computer code system for thermal and structural analysis of nuclear fuel shipping casks (version 2)

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1991-08-01

    A computer program CASKETSS-2 has been developed for the purpose of thermal and structural analysis of nuclear fuel shipping casks. CASKETSS-2 means a modular code system for CASK Evaluation code system Thermal and Structural Safety (Version 2). Main features of CASKETSS-2 are as follow; (1) Thermal and structural analysis computer programs for one-, two-, three-dimensional geometries are contained in the code system. (2) There are simplified computer programs and a detailed one in the structural analysis part in the code system. (3) Input data generator is provided in the code system. (4) Graphic computer program is provided in the code system. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)

  9. System transient analysis code development for low pressure and low power

    International Nuclear Information System (INIS)

    Kim, Hee Cheol

    1998-02-01

    A real time reactor system analysis code, ARTIST, based on drift flux model has been developed to investigate the transient system behavior under low pressure, low flow and low power conditions with noncondensable gas present in the system. The governing equations of the ARTIST code consist of three mass continuity equations (steam, liquid and noncondensable), two energy equations (gas and mixture) and one momentum equation (mixture) constituted with the drift flux model. The capability of ARTIST in predicting two-phase flow void distribution in the system has been validated against experimental data. The results of the ARTIST axial void distribution at low pressure and low flow, are far better than the results of both the homogeneous model of TASS code and the two-fluid model of RELAP5/MOD3 code. Also, RELAP5/MOD3 calculation shows the large amplitude of void fraction oscillations at low pressure. These results imply that interfacial momentum transfer terms in the two-fluid model formulation should be carefully constituted, especially for the low pressure condition due to the big density differences between steam and water. Thermal-hydraulic state solution scheme is developed when noncondensable gas exists. Numerical consistency and convergence of obtaining equilibrium state is tested with the ideal problems for various situations including very low partial pressure conditions. Calculated thermal-hydraulic state for each test shows consistent and expected behaviour. A new multi-layer back propagation network algorithm for calculating the departure from nucleate boiling ratio (DNBR) is developed and adopted in ARTIST code in order to have real-time DNBR evaluation by eliminating the tandem procedure of the transient DNBR calculation. The algorithm trained by different patterns generated by latin hypercube sampling method on the performance space is tested for the randomly sampled untrained data and the transient DNBR data. The uncertainty of the algorithm is

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

    International Nuclear Information System (INIS)

    2014-02-01

    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

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

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Kim, Kyung Doo; Park, Cheol; Bae, Sung Won; Baek, Won Pil; Song, Cheol hwa; Jeong, Jae Jun; Lee, Seung Wook; Hwang, Moon Kyu; Lee, Chang Sup

    2011-04-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Kim, Kyung Doo; Park, Cheol; Bae, Sung Won; Baek, Won Pil; Song, Cheol hwa; Jeong, Jae Jun; Lee, Seung Wook; Hwang, Moon Kyu; Lee, Chang Sup [KAERI, Daejeon (Korea, Republic of)

    2011-04-15

    The following KAERI-VAEI collaboration works have been performed during the 2 year project ('09.4{approx}'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

  13. Design and Analysis of Self-Healing Tree-Based Hybrid Spectral Amplitude Coding OCDMA System

    Directory of Open Access Journals (Sweden)

    Waqas A. Imtiaz

    2017-01-01

    Full Text Available This paper presents an efficient tree-based hybrid spectral amplitude coding optical code division multiple access (SAC-OCDMA system that is able to provide high capacity transmission along with fault detection and restoration throughout the passive optical network (PON. Enhanced multidiagonal (EMD code is adapted to elevate system’s performance, which negates multiple access interference and associated phase induced intensity noise through efficient two-matrix structure. Moreover, system connection availability is enhanced through an efficient protection architecture with tree and star-ring topology at the feeder and distribution level, respectively. The proposed hybrid architecture aims to provide seamless transmission of information at minimum cost. Mathematical model based on Gaussian approximation is developed to analyze performance of the proposed setup, followed by simulation analysis for validation. It is observed that the proposed system supports 64 subscribers, operating at the data rates of 2.5 Gbps and above. Moreover, survivability and cost analysis in comparison with existing schemes show that the proposed tree-based hybrid SAC-OCDMA system provides the required redundancy at minimum cost of infrastructure and operation.

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

    Science.gov (United States)

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

    2013-07-01

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

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

    International Nuclear Information System (INIS)

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

    1996-02-01

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

  16. 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.

  17. 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.

  18. DANDE-a linked code system for core neutronics/depletion analysis

    International Nuclear Information System (INIS)

    LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.

    1986-01-01

    This report describes DANDE-a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the course of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of the reactor fuel under increased burn conditions. The operation of the code system is illustrated in this report by two actual problems

  19. DANDE: a linked code system for core neutronics/depletion analysis

    International Nuclear Information System (INIS)

    LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.

    1986-01-01

    This report describes DANDE - a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the cource of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of reactor fuel under increased burn conditions. The operation of the code system is illustrated in this report by two sample problems. 25 refs

  20. DANDE: a linked code system for core neutronics/depletion analysis

    International Nuclear Information System (INIS)

    LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.

    1985-06-01

    This report describes DANDE - a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the course of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of reactor fuel under increased burn conditions. The operation of the code system is made clear in this report by following a sample problem

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

    International Nuclear Information System (INIS)

    Lee, Y. G.; Kim, J. W.; Yoon, S. J.; Park, G. C.

    2010-10-01

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-10-15

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

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

    International Nuclear Information System (INIS)

    Oliveira, Carlos; Salgado, Jose

    1998-01-01

    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

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

    International Nuclear Information System (INIS)

    Cincotti, V.; D'Andrea, A.

    1989-07-01

    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

  5. Integrated intra-subassembly treatment in the SASSYS-1 LMR systems analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Dunn, F.

    1992-09-01

    This report discusses a hot channel treatment which has been added to the SASSYS-1 LMR systems analysis code by providing for a multiple pin treatment of each of one or more subassemblies. This is an explicit calculation of intra-subassembly effects, not a hot-channel adjustment to a calculated average channel. Thus, the code can account for effects such as transient flow redistribution, both within a subassembly and among subassemblies. The code now provides a total integrated thermal hydraulic treatment including a multiple pin treatment within subassemblies, a multi-channel treatment of the whole core, and models for the primary coolant loops, the intermediate coolant loops, the steam generators, and the balance of plant. Currently the multiple-pin option is only implemented for single-phase calculations. It is not applicable after the onset of boiling or pin disruption. The new multiple pin treatment is being verified with detailed temperature data from instrumented subassemblies in EBR-II, both steady-state and transient, with special emphasis on passive safety tests such as SHRT-45. For the SHRT-45 test, excellent agreement is obtained between code predictions and experimental measurements of coolant temperatures.

  6. Integrated intra-subassembly treatment in the SASSYS-1 LMR systems analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Dunn, F.

    1992-01-01

    This report discusses a hot channel treatment which has been added to the SASSYS-1 LMR systems analysis code by providing for a multiple pin treatment of each of one or more subassemblies. This is an explicit calculation of intra-subassembly effects, not a hot-channel adjustment to a calculated average channel. Thus, the code can account for effects such as transient flow redistribution, both within a subassembly and among subassemblies. The code now provides a total integrated thermal hydraulic treatment including a multiple pin treatment within subassemblies, a multi-channel treatment of the whole core, and models for the primary coolant loops, the intermediate coolant loops, the steam generators, and the balance of plant. Currently the multiple-pin option is only implemented for single-phase calculations. It is not applicable after the onset of boiling or pin disruption. The new multiple pin treatment is being verified with detailed temperature data from instrumented subassemblies in EBR-II, both steady-state and transient, with special emphasis on passive safety tests such as SHRT-45. For the SHRT-45 test, excellent agreement is obtained between code predictions and experimental measurements of coolant temperatures.

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

    International Nuclear Information System (INIS)

    Likhanskii, V.; Evdokimov, I.; Sorokin, A.; Khromov, A.; Kanukova, V.; Apollonova, O.; Ugryumov, A.

    2008-01-01

    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)

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

    Science.gov (United States)

    Konnik, Mikhail V.

    2012-04-01

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

  9. 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.

  10. 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.

  11. Analysis of recent post irradiation tests by Japanese and French burnup analysis code systems

    International Nuclear Information System (INIS)

    Iwasaki, Tomohiko; Hiraizumi, Hiroaki; Youinou, Gilles

    2002-01-01

    Benchmark problem based on Japanese Post Irradiation Experiment (PIE) data was analyzed by Japanese burnup analysis code and French one under the cooperative research program between the Japanese University Association (JUA) in Japan and Commissariat a l'Enegie Atomique (CEA) in France. Significant discrepancies over 10% were found between the Japanese and French results for 238 Pu, 243 Am, 244 Cm, 125 Sb, 154 Eu, 134 Cs and 144 Ce. It is supposed that the difference of C/E for 243 Am and 244 Cm between Japanese results and French ones is due to the (n,gamma) reaction of 242m Am. For 125 Sb and 154 Eu, the C/E values are improved by using new cross section and fission yield libraries. (author)

  12. PRAAGE-1988: An interactive IBM-PC code for aging analysis of NUREG-1150 systems

    International Nuclear Information System (INIS)

    Fullwood, R.R.; Shier, W.G.

    1988-01-01

    Probabilistic Risk Assessments (PRA) contain a great deal of information for estimating the risk of a nuclear power plant but do not consider aging. PRAAGE (PRA+AGE) is an interactive, IBM-PC code for processing PRA-developed system models using non-aged failure rate data in conjunction with user-supplied time-dependent nuclear plant experience component failure rate data to determine the effects of component aging on a system's reliability as well as providing the age-dependent importances of various generic components. This paper describes the structure, use and application of PRAAGE to the aging analysis of the Peach Bottom 2 RHR system in the LPCI and SDC modes of operation. 4 refs., 15 figs., 5 tabs

  13. Analysis of a distributed pulse power system using a circuit analysis code

    International Nuclear Information System (INIS)

    Hoeft, L.O.; BDM Corp., Albuquerque, NM)

    1979-01-01

    A sophisticated computer code (SCEPTRE), used to analyze electronic circuits, was used to evaluate the performance of a large flash x-ray machine. This device was considered to be a transmission line whose impedance varied with position. This distributed system was modeled by lumped parameter sections with time constants of 1 ns. The model was used to interpret voltage, current, and radiation measurements in terms of diode performance. The effects of tube impedance, diode model, switch behavior, and potential geometric modifications were determined. The principal conclusions were that, since radiation output depends strongly on voltage, diode impedance was much more important than the other parameters, and the charge voltage must be accurately known

  14. Fuel performance analysis code 'FAIR'

    International Nuclear Information System (INIS)

    Swami Prasad, P.; Dutta, B.K.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

    1994-01-01

    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

  15. Development of the fire PSA methodology and the fire analysis computer code system

    International Nuclear Information System (INIS)

    Katsunori, Ogura; Tomomichi, Ito; Tsuyoshi, Uchida; Yusuke, Kasagawa

    2009-01-01

    Fire PSA methodology has been developed and was applied to NPPs in Japan for power operation and LPSD states. CDFs of preliminary fire PSA for power operation were the higher than that of internal events. Fire propagation analysis code system (CFAST/FDS Network) was being developed and verified thru OECD-PRISME Project. Extension of the scope for LPSD state is planned to figure out the risk level. In order to figure out the fire risk level precisely, the enhancement of the methodology is planned. Verification and validation of phenomenological fire propagation analysis code (CFAST/FDS Network) in the context of Fire PSA. Enhancement of the methodology such as an application of 'Electric Circuit Analysis' in NUREG/CR-6850 and related tests in order to quantify the hot-short effect precisely. Development of seismic-induced fire PSA method being integration of existing seismic PSA and fire PSA methods is ongoing. Fire PSA will be applied to review the validity of fire prevention and mitigation measures

  16. 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.

  17. ESCADRE and ICARE code systems

    International Nuclear Information System (INIS)

    Reocreux, M.; Gauvain, J.

    1992-01-01

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

  18. Fuel relocation modeling in the SAS4A accident analysis code system

    International Nuclear Information System (INIS)

    Tentner, A.M.; Miles, K.J.; Kalimullah; Hill, D.J.

    1986-01-01

    The SAS4A code system has been designed for the analysis of the initial phase of Hypothetical Core Disruptive Accidents (HCDAs) up to gross melting or failure of the subassembly walls. During such postulated accident scenarios as the Loss-of-Flow (LOF) and Transient-Overpower (TOP) events, the relocation of the fuel plays a key role in determining the sequence of events and the amount of energy produced before neutronic shutdown. This paper discusses the general strategy used in modelong the various phenomena which lead to fuel relocation and presents the key fuel relocation models used in SAS4A. The implications of these models for the whole-core accident analysis as well as recent results of fuel relocation are emphasized. 12 refs

  19. 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.

  20. Code division multiple-access techniques in optical fiber networks. II - Systems performance analysis

    Science.gov (United States)

    Salehi, Jawad A.; Brackett, Charles A.

    1989-08-01

    A technique based on optical orthogonal codes was presented by Salehi (1989) to establish a fiber-optic code-division multiple-access (FO-CDMA) communications system. The results are used to derive the bit error rate of the proposed FO-CDMA system as a function of data rate, code length, code weight, number of users, and receiver threshold. The performance characteristics for a variety of system parameters are discussed. A means of reducing the effective multiple-access interference signal by placing an optical hard-limiter at the front end of the desired optical correlator is presented. Performance calculations are shown for the FO-CDMA with an ideal optical hard-limiter, and it is shown that using a optical hard-limiter would, in general, improve system performance.

  1. 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.

  2. ETR/ITER systems code

    International Nuclear Information System (INIS)

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

    1988-04-01

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

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

    International Nuclear Information System (INIS)

    Kawai, Hiroshi

    2002-01-01

    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)

  4. Preliminary design of a small air loop for system analysis and validation of Cathare code

    International Nuclear Information System (INIS)

    Marchand, M.; Saez, M.; Tauveron, N.; Tenchine, D.; Germain, T.; Geffraye, G.; Ruby, G.P.

    2007-01-01

    The French Atomic Energy Commission (Cea) is carrying on the design of a Small Air Loop for System Analysis (SALSA), devoted to the study of gas cooled nuclear reactors behaviour in normal and incidental/accidental operating conditions. The reduced size of the SALSA components compared to a full-scale reactor and air as gaseous coolant instead of Helium will allow an easy management of the loop. The main purpose of SALSA will be the validation of the associated thermal hydraulic safety simulation codes, like CATHARE. The main goal of this paper is to present the methodology used to define the characteristics of the loop. In a first step, the study has been focused on a direct-cycle system for the SALSA loop with few global constraints using a similarity analysis to support the definition and design of the loop. Similarity requirements have been evaluated to determine the scale factors which have to be applied to the SALSA loop components. The preliminary conceptual design of the SALSA plant with a definition of each component has then be carried out. The whole plant has been modelled using the CATHARE code. Calculations of the SALSA steady-state in nominal conditions and of different plant transients in direct-cycle have been made. The first system results obtained on the global behaviour of the loop confirm that SALSA can be representative of a Gas-Cooled nuclear reactor with some minor design modifications. In a second step, the current prospects focus on the SALSA loop capability to reproduce correctly the heat transfer occurring in specific incidental situations. Heat decay removal by natural convection is a crucial point of interest. The first results show that the behaviour and the efficiency of the loop are strongly influenced by the definition of the main parameters for each component. A complete definition of SALSA is under progress. (authors)

  5. Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors

    International Nuclear Information System (INIS)

    Kobori, Hikaru; Kasada, Ryuta; Hiwatari, Ryoji; Konishi, Satoshi

    2016-01-01

    Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO_2 emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO_2 emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO_2 emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO_2 emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO_2 emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

  6. Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kobori, Hikaru [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hiwatari, Ryoji [Central Research Institute of Electric Power Industry, Tokyo (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan)

    2016-11-01

    Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO{sub 2} emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

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

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Jeong, Jae Jun; Lee, Won Jae

    1998-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Jeong, Jae Jun; Lee, Won Jae [KAERI, Taejon (Korea, Republic of)

    1998-10-01

    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.

  9. SCALE Code System

    Energy Technology Data Exchange (ETDEWEB)

    Jessee, Matthew Anderson [ORNL

    2016-04-01

    The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.SCALE 6.2 provides many new capabilities and significant improvements of existing features.New capabilities include:• ENDF/B-VII.1 nuclear data libraries CE and MG with enhanced group structures,• Neutron covariance data based on ENDF/B-VII.1 and supplemented with ORNL data,• Covariance data for fission product yields and decay constants,• Stochastic uncertainty and correlation quantification for any SCALE sequence with Sampler,• Parallel calculations with KENO,• Problem-dependent temperature corrections for CE calculations,• CE shielding and criticality accident alarm system analysis with MAVRIC,• CE

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

    International Nuclear Information System (INIS)

    Kawamoto, Yosuke; Chiba, Go; Tsuji, Masashi; Narabayashi, Tadashi

    2013-01-01

    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)

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  12. Fuel relocation modeling in the SAS4A accident analysis code system

    International Nuclear Information System (INIS)

    Tentner, A.M.; Miles, K.J.

    1985-01-01

    SAS4A is a new code system which has been designed for analyzing the initial phase of Hypothetical Core Disruptive Accidents (HCDAs) up to gross melting or failure of the subassembly walls. During such postulated accident scenarios as the Loss-of-Flow (LOF) and Transient-Overpower (TOP) events, the relocation of the fuel plays a key role in determining the sequence of events and the amount of energy produced before neutronic shutdown. This paper discusses the general strategy used in modeling the various phenomena which lead to fuel relocation and presents the key fuel relocation models used in SAS4A. The implications of these models for the whole-core accident analysis as well as recent results of fuel motion experiment analyses are also presented

  13. A new balance-of-plant model for the SASSYS-1 LMR systems analysis code

    International Nuclear Information System (INIS)

    Briggs, L.L.

    1989-01-01

    A balance-of-plant model has been added to the SASSYS-1 liquid metal reactor systems analysis code. Until this addition, the only waterside component which SASSYS-1 could explicitly model was the water side of a steam generator, with the remainder of the water side represented by boundary conditions on the steam generator. The balance-of-plant model is based on the model used for the sodium side of the plant. It will handle subcooled liquid water, superheated steam, and saturated two-phase fluid. With the exception of heated flow paths in heaters, the model assumes adiabatic conditions along flow paths; this assumption simplifies the solution procedure while introducing very little error for a wide range of reactor plant problems. Only adiabatic flow is discussed in this report. 3 refs., 4 figs

  14. Analysis of PPM-CDMA and OPPM-CDMA communication systems with new optical code

    Science.gov (United States)

    Liu, F.; Ghafouri-Shiraz, H.

    2005-11-01

    A novel type of optical spreading sequences, named the 'new-Modified Prime Code (nMPC)', is proposed for use in synchronous direct-detection optical code-division multiple-access (CDMA) systems which employ both pulse position modulation (PPM) and overlapping pulse position modulation (OPPM) schemes. The upper bounds on the bit error rate (BER) for nMPC used in PPM-CDMA systems are derived and compared with the respective systems, using a modified prime code (MPC) and a padded modified prime code (PMPC). The nMPC is further applied to the OPPM-CDMA system and the system with a proposed interference cancellation scheme. Our results show that under the same conditions the PPM-CDMA system performances are more improved with the use of nMPC than with the two other traditional codes. Moreover, they show that the system performances are significantly enhanced by the proposed interference reduction methods, if the nMPC is used in the OPPM-CDMA systems.

  15. Development of a system code for transient analysis in a HTGR

    International Nuclear Information System (INIS)

    Lee, Tae Beom

    2004-02-01

    A GAMMA (GAs Multi-component Multi-dimensional Analysis) code is developed for transient analysis and air ingress analysis in High Temperature Gas-cooled Reactors (HTGR). The PBMR of ESKOM is selected as a reference plant for the High Temperature Gas-cooled Reactor here, which uses a direct helium cycle and pebble fuel. Physical models included in GAMMA are the pebble conduction model, radiation heat transfer model, point kinetics model, decay heat model, and component models for break flow, valve, pump, cooler, power conversion unit model. The temperature distribution and the flow distribution of the PBMR are calculated for initial and accident core in the present study. In the accident analysis, typical design basis accident (DBA), including the load transient accident and depressurization accident into the system are selected and analyzed in detail. The predictions by GAMMA for PBMR at 100% power are compared with those by VSOP and PBR S IM. It turns out that the temperature in the upper region in the third channel predicted by GAMMA is about 62 .deg. C at maximum higher than that by VSOP, but is pretty close to that by PBR S IM. The center temperature of the fuel shows that that predicted by considering swelling effect is higher than that without swelling effect by about 10 .deg. C. The net efficiency of direct system is higher than that of indirect system due to an effect of the circulator power. The transient capability of GAMMA is validated through analytical solution and PBR S IM analyzing the depressurization (Loss Of Coolant Accident, LOCA) and load transient accident. After the LOCA the system pressure decreases dramatically from 8MPa to 0.4MPa within 2 sec. After the PI (Proportional-plus-Integral) controller senses that the power shaft is over the set-point of 3,600 rpm, the bypass valve makes shaft speed back to the set-point

  16. Light-water reactor safety analysis codes

    International Nuclear Information System (INIS)

    Jackson, J.F.; Ransom, V.H.; Ybarrondo, L.J.; Liles, D.R.

    1980-01-01

    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

  17. 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.

  18. Error analysis of supercritical water correlations using ATHLET system code under DHT conditions

    Energy Technology Data Exchange (ETDEWEB)

    Samuel, J., E-mail: jeffrey.samuel@uoit.ca [Univ. of Ontario Inst. of Tech., Oshawa, ON (Canada)

    2014-07-01

    The thermal-hydraulic computer code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients) is used for analysis of anticipated and abnormal plant transients, including safety analysis of Light Water Reactors (LWRs) and Russian Graphite-Moderated High Power Channel-type Reactors (RBMKs). The range of applicability of ATHLET has been extended to supercritical water by updating the fluid-and transport-properties packages, thus enabling the code to the used in analysis of SuperCritical Water-cooled Reactors (SCWRs). Several well-known heat-transfer correlations for supercritical fluids were added to the ATHLET code and a numerical model was created to represent an experimental test section. In this work, the error in the Heat Transfer Coefficient (HTC) calculation by the ATHLET model is studied along with the ability of the various correlations to predict different heat transfer regimes. (author)

  19. Analysis of Korean Nuclear Fuel Cycle System by Using DANESS Code

    International Nuclear Information System (INIS)

    Jeong, Chang Joon

    2009-08-01

    Korean fast reactor scenarios have been analyzed for various kinds of conversion ratio (CR) by the DANESS system dynamic analysis code. The once-through fuel cycle analysis was modeled based on the Korean 'National Energy Basic Plan' up to 2030 and a postulated nuclear demand growth rate until 2150. The fast reactor scenario analysis has been performed for three kinds of conversion ratios such as 0.3, 0.61 and 1.0. Through the calculations, the nuclear reactor deployment scenario, front-end cycle, back-end cycle, and long-term heat load have been investigated. From the once-through results, it is shown that the nuclear power demand would be ∼70 GWe and the total amount of the spent fuel accumulated by 2150 would be ∼168000 t. The fast reactor (FR) scenario analysis results show that the spent fuel inventory and out-pile transuranic element (TRU) can be reduced by increasing the fast reactor conversion ratio. Furthermore, the long-term heat load of spent fuel decreases with increasing the conversion ratio. However, it is known that the deployment of a fast reactor of low conversion ratio does not much reduce the spent fuel and out-pile TRU inventory due to the fast reactor deployment limitation which is related to the availability of TRU

  20. 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.

  1. Benchmarking the MCNP code for Monte Carlo modelling of an in vivo neutron activation analysis system.

    Science.gov (United States)

    Natto, S A; Lewis, D G; Ryde, S J

    1998-01-01

    The Monte Carlo computer code MCNP (version 4A) has been used to develop a personal computer-based model of the Swansea in vivo neutron activation analysis (IVNAA) system. The model included specification of the neutron source (252Cf), collimators, reflectors and shielding. The MCNP model was 'benchmarked' against fast neutron and thermal neutron fluence data obtained experimentally from the IVNAA system. The Swansea system allows two irradiation geometries using 'short' and 'long' collimators, which provide alternative dose rates for IVNAA. The data presented here relate to the short collimator, although results of similar accuracy were obtained using the long collimator. The fast neutron fluence was measured in air at a series of depths inside the collimator. The measurements agreed with the MCNP simulation within the statistical uncertainty (5-10%) of the calculations. The thermal neutron fluence was measured and calculated inside the cuboidal water phantom. The depth of maximum thermal fluence was 3.2 cm (measured) and 3.0 cm (calculated). The width of the 50% thermal fluence level across the phantom at its mid-depth was found to be the same by both MCNP and experiment. This benchmarking exercise has given us a high degree of confidence in MCNP as a tool for the design of IVNAA systems.

  2. Advanced video coding systems

    CERN Document Server

    Gao, Wen

    2015-01-01

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

  3. Analysis of functioning and efficiency of a code blue system in a tertiary care hospital.

    Science.gov (United States)

    Monangi, Srinivas; Setlur, Rangraj; Ramanathan, Ramprasad; Bhasin, Sidharth; Dhar, Mridul

    2018-01-01

    "Code blue" (CB) is a popular hospital emergency code, which is used by hospitals to alert their emergency response team of any cardiorespiratory arrest. The factors affecting the outcomes of emergencies are related to both the patient and the nature of the event. The primary objective was to analyze the survival rate and factors associated with survival and also practical problems related to functioning of a CB system (CBS). After the approval of hospital ethics committee, an analysis and audit was conducted of all patients on whom a CB had been called in our tertiary care hospital over 24 months. Data collected were demographic data, diagnosis, time of cardiac arrest and activation of CBS, time taken by CBS to reach the patient, presenting rhythm on arrival of CB team, details of cardiopulmonary resuscitation (CPR) such as duration and drugs given, and finally, events and outcomes. Chi-square test and logistic regression analysis were used to analyze the data. A total of 720 CB calls were initiated during the period. After excluding 24 patients, 694 calls were studied and analyzed. Six hundred and twenty were true calls and 74 were falls calls. Of the 620, 422 were cardiac arrests and 198 were medical emergencies. Overall survival was 26%. Survival in patients with cardiac arrests was 11.13%. Factors such as age, presenting rhythm, and duration of CPR were found to have a significant effect on survival. Problems encountered were personnel and equipment related. A CBS is effective in improving the resuscitation efforts and survival rates after inhospital cardiac arrests. Age, presenting rhythm at the time of arrest, and duration of CPR have significant effect on survival of the patient after a cardiac arrest. Technical and staff-related problems need to be considered and improved upon.

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

    International Nuclear Information System (INIS)

    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

  5. 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.

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

    International Nuclear Information System (INIS)

    Haney, S.W.; Barr, W.L.; Crotinger, J.A.; Perkins, L.J.; Solomon, C.J.; Chaniotakis, E.A.; Freidberg, J.P.; Wei, J.; Galambos, J.D.; Mandrekas, J.

    1992-01-01

    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

  7. The SAS4A/SASSYS-1 Safety Analysis Code System, Version 5

    Energy Technology Data Exchange (ETDEWEB)

    Fanning, T. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-01-01

    The SAS4A/SASSYS-1 computer code is developed by Argonne National Laboratory for thermal, hydraulic, and neutronic analysis of power and flow transients in liquidmetal- cooled nuclear reactors (LMRs). SAS4A was developed to analyze severe core disruption accidents with coolant boiling and fuel melting and relocation, initiated by a very low probability coincidence of an accident precursor and failure of one or more safety systems. SASSYS-1, originally developed to address loss-of-decay-heat-removal accidents, has evolved into a tool for margin assessment in design basis accident (DBA) analysis and for consequence assessment in beyond-design-basis accident (BDBA) analysis. SAS4A contains detailed, mechanistic models of transient thermal, hydraulic, neutronic, and mechanical phenomena to describe the response of the reactor core, its coolant, fuel elements, and structural members to accident conditions. The core channel models in SAS4A provide the capability to analyze the initial phase of core disruptive accidents, through coolant heat-up and boiling, fuel element failure, and fuel melting and relocation. Originally developed to analyze oxide fuel clad with stainless steel, the models in SAS4A have been extended and specialized to metallic fuel with advanced alloy cladding. SASSYS-1 provides the capability to perform a detailed thermal/hydraulic simulation of the primary and secondary sodium coolant circuits and the balance-ofplant steam/water circuit. These sodium and steam circuit models include component models for heat exchangers, pumps, valves, turbines, and condensers, and thermal/hydraulic models of pipes and plena. SASSYS-1 also contains a plant protection and control system modeling capability, which provides digital representations of reactor, pump, and valve controllers and their response to input signal changes.

  8. 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

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

    International Nuclear Information System (INIS)

    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

  10. 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.

  11. Analysis of pressure wave transients and seismic response in LMFBR piping systems using the SHAPS code

    International Nuclear Information System (INIS)

    Zeuch, W.R.; Wang, C.Y.

    1985-01-01

    This paper presents some of the current capabilities of the three-dimensional piping code SHAPS and demonstrates their usefulness in handling analyses encountered in typical LMFBR studies. Several examples demonstrate the utility of the SHAPS code for problems involving fluid-structure interactions and seismic-related events occurring in three-dimensional piping networks. Results of two studies of pressure wave propagation demonstrate the dynamic coupling of pipes and elbows producing global motion and rigorous treatment of physical quantities such as changes in density, pressure, and strain energy. Results of the seismic analysis demonstrate the capability of SHAPS to handle dynamic structural response within a piping network over an extended transient period of several seconds. Variation in dominant stress frequencies and global translational frequencies were easily handled with the code. 4 refs., 10 figs

  12. A new balance-of-plant model for the SASSYS-1 LMR systems analysis code

    International Nuclear Information System (INIS)

    Briggs, L.L.

    1991-01-01

    In this paper, a balance-of-plant model is developed for the SASSYS code. This model represents the balance of plant as a network of components. It interfaces with the existing SASSYS code through the water side of the steam generator. The network representation provides a discretization of the mass, momentum, and energy equations and the equation of state and allows a simultaneous solution for the changes in pressure, flow, and enthalpy throughout the waterside system. The model has been tested for several types of transients and been found to perform both accurately and efficiently

  13. The development of the thermohydraulic analysis code for the passive containment cooling system: PCCSAC

    International Nuclear Information System (INIS)

    Wang Jianyu; Zhang Shenru; Min Yuanyou

    1994-01-01

    To estimate the performance of the passive containment cooling system (PCCS) of the AC-600 nuclear power plant, the PCCSAC code is developed currently by the jointed efforts between Tsinghua University and NPIC. Different features on the passive behavior of the system and the main components of the containment are considered in the code which is needed by the further AC-600 R and D Program. With a brief description of the AC-600 passive containment cooling system and components, the main thermohydraulic models and numerical scheme used in the PCCSAC code are introduced and the selected results of the verification and the prediction for the performance of the AC-600 passive containment cooling system under LOCA and a steam line break accident are presented to preliminarily demonstrate the applicability and reliability of the PCCSAC model. The current PCCSAC model is conservative and a further 2-D PCCSAC version is under consideration in addition to provide the database for models from some tests associated with the components and systems unique to AC-600 nuclear power plant to meet the requirement of the more realistic modelization for the AC-600 passive containment cooling system. (author)

  14. Inclusive bit error rate analysis for coherent optical code-division multiple-access system

    Science.gov (United States)

    Katz, Gilad; Sadot, Dan

    2002-06-01

    Inclusive noise and bit error rate (BER) analysis for optical code-division multiplexing (OCDM) using coherence techniques is presented. The analysis contains crosstalk calculation of the mutual field variance for different number of users. It is shown that the crosstalk noise depends deeply on the receiver integration time, the laser coherence time, and the number of users. In addition, analytical results of the power fluctuation at the received channel due to the data modulation at the rejected channels are presented. The analysis also includes amplified spontaneous emission (ASE)-related noise effects of in-line amplifiers in a long-distance communication link.

  15. The CORSYS neutronics code system

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  16. Computer codes for safety analysis

    International Nuclear Information System (INIS)

    Holland, D.F.

    1986-11-01

    Computer codes for fusion safety analysis have been under development in the United States for about a decade. This paper will discuss five codes that are currently under development by the Fusion Safety Program. The purpose and capability of each code will be presented, a sample given, followed by a discussion of the present status and future development plans

  17. Tandem Mirror Reactor Systems Code (Version I)

    International Nuclear Information System (INIS)

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

    1985-09-01

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

  18. 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.

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

    International Nuclear Information System (INIS)

    Jang, H. S.; Jeon, M. H.; Cho, N. J. and others

    1992-01-01

    The objectives of this study is to develop a system of computer codes for postulated severe accident analyses in nuclear power plants. This system of codes is necessary to conduct Individual Plant Examination for domestic nuclear power plants. As a result of this study, one can conduct severe accident assessments more easily, and can extract the plant-specific vulnerabilities for severe accidents and at the same time the ideas for enhancing overall accident-resistance. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Carlos, S., E-mail: scarlos@iqn.upv.es [Department d’Enginyeria Química i Nuclear, Universitat Politècnica de València, Camí de Vera s.n, 46022 València (Spain); Sánchez, A. [Department d’Estadística Aplicada i Qualitat, Universitat Politècnica de València, Camí de Vera s.n, 46022 València (Spain); Ginestar, D. [Department de Matemàtica Aplicada, Universitat Politècnica de València, Camí de Vera s.n, 46022 València (Spain); Martorell, S. [Department d’Enginyeria Química i Nuclear, Universitat Politècnica de València, Camí de Vera s.n, 46022 València (Spain)

    2013-09-15

    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

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

    International Nuclear Information System (INIS)

    Austregesilo, Henrique; Bals, Christine; Trambauer, Klaus

    2007-01-01

    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 4 C oxidation do not affect significantly the total calculated hydrogen release rates

  2. 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)

  3. Neutronics/Thermo-fluid Coupled Analysis of PMR-200 Equilibrium Cycle by CAPP/GAMMA+ Code System

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyun Chul; Tak, Nam-il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The equilibrium core was obtained by performing CAPP stand-alone multi-cycle depletion calculation with critical rod position search. In this work, a code system for coupled neutronics and thermo-fluids simulation was developed using CAPP and GAMMA+ codes. A server program, INTCA, controls the two codes for coupled calculations and performs the mapping between the variables of the two codes based on the nodalization of the two codes. In order to extend the knowledge about the coupled behavior of a prismatic VHTR, the CAPP/GAMMA+ code system was applied to steady state performance analysis of PMR-200. The coupled calculation was carried out for the equilibrium core of PMR-200 from BOC to EOC. The peak fuel temperature was predicted to be 1372 .deg. C near MOC. However, the cycle-average fuel temperature was calculated as 1230 .deg. C, which is slightly below the design target of 1250 .deg. C. In addition, significant impact of the bypass flow on the central reflector temperature was found. Without bypass flow, the temperature of the active core region was slightly decreased while the temperature of the central and side reflector region was increased much. The both changes in the temperature increase the multiplication factor and the total change of the multiplication factor was more than 300 pcm. On the other hand, the effect of the bypass flow on the power density profile was not significant.

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

    Science.gov (United States)

    Zhao, Hui; Wei, Jingxuan

    2014-09-01

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

  5. Performance Analysis for Cooperative Communication System with QC-LDPC Codes Constructed with Integer Sequences

    Directory of Open Access Journals (Sweden)

    Yan Zhang

    2015-01-01

    Full Text Available This paper presents four different integer sequences to construct quasi-cyclic low-density parity-check (QC-LDPC codes with mathematical theory. The paper introduces the procedure of the coding principle and coding. Four different integer sequences constructing QC-LDPC code are compared with LDPC codes by using PEG algorithm, array codes, and the Mackey codes, respectively. Then, the integer sequence QC-LDPC codes are used in coded cooperative communication. Simulation results show that the integer sequence constructed QC-LDPC codes are effective, and overall performance is better than that of other types of LDPC codes in the coded cooperative communication. The performance of Dayan integer sequence constructed QC-LDPC is the most excellent performance.

  6. Parallel processing of structural integrity analysis codes

    International Nuclear Information System (INIS)

    Swami Prasad, P.; Dutta, B.K.; Kushwaha, H.S.

    1996-01-01

    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

  7. From the direct numerical simulation to system codes-perspective for the multi-scale analysis of LWR thermal hydraulics

    International Nuclear Information System (INIS)

    Bestion, D.

    2010-01-01

    A multi-scale analysis of water-cooled reactor thermal hydraulics can be used to take advantage of increased computer power and improved simulation tools, including Direct Numerical Simulation (DNS), Computational Fluid Dynamics (CFD) (in both open and porous mediums), and system thermalhydraulic codes. This paper presents a general strategy for this procedure for various thermalhydraulic scales. A short state of the art is given for each scale, and the role of the scale in the overall multi-scale analysis process is defined. System thermalhydraulic codes will remain a privileged tool for many investigations related to safety. CFD in porous medium is already being frequently used for core thermal hydraulics, either in 3D modules of system codes or in component codes. CFD in open medium allows zooming on some reactor components in specific situations, and may be coupled to the system and component scales. Various modeling approaches exist in the domain from DNS to CFD which may be used to improve the understanding of flow processes, and as a basis for developing more physically based models for macroscopic tools. A few examples are given to illustrate the multi-scale approach. Perspectives for the future are drawn from the present state of the art and directions for future research and development are given

  8. Preliminary Study of 1D Thermal-Hydraulic System Analysis Code Using the Higher-Order Numerical Scheme

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Woong; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    The existing nuclear system analysis codes such as RELAP5, TRAC, MARS and SPACE use the first-order numerical scheme in both space and time discretization. However, the first-order scheme is highly diffusive and less accurate due to the first order of truncation error. So, the numerical diffusion problem which makes the gradients to be smooth in the regions where the gradients should be steep can occur during the analysis, which often predicts less conservatively than the reality. Therefore, the first-order scheme is not always useful in many applications such as boron solute transport. RELAP7 which is an advanced nuclear reactor system safety analysis code using the second-order numerical scheme in temporal and spatial discretization is being developed by INL (Idaho National Laboratory) since 2011. Therefore, for better predictive performance of the safety of nuclear reactor systems, more accurate nuclear reactor system analysis code is needed for Korea too to follow the global trend of nuclear safety analysis. Thus, this study will evaluate the feasibility of applying the higher-order numerical scheme to the next generation nuclear system analysis code to provide the basis for the better nuclear system analysis code development. The accuracy is enhanced in the spatial second-order scheme and the numerical diffusion problem is alleviated while indicates significantly lower maximum Courant limit and the numerical dispersion issue which produces spurious oscillation and non-physical results in the higher-order scheme. If the spatial scheme is the first order scheme then the temporal second-order scheme provides almost the same result with the temporal firstorder scheme. However, when the temporal second order scheme and the spatial second-order scheme are applied together, the numerical dispersion can occur more severely. For the more in-depth study, the verification and validation of the NTS code built in MATLAB will be conducted further and expanded to handle two

  9. Development of long-lived radionuclide transmutation technology - Development of a code system for core analysis of the transmutation reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Nam Zin; Kim, Yong Hee; Kim, Tae Hyung; Jo, Chang Keun; Park, Chang Je [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1996-07-01

    The objective of this study is to develop a code system for core analysis= of the critical transmutation reactors utilizing fast neutrons. Core characteristics of the transmutation reactors were identified and four codes, HANCELL for pincell calculation, PRISM and AFEN-H3D for core calculation, and MA{sub B}URN for depletion calculation, were developed. The pincell calculation code is based on one-dimensional collision probability method and may provide homogenized/condensed parameters of a pincell and also can homogenize the control assembly via a nonlinear iterative method. The core calculation codes, PRISM and AFEN-H3D, solve the multi-group, multi-dimensional neutron diffusion equations for a hexagonal geometry and they are based on the finite difference method and analytic function expansion nodal (AFEN) method, respectively. The MA{sub B}URN code san analyze the behavior of actinides and fission products in a reactor core. Through benchmarking, we confirmed that the newly developed codes provide accurate solutions. 30 refs., 10 tabs., 8 figs. (author)

  10. Behavioral analysis of malicious code through network traffic and system call monitoring

    Science.gov (United States)

    Grégio, André R. A.; Fernandes Filho, Dario S.; Afonso, Vitor M.; Santos, Rafael D. C.; Jino, Mario; de Geus, Paulo L.

    2011-06-01

    Malicious code (malware) that spreads through the Internet-such as viruses, worms and trojans-is a major threat to information security nowadays and a profitable business for criminals. There are several approaches to analyze malware by monitoring its actions while it is running in a controlled environment, which helps to identify malicious behaviors. In this article we propose a tool to analyze malware behavior in a non-intrusive and effective way, extending the analysis possibilities to cover malware samples that bypass current approaches and also fixes some issues with these approaches.

  11. Performance of asynchronous fiber-optic code division multiple access system based on three-dimensional wavelength/time/space codes and its link analysis.

    Science.gov (United States)

    Singh, Jaswinder

    2010-03-10

    A novel family of three-dimensional (3-D) wavelength/time/space codes for asynchronous optical code-division-multiple-access (CDMA) systems with "zero" off-peak autocorrelation and "unity" cross correlation is reported. Antipodal signaling and differential detection is employed in the system. A maximum of [(W x T+1) x W] codes are generated for unity cross correlation, where W and T are the number of wavelengths and time chips used in the code and are prime. The conditions for violation of the cross-correlation constraint are discussed. The expressions for number of generated codes are determined for various code dimensions. It is found that the maximum number of codes are generated for S systems. The codes have a code-set-size to code-size ratio greater than W/S. For instance, with a code size of 2065 (59 x 7 x 5), a total of 12,213 users can be supported, and 130 simultaneous users at a bit-error rate (BER) of 10(-9). An arrayed-waveguide-grating-based reconfigurable encoder/decoder design for 2-D implementation for the 3-D codes is presented so that the need for multiple star couplers and fiber ribbons is eliminated. The hardware requirements of the coders used for various modulation/detection schemes are given. The effect of insertion loss in the coders is shown to be significantly reduced with loss compensation by using an amplifier after encoding. An optical CDMA system for four users is simulated and the results presented show the improvement in performance with the use of loss compensation.

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

    International Nuclear Information System (INIS)

    Jang, Sung Soon; Kwak, Sung Woo; Yoo, Ho Sik; Kim, Jung Soo; Yoon, Wan Ki

    2008-01-01

    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

  13. Elements of algebraic coding systems

    CERN Document Server

    Cardoso da Rocha, Jr, Valdemar

    2014-01-01

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

  14. Performance Analysis of a De-correlated Modified Code Tracking Loop for Synchronous DS-CDMA System under Multiuser Environment

    Science.gov (United States)

    Wu, Ya-Ting; Wong, Wai-Ki; Leung, Shu-Hung; Zhu, Yue-Sheng

    This paper presents the performance analysis of a De-correlated Modified Code Tracking Loop (D-MCTL) for synchronous direct-sequence code-division multiple-access (DS-CDMA) systems under multiuser environment. Previous studies have shown that the imbalance of multiple access interference (MAI) in the time lead and time lag portions of the signal causes tracking bias or instability problem in the traditional correlating tracking loop like delay lock loop (DLL) or modified code tracking loop (MCTL). In this paper, we exploit the de-correlating technique to combat the MAI at the on-time code position of the MCTL. Unlike applying the same technique to DLL which requires an extensive search algorithm to compensate the noise imbalance which may introduce small tracking bias under low signal-to-noise ratio (SNR), the proposed D-MCTL has much lower computational complexity and exhibits zero tracking bias for the whole range of SNR, regardless of the number of interfering users. Furthermore, performance analysis and simulations based on Gold codes show that the proposed scheme has better mean square tracking error, mean-time-to-lose-lock and near-far resistance than the other tracking schemes, including traditional DLL (T-DLL), traditional MCTL (T-MCTL) and modified de-correlated DLL (MD-DLL).

  15. CFD and system analysis code investigations of the multidimensional flow mixing phenomena in the reactor pressure vessel

    International Nuclear Information System (INIS)

    Ceuca, S.C.; Herb, J.; Schoeffel, P.J.; Hollands, T.; Austregesilo, H.; Hristov, H.V.

    2017-01-01

    The realistic numerical prediction of transient fluid-dynamic scenarios including the complex, three-dimensional flow mixing phenomena occurring in the reactor pressure vessel (RPV) both in normal or abnormal operation are an important issue in today's reactor safety assessment studies. Both Computational Fluid Dynamics (CFD) tools as well as fluid-dynamic system analysis codes, each with its advantages and drawbacks, are commonly used to model such transients. Simulation results obtained with the open-source CFD tool-box OpenFOAM and the German thermal-hydraulic system code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients), the later developed by Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) for the analysis of the whole spectrum of operational transients, design-basis accidents and beyond design basis accidents anticipated for nuclear energy facilities, are compared against experimental data from the ROssendorf Coolant Mixing (ROCOM) test facility. In the case of the OpenFOAM CFD simulations the influence of various turbulence models and numerical schemes has been assessed while in the case of the system analysis code ATHLET a multidimensional nodalization recommended for real power plant applications has been employed. The simulation results show a good agreement with the experimental data, indicating that both OpenFOAM and ATHLET can capture the key flow features of the mixing processes in the Reactor Pressure Vessel (RPV). (author)

  16. System Based Code: Principal Concept

    International Nuclear Information System (INIS)

    Yasuhide Asada; Masanori Tashimo; Masahiro Ueta

    2002-01-01

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

  17. ANTEO+: A subchannel code for thermal-hydraulic analysis of liquid metal cooled systems

    Energy Technology Data Exchange (ETDEWEB)

    Lodi, F., E-mail: francesco.lodi5@unibo.it [DIN – Laboratory of Montecuccolino, University of Bologna, Via dei Colli 16, 40136 Bologna (Italy); Grasso, G., E-mail: giacomo.grasso@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Mattioli, D., E-mail: davide.mattioli@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Sumini, M., E-mail: marco.sumini@unibo.it [DIN – Laboratory of Montecuccolino, University of Bologna, Via dei Colli 16, 40136 Bologna (Italy)

    2016-05-15

    Highlights: • The code structure is presented in detail. • The performed validation is outlined. • Results are critically discussed assessing code accuracy. • Conclusions are drawn and ground for future work identified. - Abstract: Liquid metal cooled fast reactors are promising options for achieving the high degrees of safety and sustainability demanded by the Generation IV paradigm. Among the critical aspects to be addressed in the design process, thermal-hydraulics is one of the most challenging; in order to embed safety in the core conceptualization, these aspects are to be considered at the very beginning of the design process, and translated in a design perspective. For achieving these objectives the subchannel code ANTEO+ has been conceived, able to simulate pin bundle arrangements cooled by liquid metals. The main purposes of ANTEO+ are simplifying the problem description maintaining the required accuracy, enabling a more transparent interface with the user, and having a clear and identifiable application domain, in order to help the user interpreting the results and, mostly, defining their confidence. Since ANTEO+ relies on empirical correlations, the validation phase is of paramount importance along with a clear discussion on the simplifications adopted in modeling the conservation equations. In the present work a detailed description of ANTEO+ structure is given along with a thorough validation of the main models implemented for flow split, pressure drops and subchannel temperatures. The analysis confirmed the ability of ANTEO+ in reproducing experimental data in its anticipated validity domain, with a relatively high degree of accuracy when compared to other classical subchannel tools like ENERGY-II, COBRA-IV-I-MIT and BRS-TVS.

  18. 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)

    Carvalho, F. de A.T. de.

    1985-01-01

    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) [pt

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-04-15

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

  20. Analysis of the VVER-1000 coolant transient benchmark phase 1 with the code system RELAP5/PARCS

    International Nuclear Information System (INIS)

    Victor Hugo Sanchez Espinoza

    2005-01-01

    Full text of publication follows: As part of the reactor dynamics activities of FZK/IRS, the qualification of best-estimate coupled code systems for reactor safety evaluations is a key step toward improving their prediction capability and acceptability. The VVER-1000 Coolant Transient Benchmark Phase 1 represents an excellent opportunity to validate the simulation capability of the coupled code system RELAP5/PACRS regarding both the thermal hydraulic plant response (RELAP5) using measured data obtained during commissioning tests at the Kozloduy nuclear power plant unit 6 and the neutron kinetics models of PARCS for hexagonal geometries. The Phase 1 is devoted to the analysis of the switching on of one main coolant pump while the other three pumps are in operation. It includes the following exercises: (a) investigation of the integral plant response using a best-estimate thermal hydraulic system code with a point kinetics model (b) analysis of the core response for given initial and transient thermal hydraulic boundary conditions using a coupled code system with 3D-neutron kinetics model and (c) investigation of the integral plant response using a best-estimate coupled code system with 3D-neutron kinetics. Already before the test, complex flow conditions exist within the RPV e.g. coolant mixing in the upper plenum caused by the reverse flow through the loop-3 with the stopped pump. The test is initiated by switching on the main coolant pump of loop-3 that leads to a reversal of the flow through the respective piping. After about 13 s the mass flow rate through this loop reaches values comparable with the one of the other loops. During this time period, the increased primary coolant flow causes a reduction of the core averaged coolant temperature and thus an increase of the core power. Later on, the power stabilizes at a level higher than the initial power. In this analysis, special attention is paid on the prediction of the spatial asymmetrical core cooling during

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

    International Nuclear Information System (INIS)

    Goto, Minoru; Takamatsu, Kuniyoshi

    2007-03-01

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

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

    International Nuclear Information System (INIS)

    Kliem, Soeren; Hoehne, Thomas; Rohde, Ulrich; Weiss, Frank-Peter; Kozmenkov, Yaroslav

    2008-01-01

    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)

  3. Stability analysis by ERATO code

    International Nuclear Information System (INIS)

    Tsunematsu, Toshihide; Takeda, Tatsuoki; Matsuura, Toshihiko; Azumi, Masafumi; Kurita, Gen-ichi

    1979-12-01

    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)

  4. Advanced thermionic reactor systems design code

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  5. A bar coding system for environmental projects

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  6. 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.

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

    International Nuclear Information System (INIS)

    Kawai, H.

    2001-01-01

    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)

  8. Performance analysis of super-orthogonal space-frequency trellis coded OFDM system

    CSIR Research Space (South Africa)

    Sokoya, O

    2009-08-01

    Full Text Available that is used with OFDM. SOSFTC-OFDM utilizes the diversities in frequency and space domain by assuming that coding is done along adjacent subcarrier in an OFDM environment. This paper evaluates the exact pairwise error probability (PEP) of the SOSFTC...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-02-01

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

  10. CINETHICA - Core accident analysis code

    International Nuclear Information System (INIS)

    Nakata, H.

    1989-10-01

    A computer program for nuclear accident analysis has been developed based on the point-kinetics approximation and one-dimensional heat transfer model for reactivity feedback calculation. Hansen's method/1/ were used for the kinetics equation solution and explicit Euler method were adopted for the thermohidraulic equations. The results were favorably compared to those from the GAPOTKIN Code/2/. (author) [pt

  11. R-matrix analysis code (RAC)

    International Nuclear Information System (INIS)

    Chen Zhenpeng; Qi Huiquan

    1990-01-01

    A comprehensive R-matrix analysis code has been developed. It is based on the multichannel and multilevel R-matrix theory and runs in VAX computer with FORTRAN-77. With this code many kinds of experimental data for one nuclear system can be fitted simultaneously. The comparisions between code RAC and code EDA of LANL are made. The data show both codes produced the same calculation results when one set of R-matrix parameters was used. The differential cross section of 10 B (n, α) 7 Li for E n = 0.4 MeV and the polarization of 16 O (n,n) 16 O for E n = 2.56 MeV are presented

  12. Analysis of piping systems by finite element method using code SAP-IV

    International Nuclear Information System (INIS)

    Cizelj, L.; Ogrizek, D.

    1987-01-01

    Due to extensive and multiple use of the computer code SAP-IV we have decided to install it on VAX 11/750 machine. Installation required a large quantity of programming due to great discrepancies between the CDC (the original program version) and the VAX. Testing was performed basically in the field of pipe elements, based on a comparison between results obtained with the codes PSAFE2, DOCIJEV, PIPESD and SAP -V. Besides, the model of reactor pressure vessel with 3-D thick shell elements was done. The capabilities show good agreement with the results of other programs mentioned above. Along with the package installation, the graphical postprocessors being developed for mesh plotting. (author)

  13. Containers analysis code of zero order (CACO0) - A basic design system for Type B packages

    International Nuclear Information System (INIS)

    Gaspar, C.; Benito, G.; Rey, J.C.

    1989-01-01

    Very frequently, the principal issues that have to be assessed in the design of a type B(U) package are radiation shielding and evaluation of mechanical and thermal test effects. Thermal behavior during normal transport conditions has also to be considered when the material must dissipate high thermal power. If the transported material is fissile it should be assured that it remains subcritical during transport. The containment of radioactive material must always be assured. In some cases this requires considerable effort. Usually these different design issues are very closely coupled. This coupling does not permit independent consideration. Also, some issues are competitive and generate conflicting design criteria. Given the goal of meeting pertinent transport regulations at a reasonable cost, all design-relevant issues must be balanced in order to obtain a good design. For each design-relevant issue there exists a number of methods of varying efficiency and cost, which can be used to define the key parameters of those particular issues. The overall design methodology must taken into account interactions between parameters of different issues. CACO0 is a system that integrates all design relevant issues and their interactions. The system consists of different modules, each one oriented to a different design issue. The modules are related by a control structure that enables sequentation or iteration during design in a fast and simple manner. Modules can easily be replaced or added, so the system can be updated or adapted to new design problems. The system was designed for use in factibility analysis, cost estimation, conceptual design and initial stages of basic design of type B(U) packages. To accomplish those ends, simple, fast and conservative methods are used

  14. Interrelations of codes in human semiotic systems.

    OpenAIRE

    Somov, Georgij

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

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

    International Nuclear Information System (INIS)

    Lee, Young Jin; Chung, Bub Dong

    2004-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Stevan M. Berber

    2014-06-01

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

  18. System Design Description for the TMAD Code

    International Nuclear Information System (INIS)

    Finfrock, S.H.

    1995-01-01

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

  19. 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)

    2003-11-01

    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

  20. Comprehensive safety analysis code system for nuclear fusion reactors II: Thermal analysis during plasma disruptions for international thermonuclear experimental reactor

    International Nuclear Information System (INIS)

    Honda, T.; Maki, K.; Okazaki, T.

    1994-01-01

    Thermal characteristics of a fusion reactor [International Thermonuclear Experimental Reactor (ITER) Conceptual Design Activity] during plasma disruptions have been analyzed by using a comprehensive safety analysis code for nuclear fusion reactors. The erosion depth due to disruptions for the armor of the first wall depends on the current quench time of disruptions occurring in normal operation. If it is possible to extend the time up to ∼50 ms, the erosion depth is considerably reduced. On the other hand, the erosion depth of the divertor is ∼570 μm for only one disruption, which is determined only by the thermal flux during the thermal quench. This means that the divertor plate should be exchanged after about nine disruptions. Counter-measures are necessary for the divertor to relieve disruption influences. As other scenarios of disruptions, beta-limit disruptions and vertical displacement events were also investigated quantitatively. 13 refs., 5 figs

  1. 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

  2. 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

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

    International Nuclear Information System (INIS)

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

    1988-12-01

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

  4. 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)

    Fernandes Filho, T.L.

    1982-11-01

    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.) [pt

  5. 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)

    Carvalho, F. de A.T. de.

    1985-01-01

    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) [pt

  6. 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.

  7. DEFORM-4: fuel pin characterization and transient response in the SAS4A accident analysis code system

    International Nuclear Information System (INIS)

    Miles, K.J.; Hill, D.J.

    1986-01-01

    The DEFORM-4 module is the segment of the SAS4A Accident Analysis Code System that calculates the fuel pin characterization in response to a steady state irradiation history, thereby providing the initial conditions for the transient calculation. The various phenomena considered include fuel porosity migration, fission gas bubble induced swelling, fuel cracking and healing, fission gas release, cladding swelling, and the thermal-mechanical state of the fuel and cladding. In the transient state, the module continues the thermal-mechanical response calculation, including fuel melting and central cavity pressurization, until cladding failure is predicted and one of the failed fuel modules is initiated. Comparisons with experimental data have demonstrated the validity of the modeling approach

  8. A new balance-of-plant model for the SASSYS-1 LMR [liquid metal reactor] systems analysis code

    International Nuclear Information System (INIS)

    Briggs, L.L.

    1989-01-01

    A balance-of-plant (BOP) model has been developed for use within the SASSYS-1 liquid-metal reactor systems analysis code. This model expands the scope of SASSYS-1 so that the code can explicitly model the waterside components of a nuclear power plant; previously, only the water side of the steam generators could be modeled, with the remainder of the water side represented by boundary conditions on the steam generator. The model represents the BOP a set of flow paths and path junctions; the mass and energy equations are solved at the junctions, and the momentum equation is solved along the flow paths. The junctions are thus mass and energy cells, and the paths are momentum cells. The various waterside component models (pumps, valves, etc.) are specialized types of energy or momentum cells, as appropriate. The solution scheme implicitly couples the energy cells through the momentum cells and solves simultaneously for pressures and enthalpies within the energy cells and for flows within the momentum cells

  9. A COMPARATIVE ANALYSIS OF GLOBAL POSITIONING SYSTEM SCHEMES BASED ON BLOCK CODES

    OpenAIRE

    Prasad Janga; Dr. R. L. Sharma

    2017-01-01

    Global Positioning System (GPS) is a satellite based positioning system based on radio ranging technique. The GPS will provide very accurate three-dimensional position, velocity and timing information to users anywhere in the world. GPS can also be used in other applications such as vehicle monitoring for traffic management in urban areas, Geographical Information System (GIS), 4G Communications, marine navigation, search and rescue and military applications. As GPS accuracy is limited by ion...

  10. NORTICA - a new code for cyclotron analysis

    International Nuclear Information System (INIS)

    Gorelov, D.; Johnson, D.; Marti, F.

    2001-01-01

    The new package NORTICA (Numerical ORbit Tracking In Cyclotrons with Analysis) of computer codes for beam dynamics simulations is under development at NSCL. The package was started as a replacement for the code MONSTER developed in the laboratory in the past. The new codes are capable of beam dynamics simulations in both CCF (Coupled Cyclotron Facility) accelerators, the K500 and K1200 superconducting cyclotrons. The general purpose of this package is assisting in setting and tuning the cyclotrons taking into account the main field and extraction channel imperfections. The computer platform for the package is Alpha Station with UNIX operating system and X-Windows graphic interface. A multiple programming language approach was used in order to combine the reliability of the numerical algorithms developed over the long period of time in the laboratory and the friendliness of modern style user interface. This paper describes the capability and features of the codes in the present state

  11. [Differentiation of coding quality in orthopaedics by special, illustration-oriented case group analysis in the G-DRG System 2005].

    Science.gov (United States)

    Schütz, U; Reichel, H; Dreinhöfer, K

    2007-01-01

    We introduce a grouping system for clinical practice which allows the separation of DRG coding in specific orthopaedic groups based on anatomic regions, operative procedures, therapeutic interventions and morbidity equivalent diagnosis groups. With this, a differentiated aim-oriented analysis of illustrated internal DRG data becomes possible. The group-specific difference of the coding quality between the DRG groups following primary coding by the orthopaedic surgeon and final coding by the medical controlling is analysed. In a consecutive series of 1600 patients parallel documentation and group-specific comparison of the relevant DRG parameters were carried out in every case after primary and final coding. Analysing the group-specific share in the additional CaseMix coding, the group "spine surgery" dominated, closely followed by the groups "arthroplasty" and "surgery due to infection, tumours, diabetes". Altogether, additional cost-weight-relevant coding was necessary most frequently in the latter group (84%), followed by group "spine surgery" (65%). In DRGs representing conservative orthopaedic treatment documented procedures had nearly no influence on the cost weight. The introduced system of case group analysis in internal DRG documentation can lead to the detection of specific problems in primary coding and cost-weight relevant changes of the case mix. As an instrument for internal process control in the orthopaedic field, it can serve as a communicative interface between an economically oriented classification of the hospital performance and a specific problem solution of the medical staff involved in the department management.

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

    International Nuclear Information System (INIS)

    Lee, Kwang Nam; Cho, Nam Zin

    1989-01-01

    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

  13. Fire-accident analysis code (FIRAC) verification

    International Nuclear Information System (INIS)

    Nichols, B.D.; Gregory, W.S.; Fenton, D.L.; Smith, P.R.

    1986-01-01

    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

  14. Implementing a modular system of computer codes

    International Nuclear Information System (INIS)

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

    1983-07-01

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

  15. Plotting system for the MINCS code

    International Nuclear Information System (INIS)

    Watanabe, Tadashi

    1990-08-01

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

  16. 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.

  17. 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.

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

    International Nuclear Information System (INIS)

    Amorim, E.S. de; D'Oliveira, A.B.; Oliveira, E.C. de

    1981-01-01

    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 238 U(n,γ), 238 U(n,f) 239 Pu(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 238 U(n,γ) based on an analysis of cost and economic benefits. (Author) [pt

  19. Structural analysis program of plant piping system. Introduction of AutoPIPE V8i new feature. JSME PPC-class 2 piping code

    International Nuclear Information System (INIS)

    Motohashi, Kazuhiko

    2009-01-01

    After an integration with ADLPipe, AutoPIPE V8i (ver.9.1) became the structural analysis program of plant piping system featured with analysis capability for the ASME NB Class 1 and JSME PPC-Class 2 piping codes including ASME NC Class 2 and ASME ND Class 3. This article described analysis capability for the JSME PPC-Class 2 piping code as well as new general features such as static analysis up to 100 thermal, 10 seismic and 10 wind load cases including different loading scenarios and pipe segment edit function: join, split, reverse and re-order segments. (T. Tanaka)

  20. GIS-assisted spatial analysis for urban regulatory detailed planning: designer's dimension in the Chinese code system

    Science.gov (United States)

    Yu, Yang; Zeng, Zheng

    2009-10-01

    By discussing the causes behind the high amendments ratio in the implementation of urban regulatory detailed plans in China despite its law-ensured status, the study aims to reconcile conflict between the legal authority of regulatory detailed planning and the insufficient scientific support in its decision-making and compilation by introducing into the process spatial analysis based on GIS technology and 3D modeling thus present a more scientific and flexible approach to regulatory detailed planning in China. The study first points out that the current compilation process of urban regulatory detailed plan in China employs mainly an empirical approach which renders it constantly subjected to amendments; the study then discusses the need and current utilization of GIS in the Chinese system and proposes the framework of a GIS-assisted 3D spatial analysis process from the designer's perspective which can be regarded as an alternating processes between the descriptive codes and physical design in the compilation of regulatory detailed planning. With a case study of the processes and results from the application of the framework, the paper concludes that the proposed framework can be an effective instrument which provides more rationality, flexibility and thus more efficiency to the compilation and decision-making process of urban regulatory detailed plan in China.

  1. Code assessment and modelling for Design Basis Accident Analysis of the European sodium fast reactor design. Part I: System description, modelling and benchmarking

    International Nuclear Information System (INIS)

    Lázaro, A.; Ammirabile, L.; Bandini, G.; Darmet, G.; Massara, S.; Dufour, Ph.; Tosello, A.; Gallego, E.; Jimenez, G.; Mikityuk, K.; Schikorr, M.; Bubelis, E.; Ponomarev, A.; Kruessmann, R.; Stempniewicz, M.

    2014-01-01

    Highlights: • Ten system-code models of the ESFR were developed in the frame of the CP-ESFR project. • Eight different thermohydraulic system codes adapted to sodium fast reactor's technology. • Benchmarking exercise settled to check the consistency of the calculations. • Upgraded system codes able to simulate the reactivity feedback and key safety parameters. -- Abstract: The new reactor concepts proposed in the Generation IV International Forum (GIF) are conceived to improve the use of natural resources, reduce the amount of high-level radioactive waste and excel in their reliability and safe operation. Among these novel designs sodium fast reactors (SFRs) stand out due to their technological feasibility as demonstrated in several countries during the last decades. As part of the contribution of EURATOM to GIF the CP-ESFR is a collaborative project with the objective, among others, to perform extensive analysis on safety issues involving renewed SFR demonstrator designs. The verification of computational tools able to simulate the plant behaviour under postulated accidental conditions by code-to-code comparison was identified as a key point to ensure reactor safety. In this line, several organizations employed coupled neutronic and thermal-hydraulic system codes able to simulate complex and specific phenomena involving multi-physics studies adapted to this particular fast reactor technology. In the “Introduction” of this paper the framework of this study is discussed, the second section describes the envisaged plant design and the commonly agreed upon modelling guidelines. The third section presents a comparative analysis of the calculations performed by each organisation applying their models and codes to a common agreed transient with the objective to harmonize the models as well as validating the implementation of all relevant physical phenomena in the different system codes

  2. Code assessment and modelling for Design Basis Accident Analysis of the European sodium fast reactor design. Part I: System description, modelling and benchmarking

    Energy Technology Data Exchange (ETDEWEB)

    Lázaro, A., E-mail: aurelio.lazaro-chueca@ec.europa.eu [JRC-IET European Commission—Westerduinweg 3, PO Box-2, 1755 ZG Petten (Netherlands); UPV—Universidad Politecnica de Valencia, Cami de vera s/n-46002, Valencia (Spain); Ammirabile, L. [JRC-IET European Commission—Westerduinweg 3, PO Box-2, 1755 ZG Petten (Netherlands); Bandini, G. [ENEA, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Darmet, G.; Massara, S. [EDF, 1 avenue du Général de Gaulle, 92141 Clamart (France); Dufour, Ph.; Tosello, A. [CEA, St Paul lez Durance, 13108 Cadarache (France); Gallego, E.; Jimenez, G. [UPM, José Gutiérrez Abascal, 2-28006 Madrid (Spain); Mikityuk, K. [PSI—Paul Scherrer Institut, 5232 Villigen Switzerland (Switzerland); Schikorr, M.; Bubelis, E.; Ponomarev, A.; Kruessmann, R. [KIT—Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen Germany (Germany); Stempniewicz, M. [NRG, Utrechtseweg 310, PO Box 9034 6800 ES, Arnhem (Netherlands)

    2014-01-15

    Highlights: • Ten system-code models of the ESFR were developed in the frame of the CP-ESFR project. • Eight different thermohydraulic system codes adapted to sodium fast reactor's technology. • Benchmarking exercise settled to check the consistency of the calculations. • Upgraded system codes able to simulate the reactivity feedback and key safety parameters. -- Abstract: The new reactor concepts proposed in the Generation IV International Forum (GIF) are conceived to improve the use of natural resources, reduce the amount of high-level radioactive waste and excel in their reliability and safe operation. Among these novel designs sodium fast reactors (SFRs) stand out due to their technological feasibility as demonstrated in several countries during the last decades. As part of the contribution of EURATOM to GIF the CP-ESFR is a collaborative project with the objective, among others, to perform extensive analysis on safety issues involving renewed SFR demonstrator designs. The verification of computational tools able to simulate the plant behaviour under postulated accidental conditions by code-to-code comparison was identified as a key point to ensure reactor safety. In this line, several organizations employed coupled neutronic and thermal-hydraulic system codes able to simulate complex and specific phenomena involving multi-physics studies adapted to this particular fast reactor technology. In the “Introduction” of this paper the framework of this study is discussed, the second section describes the envisaged plant design and the commonly agreed upon modelling guidelines. The third section presents a comparative analysis of the calculations performed by each organisation applying their models and codes to a common agreed transient with the objective to harmonize the models as well as validating the implementation of all relevant physical phenomena in the different system codes.

  3. Development of chemical equilibrium analysis code 'CHEEQ'

    International Nuclear Information System (INIS)

    Nagai, Shuichiro

    2006-08-01

    'CHEEQ' code which calculates the partial pressure and the mass of the system consisting of ideal gas and pure condensed phase compounds, was developed. Characteristics of 'CHEEQ' code are as follows. All the chemical equilibrium equations were described by the formation reactions from the mono-atomic gases in order to simplify the code structure and input preparation. Chemical equilibrium conditions, Σν i μ i =0 for the gaseous compounds and precipitated condensed phase compounds and Σν i μ i > 0 for the non-precipitated condensed phase compounds, were applied. Where, ν i and μ i are stoichiometric coefficient and chemical potential of component i. Virtual solid model was introduced to perform the calculation of constant partial pressure condition. 'CHEEQ' was consisted of following 3 parts, (1) analysis code, zc132. f. (2) thermodynamic data base, zmdb01 and (3) input data file, zindb. 'CHEEQ' code can calculate the system which consisted of elements (max.20), condensed phase compounds (max.100) and gaseous compounds. (max.200). Thermodynamic data base, zmdb01 contains about 1000 elements and compounds, and 200 of them were Actinide elements and their compounds. This report describes the basic equations, the outline of the solution procedure and instructions to prepare the input data and to evaluate the calculation results. (author)

  4. Application of coupled codes for safety analysis and licensing issues

    International Nuclear Information System (INIS)

    Langenbuch, S.; Velkov, K.

    2006-01-01

    An overview is given on the development and the advantages of coupled codes which integrate 3D neutron kinetics into thermal-hydraulic system codes. The work performed within GRS by coupling the thermal-hydraulic system code ATHLET and the 3D neutronics code QUABOX/CUBBOX is described as an example. The application of the coupled codes as best-estimate simulation tools for safety analysis is discussed. Some examples from German licensing practices are given which demonstrate how the improved analytical methods of coupled codes have contributed to solve licensing issues related to optimized and more economical use of fuel. (authors)

  5. 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.

  6. Sandia National Laboratories analysis code data base

    Science.gov (United States)

    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.

  7. Analysis of a small PWR core with the PARCS/Helios and PARCS/Serpent code systems

    International Nuclear Information System (INIS)

    Baiocco, G.; Petruzzi, A.; Bznuni, S.; Kozlowski, T.

    2017-01-01

    Highlights: • The consistency between Helios and Serpent few-group cross sections is shown. • The PARCS model is validated against a Monte Carlo 3D model. • The fission and capture rates are compared. • The influence of the spacer grids on the axial power distribution is shown. - Abstract: Lattice physics codes are primarily used to generate cross-section data for nodal codes. In this work the methodology of homogenized constant generation was applied to a small Pressurized Water Reactor (PWR) core, using the deterministic code Helios and the Monte Carlo code Serpent. Subsequently, a 3D analysis of the PWR core was performed with the nodal diffusion code PARCS using the two-group cross section data sets generated by Helios and Serpent. Moreover, a full 3D model of the PWR core was developed using Serpent in order to obtain a reference solution. Several parameters, such as k eff , axial and radial power, fission and capture rates were compared and found to be in good agreement.

  8. 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.

  9. Web interface for plasma analysis codes

    International Nuclear Information System (INIS)

    Emoto, M.; Murakami, S.; Yoshida, M.; Funaba, H.; Nagayama, Y.

    2008-01-01

    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

  10. Incorporation of a Helical Tube Heat Transfer Model in the MARS Thermal Hydraulic Systems Analysis Code for the T/H Analyses of the SMART Reactor

    International Nuclear Information System (INIS)

    Young Jin Lee; Bub Dong Chung; Jong Chull Jo; Hho Jung Kim; Un Chul Lee

    2004-01-01

    SMART is a medium sized integral type advanced pressurized water reactor currently under development at KAERI. The steam generators of SMART are designed with helically coiled tubes and these are designed to produce superheated steam. The helical shape of the tubes can induce strong centrifugal effect on the secondary coolant as it flows inside the tubes. The presence of centrifugal effect is expected to enhance the formation of cross-sectional circulation flows within the tubes that will increase the overall heat transfer. Furthermore, the centrifugal effect is expected to enhance the moisture separation and thus make it easier to produce superheated steam. MARS is a best-estimate thermal-hydraulic systems analysis code with multi-phase, multi-dimensional analysis capability. The MARS code was produced by restructuring and merging the RELAP5 and the COBRA-TF codes. However, MARS as well as most other best-estimate systems analysis codes in current use lack the detailed models needed to describe the thermal hydraulics of helically coiled tubes. In this study, the heat transfer characteristics and relevant correlations for both the tube and shell sides of helical tubes have been investigated, and the appropriate models have been incorporated into the MARS code. The newly incorporated helical tube heat transfer package is available to the MARS users via selection of the appropriate option in the input. A performance analysis on the steam generator of SMART under full power operation was carried out using the modified MARS code. The results of the analysis indicate that there is a significant improvement in the code predictability. (authors)

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

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

  13. CONTEMPT-DG containment analysis code

    International Nuclear Information System (INIS)

    Deem, R.E.; Rousseau, K.

    1982-01-01

    The assessment of hydrogen burning in a containment building during a degraded core event requires a knowledge of various system responses. These system responses (i.e. heat sinks, fan cooler units, sprays, etc.) can have a marked effect on the overall containment integrity results during a hydrogen burn. In an attempt to properly handle the various system responses and still retain the capability to perform sensitivity analysis on various parameters, the CONTEMPT-DG computer code was developed. This paper will address the historical development of the code, its various features, and the rationale for its development. Comparisons between results from the CONTEMPT-DG analyses and results from similar MARCH analyses will also be given

  14. Error-correction coding and decoding bounds, codes, decoders, analysis and applications

    CERN Document Server

    Tomlinson, Martin; Ambroze, Marcel A; Ahmed, Mohammed; Jibril, Mubarak

    2017-01-01

    This book discusses both the theory and practical applications of self-correcting data, commonly known as error-correcting codes. The applications included demonstrate the importance of these codes in a wide range of everyday technologies, from smartphones to secure communications and transactions. Written in a readily understandable style, the book presents the authors’ twenty-five years of research organized into five parts: Part I is concerned with the theoretical performance attainable by using error correcting codes to achieve communications efficiency in digital communications systems. Part II explores the construction of error-correcting codes and explains the different families of codes and how they are designed. Techniques are described for producing the very best codes. Part III addresses the analysis of low-density parity-check (LDPC) codes, primarily to calculate their stopping sets and low-weight codeword spectrum which determines the performance of these codes. Part IV deals with decoders desi...

  15. Status of SPACE Safety Analysis Code Development

    International Nuclear Information System (INIS)

    Lee, Dong Hyuk; Yang, Chang Keun; Kim, Se Yun; Ha, Sang Jun

    2009-01-01

    In 2006, the Korean the Korean nuclear industry started developing a thermal-hydraulic analysis code for safety analysis of PWR(Pressurized Water Reactor). The new code is named as SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). The SPACE code can solve two-fluid, three-field governing equations in one dimensional or three dimensional geometry. The SPACE code has many component models required for modeling a PWR, such as reactor coolant pump, safety injection tank, etc. The programming language used in the new code is C++, for new generation of engineers who are more comfortable with C/C++ than old FORTRAN language. This paper describes general characteristics of SPACE code and current status of SPACE code development

  16. Expansion of the CHR bone code system

    International Nuclear Information System (INIS)

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

    1976-01-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  18. Coding-Spreading Tradeoff in CDMA Systems

    National Research Council Canada - National Science Library

    Bolas, Eduardo

    2002-01-01

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

  19. Some uncertainty results obtained by the statistical version of the KARATE code system related to core design and safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Panka, Istvan; Hegyi, Gyoergy; Maraczy, Csaba; Temesvari, Emese [Hungarian Academy of Sciences, Budapest (Hungary). Reactor Analysis Dept.

    2017-11-15

    The best-estimate KARATE code system has been widely used for core design calculations and simulations of slow transients of VVER reactors. Recently there has been an increasing need for assessing the uncertainties of such calculations by propagating the basic input uncertainties of the models through the full calculation chain. In order to determine the uncertainties of quantities of interest during the burnup, the statistical version of the KARATE code system has been elaborated. In the first part of the paper, the main features of the new code system are discussed. The applied statistical method is based on Monte-Carlo sampling of the considered input data taking into account mainly the covariance matrices of the cross sections and/or the technological uncertainties. In the second part of the paper, only the uncertainties of cross sections are considered and an equilibrium cycle related to a VVER-440 type reactor is investigated. The burnup dependence of the uncertainties of some safety related parameters (e.g. critical boron concentration, rod worth, feedback coefficients, assembly-wise radial power and burnup distribution) are discussed and compared to the recently used limits.

  20. The coupled code system TORT-TD/ATTICA3D for 3-D transient analysis of pebble-bed HTGR

    International Nuclear Information System (INIS)

    Seubert, A.; Sureda, A.; Lapins, J.; Buck, M.; Laurien, E.; Bader, J.; EnBW Kernkraft GmbH, Philippsburg

    2012-01-01

    This paper describes the time-dependent 3-D discrete-ordinates based coupled code system TORT-TD/ATTICA3D and its application to HTGR of pebble bed type. TORT-TD/ATTICA3D is represented by a single executable and adapts the so-called internal coupling approach. Three-dimensional distributions of temperatures from ATTICA3D and power density from TORT-TD are efficiently exchanged by direct memory access of array elements via interface routines. Applications of TORT-TD/ATTICA3D to three transients based on the PBMR-400 benchmark (total and partial control rod withdrawal and cold helium ingress) and the full power steady state of the HTR-10 are presented. For the partial control rod withdrawal, 3-D effects of local neutron flux redistributions are clearly identified. The results are very promising and demonstrate that the coupled code system TORT-TD/ATTICA3D may represent a key component in a future comprehensive 3-D code system for HTGR of pebble bed type. (orig.)

  1. Integrated burnup calculation code system SWAT

    International Nuclear Information System (INIS)

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

    1997-11-01

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

  2. The octopus burnup and criticality code system

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-09-01

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

  3. The OCTOPUS burnup and criticality code system

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-06-01

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

  4. The octopus burnup and criticality code system

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  5. The OCTOPUS burnup and criticality code system

    International Nuclear Information System (INIS)

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

    1996-06-01

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

  6. Development and assessment of Multi-dimensional flow models in the thermal-hydraulic system analysis code MARS

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Bae, S. W.; Jeong, J. J.; Lee, S. M

    2005-04-15

    A new multi-dimensional component has been developed to allow for more flexible 3D capabilities in the system code, MARS. This component can be applied in the Cartesian and cylindrical coordinates. For the development of this model, the 3D convection and diffusion terms are implemented in the momentum and energy equation. And a simple Prandtl's mixing length model is applied for the turbulent viscosity. The developed multi-dimensional component was assessed against five conceptual problems with analytic solution. And some SETs are calculated and compared with experimental data. With this newly developed multi-dimensional flow module, the MARS code can realistic calculate the flow fields in pools such as those occurring in the core, steam generators and IRWST.

  7. Development and assessment of Multi-dimensional flow models in the thermal-hydraulic system analysis code MARS

    International Nuclear Information System (INIS)

    Chung, B. D.; Bae, S. W.; Jeong, J. J.; Lee, S. M.

    2005-04-01

    A new multi-dimensional component has been developed to allow for more flexible 3D capabilities in the system code, MARS. This component can be applied in the Cartesian and cylindrical coordinates. For the development of this model, the 3D convection and diffusion terms are implemented in the momentum and energy equation. And a simple Prandtl's mixing length model is applied for the turbulent viscosity. The developed multi-dimensional component was assessed against five conceptual problems with analytic solution. And some SETs are calculated and compared with experimental data. With this newly developed multi-dimensional flow module, the MARS code can realistic calculate the flow fields in pools such as those occurring in the core, steam generators and IRWST

  8. Uncertainty analysis of the FRAP code

    International Nuclear Information System (INIS)

    Peck, S.O.

    1978-01-01

    A user oriented, automated uncertainty analysis capability has been built into the FRAP code (Fuel Rod Analysis Program) and applied to a PWR fuel rod undergoing a LOCA. The method of uncertainty analysis is the Response Surface Method (RSM). (author)

  9. Code systems for effective and precise calculation of the basic neutron characteristics, core loading optimization, analysis and estimation of the operation regimes of WWER type reactors

    International Nuclear Information System (INIS)

    Apostolov, T.; Ivanov, K.; Prodanova, R.; Manolova, M.; Petrova, T.; Alekova, G.

    1993-01-01

    Two directions for investigations are suggested: 1) Analysis and evaluation of the real loading patterns and operational regimes for Kozloduy NPP WWER-440 and WWER-1000 in the frame of the recent safety criteria and nuclear power plant operating limits. 2) Development of modern code system for WWER type reactor core analysis with advanced features: new design and materials for fuel and control rods, increasing the fuel enrichment, using the integral and discrete burnable absorbers etc. The fuel technology design evolution maximizes the fuel utilization efficiency, improves operation performance and enhances safety margins. By the joint efforts of specialists from INRNE, Sofia (BG) and KAB, Berlin (GE), the codes NESSEL-IV-EC, PYTHIA and DERAB have been developed and verified. In the frame of the PHARE programme the joint project ASPERCA has been proposed intended for reactor physics calculations with PHYBER-WWER code for safety enhancement and operation reliability improvement. In-core fuel management benchmarks for 4 cycles of unit 2 (WWER-440) and 2 cycles of unit 5 (WWER-1000) have been performed. The coordination of burnable absorber design implementation, low leakage loadings usage, reloading enrichment increase and steel content reduction in the core have made the reactor core analysis more demanding and the definition of loading patterns - more difficult. This complexity requires routine use of three-dimensional fast accurate core model with extended and updated cross section libraries. To meet the needs of WWER advanced loading patterns and in-core fuel management improvements the HEXANES code systems is being developed and qualified. Some test calculations have been carried out by the HEXANES code system investigating the influence of Gd in the fuel on the main reactor physics parameters. For reevaluation of the core safety-related design limits forming the basis of licensing procedure, the code DYN3D/M2 is used. 16 refs., 3 figs. (author)

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

    International Nuclear Information System (INIS)

    Li Chunhuai; Fu Shouxin; Liu Guilian

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    Forster, R.A.; Parsons, K.

    1997-01-01

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

  12. High dynamic range coding imaging system

    Science.gov (United States)

    Wu, Renfan; Huang, Yifan; Hou, Guangqi

    2014-10-01

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

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

    Science.gov (United States)

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

    2007-04-01

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

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

    Science.gov (United States)

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

    2016-03-01

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

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

    International Nuclear Information System (INIS)

    Hwang, Moon Kyu; Kim, Soo Hyung; Kim, Byung Jae; Chung, Bub Dong; Kim, Hee Cheol

    2010-04-01

    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)

  16. Channel coding in the space station data system network

    Science.gov (United States)

    Healy, T.

    1982-01-01

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

  17. 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.

  18. OPR1000 RCP Flow Coastdown Analysis using SPACE Code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong-Hyuk; Kim, Seyun [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The Korean nuclear industry developed a thermal-hydraulic analysis code for the safety analysis of PWRs, named SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). Current loss of flow transient analysis of OPR1000 uses COAST code to calculate transient RCS(Reactor Coolant System) flow. The COAST code calculates RCS loop flow using pump performance curves and RCP(Reactor Coolant Pump) inertia. In this paper, SPACE code is used to reproduce RCS flowrates calculated by COAST code. The loss of flow transient is transient initiated by reduction of forced reactor coolant circulation. Typical loss of flow transients are complete loss of flow(CLOF) and locked rotor(LR). OPR1000 RCP flow coastdown analysis was performed using SPACE using simplified nodalization. Complete loss of flow(4 RCP trip) was analyzed. The results show good agreement with those from COAST code, which is CE code for calculating RCS flow during loss of flow transients. Through this study, we confirmed that SPACE code can be used instead of COAST code for RCP flow coastdown analysis.

  19. SRAC2006: A comprehensive neutronics calculation code system

    International Nuclear Information System (INIS)

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

    2007-02-01

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

  20. Application of ASTEC, MELCOR, and MAAP Computer Codes for Thermal Hydraulic Analysis of a PWR Containment Equipped with the PCFV and PAR Systems

    Directory of Open Access Journals (Sweden)

    Siniša Šadek

    2017-01-01

    Full Text Available The integrity of the containment will be challenged during a severe accident due to pressurization caused by the accumulation of steam and other gases and possible ignition of hydrogen and carbon monoxide. Installation of a passive filtered venting system and passive autocatalytic recombiners allows control of the pressure, radioactive releases, and concentration of flammable gases. Thermal hydraulic analysis of the containment equipped with dedicated passive safety systems after a hypothetical station blackout event is performed for a two-loop pressurized water reactor NPP with three integral severe accident codes: ASTEC, MELCOR, and MAAP. MELCOR and MAAP are two major US codes for severe accident analyses, and the ASTEC code is the European code, joint property of Institut de Radioprotection et de Sûreté Nucléaire (IRSN, France and Gesellschaft für Anlagen und Reaktorsicherheit (GRS, Germany. Codes’ overall characteristics, physics models, and the analysis results are compared herein. Despite considerable differences between the codes’ modelling features, the general trends of the NPP behaviour are found to be similar, although discrepancies related to simulation of the processes in the containment cavity are also observed and discussed in the paper.

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

    International Nuclear Information System (INIS)

    Mohd Faiz Salim; Ridha Roslan; Mohd Rizal Mamat

    2013-01-01

    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)

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

    International Nuclear Information System (INIS)

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

    2014-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-02-12

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

  4. Steady state and LOCA analysis of Kartini reactor using RELAP5/SCDAP code: The role of passive system

    Science.gov (United States)

    Antariksawan, Anhar R.; Wahyono, Puradwi I.; Taxwim

    2018-02-01

    Safety is the priority for nuclear installations, including research reactors. On the other hand, many studies have been done to validate the applicability of nuclear power plant based best estimate computer codes to the research reactor. This study aims to assess the applicability of the RELAP5/SCDAP code to Kartini research reactor. The model development, steady state and transient due to LOCA calculations have been conducted by using RELAP5/SCDAP. The calculation results are compared with available measurements data from Kartini research reactor. The results show that the RELAP5/SCDAP model steady state calculation agrees quite well with the available measurement data. While, in the case of LOCA transient simulations, the model could result in reasonable physical phenomena during the transient showing the characteristics and performances of the reactor against the LOCA transient. The role of siphon breaker hole and natural circulation in the reactor tank as passive system was important to keep reactor in safe condition. It concludes that the RELAP/SCDAP could be use as one of the tool to analyse the thermal-hydraulic safety of Kartini reactor. However, further assessment to improve the model is still needed.

  5. 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)

    Pham Van Lam; Le Vinh Vinh; Huynh Ton Nghiem

    2004-01-01

    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)

  6. Extending CANTUP code analysis to probabilistic evaluations

    International Nuclear Information System (INIS)

    Florea, S.

    2001-01-01

    The structural analysis with numerical methods based on final element method plays at present a central role in evaluations and predictions of structural systems which require safety and reliable operation in aggressive environmental conditions. This is the case too for the CANDU - 600 fuel channel, where besides the corrosive and thermal aggression upon the Zr97.5Nb2.5 pressure tubes, a lasting irradiation adds which has marked consequences upon the materials properties evolution. This results in an unavoidable spreading in the materials properties in time, affected by high uncertainties. Consequently, the deterministic evaluation with computation codes based on finite element method are supplemented by statistic and probabilistic methods of evaluation of the response of structural components. This paper reports the works on extending the thermo-mechanical evaluation of the fuel channel components in the frame of probabilistic structure mechanics based on statistical methods and developed upon deterministic CANTUP code analyses. CANTUP code was adapted from LAHEY 77 platform onto Microsoft Developer Studio - Fortran Power Station 4.0 platform. To test the statistical evaluation of the creeping behaviour of pressure tube, the value of longitudinal elasticity modulus (Young) was used, as random variable, with a normal distribution around value, as used in deterministic analyses. The influence of the random quantity upon the hog and effective stress developed in the pressure tube for to time values, specific to primary and secondary creep was studied. The results obtained after a five year creep, corresponding to the secondary creep are presented

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

    Science.gov (United States)

    Sikder, Somali; Ghosh, Shila

    2018-02-01

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

  8. Developments of HTGR thermofluid dynamic analysis codes and HTGR plant dynamic simulation code

    International Nuclear Information System (INIS)

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

    1983-01-01

    In nuclear power plants as well as high temperature gas-cooled reactor plants, the design is mostly performed on the basis of the results after their characteristics have been grasped by carrying out the numerical simulation using the analysis code. Also in Kawasaki Heavy Industries Ltd., on the basis of the system engineering accumulated with gas-cooled reactors since several years ago, the preparation and systematization of analysis codes have been advanced, aiming at lining up the analysis codes for heat transferring flow and control characteristics, taking up HTGR plants as the main object. In this report, a part of the results is described. The example of the analysis applying the two-dimensional compressible flow analysis codes SOLA-VOF and SALE-2D, which were developed by Los Alamos National Laboratory in USA and modified for use in Kawasaki, to HTGR system is reported. Besides, Kawasaki has developed the control characteristics analyzing code DYSCO by which the change of system composition is easy and high versatility is available. The outline, fundamental equations, fundamental algorithms and examples of application of the SOLA-VOF and SALE-2D, the present status of system characteristic simulation codes and the outline of the DYSCO are described. (Kako, I.)

  9. Current lead thermal analysis code 'CURRENT'

    International Nuclear Information System (INIS)

    Yamaguchi, Masahito; Tada, Eisuke; Shimamoto, Susumu; Hata, Kenichiro.

    1985-08-01

    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)

  10. Comprehensive safety analysis code system for nuclear fusion reactors III: Ex-vessel LOCA analyses considering passive safety

    International Nuclear Information System (INIS)

    Honda, T.; Okazaki, T.; Maki, K.; Uda, T.; Seki, Y.; Aoki, I.; Kunugi, T.

    1996-01-01

    Ex-vessel loss-of-coolant accidents (LOCAs) in a fusion reactor have been analyzed to investigate the possibility of passive plasma shutdown. For this purpose, a hybrid code of the plasma dynamics and thermal characteristics of the reactor structures, which has been modified to include the impurity emission from plasma-facing components (PFCs), has been developed. Ex-vessel LOCAs of the cooling system during the ignition operation in the International Thermonuclear Experimental Reactor (ITER), in which graphite PFCs were employed in conceptual design activity, were assumed. When double-ended break occurs at the cold leg of the divertor cooling system, the copper cooling tube begins to melt within 3 s after the LOCA, even though the plasma is passively shut down at nearly 4 s. An active plasma shutdown system will be needed for such rapid transient accidents. On the other hand, when a small (1%) break LOCA occurs there, the plasma is passively shut down at nearly 36 s, which happens before the copper cooling tube begins to melt. When the double-ended break LOCA occurs at the cold leg of the first-wall cooling system, there is enough time (nearly 100 s) to shut down the plasma with a controllable method before the reactor structures are damaged. 21 refs., 8 figs

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

    International Nuclear Information System (INIS)

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

    2010-02-01

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

  13. Development of the coupled 'system thermal-hydraulics, 3D reactor kinetics, and hot channel' analysis capability of the MARS code

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, J. J.; Chung, B. D.; Lee, W.J

    2005-02-01

    The subchannel analysis capability of the MARS 3D module has been improved. Especially, the turbulent mixing and void drift models for flow mixing phenomena in rod bundles have been assessed using some well-known rod bundle test data. Then, the subchannel analysis feature was combined to the existing coupled 'system Thermal-Hydraulics (T/H) and 3D reactor kinetics' calculation capability of MARS. These features allow the coupled 'system T/H, 3D reactor kinetics, and hot channel' analysis capability and, thus, realistic simulations of hot channel behavior as well as global system T/H behavior. In this report, the MARS code features for the coupled analysis capability are described first. The code modifications relevant to the features are also given. Then, a coupled analysis of the Main Steam Line Break (MSLB) is carried out for demonstration. The results of the coupled calculations are very reasonable and realistic, and show these methods can be used to reduce the over-conservatism in the conventional safety analysis.

  14. Long non-coding RNA XIST predicts worse prognosis in digestive system tumors: a systemic review and meta-analysis.

    Science.gov (United States)

    Liu, Xuefang; Ming, Xinliang; Jing, Wei; Luo, Ping; Li, Nandi; Zhu, Man; Yu, Mingxia; Liang, Chunzi; Tu, Jiancheng

    2018-05-11

    Aims: Increasing studies are indicating that long non-coding RNA X-inactive specific transcript (XIST) is associated with the prognosis of cancer patients. However, the results have been disputed. Therefore, we aimed to further explore the prognostic value and clinical significance of XIST in various types of cancers. Then, we focused our research on the comparison of the predictive value of XIST between digestive system tumors and non-digestive system tumors. Methods: We performed a systematic search by looking up PubMed, Embase, Cochrane Library, Web of Science, and Medline (up to January 3, 2018). Fifteen studies which matched to our inclusion criteria with a total of 920 patients for overall survival and 867 patients for clinicopathological characteristics were included in this meta-analysis. Pooled hazard ratios (HR) and odds ratios (ORs) with their corresponding 95% confidence intervals (95% CIs) were calculated to summarize the effects. Results: Our results suggested that high expression levels of XIST were associated with unfavorable overall survival in cancer patients (pooled HR=1.81, 95% CI: 1.45-2.26). Additionally, we found that XIST was more valuable in digestive system tumors (pooled HR=2.24, 95% CI: 1.73-2.92) than in non-digestive system tumors (pooled HR=1.22, 95% CI: 0.60-2.45). Additionally, elevated expression levels of XIST were connection with distant metastasis and tumor stage. Conclusion: XIST was correlated with poor prognosis, which suggested that XIST might serve as a novel predictive biomarker for cancer patients, especially for patients of digestive system tumors. ©2018 The Author(s).

  15. ENSDF analysis codes. IBM version, August 1982

    International Nuclear Information System (INIS)

    Lorenz, A.

    1982-01-01

    The nuclear structure analysis programme tape consists of physics computer processing codes used in the evaluation of mass-chain structure data. This tape was generated by the National Nuclear Data Centre, Brookhaven National Laboratory in the USA. (author)

  16. ENSDF analysis codes: IBM version. August 1982

    International Nuclear Information System (INIS)

    Lorenz, A.

    1982-09-01

    The nuclear structure analysis programme tape consists of physics computer processing codes used in the evaluation of mass-chain structure data. This tape was generated by the National Nuclear Data Centre, Brookhaven National Laboratory in the USA. (author)

  17. Analysis and Compensation for Lateral Chromatic Aberration in a Color Coding Structured Light 3D Measurement System

    Science.gov (United States)

    Huang, Junhui; Xue, Qi; Wang, Zhao; Gao, Jianmin

    2016-01-01

    While color-coding methods have improved the measuring efficiency of a structured light three-dimensional (3D) measurement system, they decreased the measuring accuracy significantly due to lateral chromatic aberration (LCA). In this study, the LCA in a structured light measurement system is analyzed, and a method is proposed to compensate the error caused by the LCA. Firstly, based on the projective transformation, a 3D error map of LCA is constructed in the projector images by using a flat board and comparing the image coordinates of red, green and blue circles with the coordinates of white circles at preselected sample points within the measurement volume. The 3D map consists of the errors, which are the equivalent errors caused by LCA of the camera and projector. Then in measurements, error values of LCA are calculated and compensated to correct the projector image coordinates through the 3D error map and a tri-linear interpolation method. Eventually, 3D coordinates with higher accuracy are re-calculated according to the compensated image coordinates. The effectiveness of the proposed method is verified in the following experiments. PMID:27598174

  18. Analysis of quantum error-correcting codes: Symplectic lattice codes and toric codes

    Science.gov (United States)

    Harrington, James William

    Quantum information theory is concerned with identifying how quantum mechanical resources (such as entangled quantum states) can be utilized for a number of information processing tasks, including data storage, computation, communication, and cryptography. Efficient quantum algorithms and protocols have been developed for performing some tasks (e.g. , factoring large numbers, securely communicating over a public channel, and simulating quantum mechanical systems) that appear to be very difficult with just classical resources. In addition to identifying the separation between classical and quantum computational power, much of the theoretical focus in this field over the last decade has been concerned with finding novel ways of encoding quantum information that are robust against errors, which is an important step toward building practical quantum information processing devices. In this thesis I present some results on the quantum error-correcting properties of oscillator codes (also described as symplectic lattice codes) and toric codes. Any harmonic oscillator system (such as a mode of light) can be encoded with quantum information via symplectic lattice codes that are robust against shifts in the system's continuous quantum variables. I show the existence of lattice codes whose achievable rates match the one-shot coherent information over the Gaussian quantum channel. Also, I construct a family of symplectic self-dual lattices and search for optimal encodings of quantum information distributed between several oscillators. Toric codes provide encodings of quantum information into two-dimensional spin lattices that are robust against local clusters of errors and which require only local quantum operations for error correction. Numerical simulations of this system under various error models provide a calculation of the accuracy threshold for quantum memory using toric codes, which can be related to phase transitions in certain condensed matter models. I also present

  19. SCALE Code System 6.2.1

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-01

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

  20. SCALE Code System 6.2.2

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-05-01

    The SCALE Code System is a widely used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including 3 deterministic and 3 Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results. SCALE 6.2 represents one of the most comprehensive revisions in the history of SCALE, providing several new capabilities and significant improvements in many existing features.

  1. SCALE Code System 6.2.1

    International Nuclear Information System (INIS)

    Rearden, Bradley T.; Jessee, Matthew Anderson

    2016-01-01

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

  2. 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)

    Bae, Byoung-Uhn; Kim, Seok; Park, Yu-Sun; Kang, Kyoung-Ho

    2014-01-01

    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

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

    International Nuclear Information System (INIS)

    Hanawa, Satoshi; Iyoku, Tatsuo

    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)

  4. ETF system code: composition and applications

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  5. A compendium of computer codes in fault tree analysis

    International Nuclear Information System (INIS)

    Lydell, B.

    1981-03-01

    In the past ten years principles and methods for a unified system reliability and safety analysis have been developed. Fault tree techniques serve as a central feature of unified system analysis, and there exists a specific discipline within system reliability concerned with the theoretical aspects of fault tree evaluation. Ever since the fault tree concept was established, computer codes have been developed for qualitative and quantitative analyses. In particular the presentation of the kinetic tree theory and the PREP-KITT code package has influenced the present use of fault trees and the development of new computer codes. This report is a compilation of some of the better known fault tree codes in use in system reliability. Numerous codes are available and new codes are continuously being developed. The report is designed to address the specific characteristics of each code listed. A review of the theoretical aspects of fault tree evaluation is presented in an introductory chapter, the purpose of which is to give a framework for the validity of the different codes. (Auth.)

  6. EXTRA·M: a computing code system for analysis of the Purex process with mixer settlers for reprocessing

    International Nuclear Information System (INIS)

    Tachimori, Shoichi

    1994-03-01

    A computer code system EXTRA·M, for simulation of transient behavior of the solutes in a multistage countercurrent extraction process, was developed aiming to predict the distribution and chemical behaviors of actinide elements, i.e., U, Pu, Np, and of technetium in the Purex process of fuel reprocessing. The mathematical model is applicable to a complete mixing stagewise contactor such as mixer settler and to the Purex, with tri-n-butylphosphate (TBP) and nitric acid system. The main characteristics of the EXTRA·M are as follows; i) Calculation of distribution ratios of the solutes is based on numerical equations of which parameter values are to be determined by a best fit method with a number of experimental data. ii) Total of 18 solutes; U(IV), U(VI), Pu(III), Pu(IV), Pu(V), Pu(VI), Np(IV), Np(V), Np(VI), Tc(IV), Tc(V), Tc(VI), Tc(VII), Zr(IV), HNO 3 , hydrazine, hydroxylamine nitrate and nitrous acid, are treated and rate equations of total 40 chemical reactions involving these solutes are incorporated. iii) Instantaneous change of flow conditions, i.e., concentration of the solutes and flow rate of the feeding solutions, is contrived by computation. iv) Reflux or bypass mode calculation, in which an aqueous raffinate stream is transferred to the preceding bank or stage, is possible. The present report explains the concept, assumptions and characteristics of the model, the material balance equations including distribution and reaction rate equations and their solution method, and the usefulness of the model by showing some examples of the verification results. A description and source program of EXTRA·M1, as an example, are listed in the annex. (J.P.N.) 63 refs

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

    International Nuclear Information System (INIS)

    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

  8. 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.

  9. Code system to compute radiation dose in human phantoms

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  10. Arabic Natural Language Processing System Code Library

    Science.gov (United States)

    2014-06-01

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

  11. Development of a Performance Analysis Code for the Off-design conditions of a S-CO2 Brayton Cycle Energy Conversion System

    International Nuclear Information System (INIS)

    Yoo, Yong-Hwan; Cha, Jae-Eun; Lee, Tae-Ho; Eoh, Jae-Hyuk; Kim, Seong-O

    2008-01-01

    For the development of a supercritical carbon dioxide (S-CO2) Brayton cycle energy conversion system coupled to KALIMER-600, a thermal balance has been established on 100% power operating conditions including all the reactor system models such as a primary heat transport system (PHTS), an intermediate heat transport system (IHTS), and an energy conversion system. The S-CO2 Brayton cycle energy conversion system consists of a sodium-CO2 heat exchanger (Hx), turbine, high temperature recuperate (HTR), low temperature recuperate (LTR), precooler, compressor no.1, and compressor no.2. Two compressors were employed to avoid a sharp change of the physical properties near their critical point with a corresponding pressure. The component locations and their operating conditions are illustrated. Energy balance of the power conversion system in KALIMER-600 was designed with the full power condition of each component. Therefore, to predict the off-design conditions and to evaluate each component, an off-design performance analysis code should be accomplished. An off-design performance analysis could be classified into overall system control logic and local system control logic. The former means that mass flow rate and power are controlled by valves, and the latter implies that a bypass or inventory control is an admitted system balance. The ultimate goal of this study is development of the overall system control logic

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

    International Nuclear Information System (INIS)

    Baratta, A.J.

    1997-01-01

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

  13. 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.

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

    Science.gov (United States)

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

    2016-11-01

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

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

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Lee, Won Jae; Jeong, Jae Jun; Ha, Kwi Seok

    1998-01-01

    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

  16. Analysis of PWR control rod ejection accident with the coupled code system SKETCH-INS/TRACE by incorporating pin power reconstruction model

    International Nuclear Information System (INIS)

    Nakajima, T.; Sakai, T.

    2010-01-01

    The pin power reconstruction model was incorporated in the 3-D nodal kinetics code SKETCH-INS in order to produce accurate calculation of three-dimensional pin power distributions throughout the reactor core. In order to verify the employed pin power reconstruction model, the PWR MOX/UO_2 core transient benchmark problem was analyzed with the coupled code system SKETCH-INS/TRACE by incorporating the model and the influence of pin power reconstruction model was studied. SKETCH-INS pin power distributions for 3 benchmark problems were compared with the PARCS solutions which were provided by the host organisation of the benchmark. SKETCH-INS results were in good agreement with the PARCS results. The capability of employed pin power reconstruction model was confirmed through the analysis of benchmark problems. A PWR control rod ejection benchmark problem was analyzed with the coupled code system SKETCH-INS/ TRACE by incorporating the pin power reconstruction model. The influence of pin power reconstruction model was studied by comparing with the result of conventional node averaged flux model. The results indicate that the pin power reconstruction model has significant effect on the pin powers during transient and hence on the fuel enthalpy

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

    International Nuclear Information System (INIS)

    Jeong, Ji Hwan; Kim, Sang Jae; Chang, Keun Sun; Lee, Jae Hun

    2001-01-01

    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

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

    Institute of Scientific and Technical Information of China (English)

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

    2017-01-01

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

  19. Study of nuclear computer code maintenance and management system

    International Nuclear Information System (INIS)

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

    1989-01-01

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

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

    International Nuclear Information System (INIS)

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

    1982-12-01

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

  1. Fault tree analysis. Implementation of the WAM-codes

    International Nuclear Information System (INIS)

    Bento, J.P.; Poern, K.

    1979-07-01

    The report describes work going on at Studsvik at the implementation of the WAM code package for fault tree analysis. These codes originally developed under EPRI contract by Sciences Applications Inc, allow, in contrast with other fault tree codes, all Boolean operations, thus allowing modeling of ''NOT'' conditions and dependent components. To concretize the implementation of these codes, the auxiliary feed-water system of the Swedish BWR Oskarshamn 2 was chosen for the reliability analysis. For this system, both the mean unavailability and the probability density function of the top event - undesired event - of the system fault tree were calculated, the latter using a Monte-Carlo simulation technique. The present study is the first part of a work performed under contract with the Swedish Nuclear Power Inspectorate. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

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

    International Nuclear Information System (INIS)

    Yokobayashi, Masao; Yoshida, Kazuo; Fujiki, Kazuo

    1989-06-01

    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)

  4. 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.

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

    International Nuclear Information System (INIS)

    Ionescu-Bujor, M.

    2000-10-01

    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.) [de

  6. Sensitivity analysis of the RESRAD, a dose assessment code

    International Nuclear Information System (INIS)

    Yu, C.; Cheng, J.J.; Zielen, A.J.

    1991-01-01

    The RESRAD code is a pathway analysis code that is designed to calculate radiation doses and derive soil cleanup criteria for the US Department of Energy's environmental restoration and waste management program. the RESRAD code uses various pathway and consumption-rate parameters such as soil properties and food ingestion rates in performing such calculations and derivations. As with any predictive model, the accuracy of the predictions depends on the accuracy of the input parameters. This paper summarizes the results of a sensitivity analysis of RESRAD input parameters. Three methods were used to perform the sensitivity analysis: (1) Gradient Enhanced Software System (GRESS) sensitivity analysis software package developed at oak Ridge National Laboratory; (2) direct perturbation of input parameters; and (3) built-in graphic package that shows parameter sensitivities while the RESRAD code is operational

  7. MELCOR Accident Consequence Code System (MACCS)

    International Nuclear Information System (INIS)

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

    1990-02-01

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

  8. MELCOR Accident Consequence Code System (MACCS)

    International Nuclear Information System (INIS)

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

    1990-02-01

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

  9. 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.

  10. Code Coupling for Multi-Dimensional Core Transient Analysis

    International Nuclear Information System (INIS)

    Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il

    2015-01-01

    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. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    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)

  13. Analysis of pipe stress using CAESAR II code

    International Nuclear Information System (INIS)

    Sitandung, Y.B.; Bandriyana, B.

    2002-01-01

    Analysis of this piping stress with the purpose of knowing stress distribution piping system in order to determine pipe supports configuration. As an example of analysis, Gas Exchanger to Warm Separator Line was chosen with, input data was firstly prepared in a document, i.e. piping analysis specification that its content named as pipe characteristics, material properties, operation conditions, guide equipment's and so on. Analysis result such as stress, load, displacement and the use support type were verified based on requirements in the code, standard, and regularities were suitable with piping system condition analyzed. As the proof that piping system is in safety condition, it can be indicated from analysis results (actual loads) which still under allowable load. From the analysis steps that have been done CAESAR II code fulfill requirements to be used as a tool of piping stress analysis as well as nuclear and non nuclear installation piping system

  14. Fast decoding algorithms for coded aperture systems

    International Nuclear Information System (INIS)

    Byard, Kevin

    2014-01-01

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

  15. CATHENA 4. A thermalhydraulics network analysis code

    International Nuclear Information System (INIS)

    Aydemir, N.U.; Hanna, B.N.

    2009-01-01

    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)

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

    International Nuclear Information System (INIS)

    Bjerke, M.A.

    1983-02-01

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

  17. 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.

  18. 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...

  19. Turbo Pascal Computer Code for PIXE Analysis

    International Nuclear Information System (INIS)

    Darsono

    2002-01-01

    To optimal utilization of the 150 kV ion accelerator facilities and to govern the analysis technique using ion accelerator, the research and development of low energy PIXE technology has been done. The R and D for hardware of the low energy PIXE installation in P3TM have been carried on since year 2000. To support the R and D of PIXE accelerator facilities in harmonize with the R and D of the PIXE hardware, the development of PIXE software for analysis is also needed. The development of database of PIXE software for analysis using turbo Pascal computer code is reported in this paper. This computer code computes the ionization cross-section, the fluorescence yield, and the stopping power of elements also it computes the coefficient attenuation of X- rays energy. The computer code is named PIXEDASIS and it is part of big computer code planed for PIXE analysis that will be constructed in the near future. PIXEDASIS is designed to be communicative with the user. It has the input from the keyboard. The output shows in the PC monitor, which also can be printed. The performance test of the PIXEDASIS shows that it can be operated well and it can provide data agreement with data form other literatures. (author)

  20. Uncertainty analysis of the FRAP code

    International Nuclear Information System (INIS)

    Peck, S.O.

    1978-01-01

    A user oriented, automated uncertainty analysis capability has been built into the Fuel Rod Analysis Program (FRAP) code and has been applied to a pressurized water reactor (PWR) fuel rod undergoing a loss-of-coolant accident (LOCA). The method of uncertainty analysis is the response surface method. The automated version significantly reduced the time required to complete the analysis and, at the same time, greatly increased the problem scope. Results of the analysis showed a significant difference in the total and relative contributions to the uncertainty of the response parameters between steady state and transient conditions

  1. Summary description of the scale modular code system

    International Nuclear Information System (INIS)

    Parks, C.V.

    1987-12-01

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

  2. 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

    Millimeter wave communications based on photonic technologies have gained increased attention to provide optic fiber-like capacity in wireless environments. However, the new hybrid fiber-wireless channel represents new challenges in terms of signal transmission performance analysis. Traditionally......, such systems use diversity schemes in combination with digital signal processing (DSP) techniques to overcome effects such as fading and inter-symbol interference (ISI). Wavelet Channel Coding (WCC) has emerged as a technique to minimize the fading effects of wireless channels, which is a mayor challenge...... in systems operating in the millimeter wave regime. This work takes the WCC one step beyond by performance evaluation in terms of bit error probability, over time-varying, frequency-selective multipath Rayleigh fading channels. The adopted propagation model follows the COST207 norm, the main international...

  3. Design and performance analysis for several new classes of codes for optical synchronous CDMA and for arbitrary-medium time-hopping synchronous CDMA communication systems

    Science.gov (United States)

    Kostic, Zoran; Titlebaum, Edward L.

    1994-08-01

    New families of spread-spectrum codes are constructed, that are applicable to optical synchronous code-division multiple-access (CDMA) communications as well as to arbitrary-medium time-hopping synchronous CDMA communications. Proposed constructions are based on the mappings from integer sequences into binary sequences. We use the concept of number theoretic quadratic congruences and a subset of Reed-Solomon codes similar to the one utilized in the Welch-Costas frequency-hop (FH) patterns. The properties of the codes are as good as or better than the properties of existing codes for synchronous CDMA communications: Both the number of code-sequences within a single code family and the number of code families with good properties are significantly increased when compared to the known code designs. Possible applications are presented. To evaluate the performance of the proposed codes, a new class of hit arrays called cyclical hit arrays is recalled, which give insight into the previously unknown properties of the few classes of number theoretic FH patterns. Cyclical hit arrays and the proposed mappings are used to determine the exact probability distribution functions of random variables that represent interference between users of a time-hopping or optical CDMA system. Expressions for the bit error probability in multi-user CDMA systems are derived as a function of the number of simultaneous CDMA system users, the length of signature sequences and the threshold of a matched filter detector. The performance results are compared with the results for some previously known codes.

  4. 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...

  5. NALAP: an LMFBR system transient code

    International Nuclear Information System (INIS)

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

    1975-07-01

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

  6. Benchmark calculation of subchannel analysis codes

    International Nuclear Information System (INIS)

    1996-02-01

    In order to evaluate the analysis capabilities of various subchannel codes used in thermal-hydraulic design of light water reactors, benchmark calculations were performed. The selected benchmark problems and major findings obtained by the calculations were as follows: (1)As for single-phase flow mixing experiments between two channels, the calculated results of water temperature distribution along the flow direction were agreed with experimental results by tuning turbulent mixing coefficients properly. However, the effect of gap width observed in the experiments could not be predicted by the subchannel codes. (2)As for two-phase flow mixing experiments between two channels, in high water flow rate cases, the calculated distributions of air and water flows in each channel were well agreed with the experimental results. In low water flow cases, on the other hand, the air mixing rates were underestimated. (3)As for two-phase flow mixing experiments among multi-channels, the calculated mass velocities at channel exit under steady-state condition were agreed with experimental values within about 10%. However, the predictive errors of exit qualities were as high as 30%. (4)As for critical heat flux(CHF) experiments, two different results were obtained. A code indicated that the calculated CHF's using KfK or EPRI correlations were well agreed with the experimental results, while another code suggested that the CHF's were well predicted by using WSC-2 correlation or Weisman-Pei mechanistic model. (5)As for droplets entrainment and deposition experiments, it was indicated that the predictive capability was significantly increased by improving correlations. On the other hand, a remarkable discrepancy between codes was observed. That is, a code underestimated the droplet flow rate and overestimated the liquid film flow rate in high quality cases, while another code overestimated the droplet flow rate and underestimated the liquid film flow rate in low quality cases. (J.P.N.)

  7. The integral analysis of 40 mm diameter pipe rupture in cooling system of fusion facility W7-X with ASTEC code

    Energy Technology Data Exchange (ETDEWEB)

    Kačegavičius, Tomas, E-mail: Tomas.Kacegavicius@lei.lt; Povilaitis, Mantas, E-mail: Mantas.Povilaitis@lei.lt

    2015-12-15

    Highlights: • The analysis of loss-of-coolant accident (LOCA) in W7-X facility. • Burst disc is sufficient to prevent pressure inside the plasma vessel exceeding 110 kPa. • Developed model of the cooling system adequately represents the expected phenomena. - Abstract: Fusion is the energy production technology, which could potentially solve problems with growing energy demand of population in the future. Wendelstein 7-X (W7-X) is an experimental facility of stellarator type, which is currently being built at the Max-Planck-Institute for Plasmaphysics located in Greifswald, Germany. W7-X shall demonstrate that in future the energy could be produced in such type of fusion reactors. The safety analysis is required before the operation of the facility could be started. A rupture of 40 mm diameter pipe, which is connected to the divertor unit (module for plasma cooling) to ensure heat removal from the vacuum vessel in case of no-plasma operation mode “baking” is one of the design basis accidents to be investigated. During “baking” mode the vacuum vessel structures and working fluid – water are heated to the temperature 160 °C. This accident was selected for the detailed analysis using integral code ASTEC, which is developed by IRSN (France) and GRS mbH (Germany). This paper presents the integral analysis of W7-X response to a selected accident scenario. The model of the main cooling circuit and “baking” circuit was developed for ASTEC code. There were analysed two cases: (1) rupture of a pipe connected to the upper divertor unit and (2) rupture of a pipe connected to the lower divertor unit. The results of analysis showed that in both cases the water is almost completely released from the units into the plasma vessel. In both cases the pressure in the plasma vessel rapidly increases and in 28 s the set point for burst disc opening is reached preventing further pressurisation.

  8. Symbol synchronization in convolutionally coded systems

    Science.gov (United States)

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

    1979-01-01

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

  9. 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.

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

    Science.gov (United States)

    Healy, T. J.

    1985-01-01

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

  11. Demonstration study on shielding safety analysis code (VI)

    Energy Technology Data Exchange (ETDEWEB)

    Sawamura, Sadashi [Hokkaido Univ., Sapporo (Japan). Faculty of Engineering

    1999-03-01

    Dose evaluation for direct radiation and skyshine from nuclear fuel facilities is one of the environment evaluation items. This evaluation is carried out by using some shielding calculation codes. Because of extremely few benchmark data of skyshine, the calculation has to be performed very conservatively. Therefore, the benchmark data of skyshine and the well-investigated code for skyshine would be necessary to carry out the rational evaluation of nuclear facilities. The purpose of this steady is to obtain the benchmark data of skyshine and to investigate the calculation code for skyshine. In this fiscal year, the followings are investigated; (1) Construction and improvement of a pulsed radiation measurement system due to the gated counting method. (2) Using the system, carried out the radiation monitoring near and in the facility of 45 MeV Linear accelerator installed at Hokkaido University. (3) Simulation analysis of the photo-neutron production and the transport by using the EGS4 and MCNP code. (author)

  12. MELCOR Accident Consequence Code System (MACCS)

    International Nuclear Information System (INIS)

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

    1990-02-01

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

  13. 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.

  14. A development of containment performance analysis methodology using GOTHIC code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B. C.; Yoon, J. I. [Future and Challenge Company, Seoul (Korea, Republic of); Byun, C. S.; Lee, J. Y. [Korea Electric Power Research Institute, Taejon (Korea, Republic of); Lee, J. Y. [Seoul National University, Seoul (Korea, Republic of)

    2003-10-01

    In a circumstance that well-established containment pressure/temperature analysis code, CONTEMPT-LT treats the reactor containment as a single volume, this study introduces, as an alternative, the GOTHIC code for an usage on multi-compartmental containment performance analysis. With a developed GOTHIC methodology, its applicability is verified for containment performance analysis for Korean Nuclear Unit 1. The GOTHIC model for this plant is simply composed of 3 compartments including the reactor containment and RWST. In addition, the containment spray system and containment recirculation system are simulated. As a result of GOTHIC calculation, under the same assumptions and conditions as those in CONTEMPT-LT, the GOTHIC prediction shows a very good result; pressure and temperature transients including their peaks are nearly the same. It can be concluded that the GOTHIC could provide reasonable containment pressure and temperature responses if considering the inherent conservatism in CONTEMPT-LT code.

  15. A development of containment performance analysis methodology using GOTHIC code

    International Nuclear Information System (INIS)

    Lee, B. C.; Yoon, J. I.; Byun, C. S.; Lee, J. Y.; Lee, J. Y.

    2003-01-01

    In a circumstance that well-established containment pressure/temperature analysis code, CONTEMPT-LT treats the reactor containment as a single volume, this study introduces, as an alternative, the GOTHIC code for an usage on multi-compartmental containment performance analysis. With a developed GOTHIC methodology, its applicability is verified for containment performance analysis for Korean Nuclear Unit 1. The GOTHIC model for this plant is simply composed of 3 compartments including the reactor containment and RWST. In addition, the containment spray system and containment recirculation system are simulated. As a result of GOTHIC calculation, under the same assumptions and conditions as those in CONTEMPT-LT, the GOTHIC prediction shows a very good result; pressure and temperature transients including their peaks are nearly the same. It can be concluded that the GOTHIC could provide reasonable containment pressure and temperature responses if considering the inherent conservatism in CONTEMPT-LT code

  16. 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)

    Hursin, M.; Shanjie, X.; Burns, A.; Hopkins, J.; Satvat, N.; Gert, G.; Tsoukalas, L. H.; Jevremovic, T.

    2006-01-01

    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)

  17. Bar-code automated waste tracking system

    International Nuclear Information System (INIS)

    Hull, T.E.

    1994-10-01

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

  18. Source Code Vulnerabilities in IoT Software Systems

    Directory of Open Access Journals (Sweden)

    Saleh Mohamed Alnaeli

    2017-08-01

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

  19. Code comparison for accelerator design and analysis

    International Nuclear Information System (INIS)

    Parsa, Z.

    1988-01-01

    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

  20. Electromagnetic field and mechanical stress analysis code

    International Nuclear Information System (INIS)

    1978-01-01

    Analysis TEXMAGST is a two stage linear finite element code for the analysis of static magnetic fields in three dimensional structures and associated mechanical stresses produced by the anti J x anti B forces within these structures. The electromagnetic problem is solved in terms of magnetic vector potential A for a given current density anti J as curl 1/μ curl anti A = anti J considering the magnetic permeability as constant. The Coulombian gauge (div anti A = o) was chosen and was implemented through the use of Lagrange multipliers. The second stage of the problem - the calculation of mechanical stresses in the same three dimensional structure is solved by using the same code with few modifications - through a restart card. Body forces anti J x anti B within each element are calculated from the solution of the first stage run and represent the input to the second stage run which will give the solution for the stress problem

  1. Computer code and users' guide for the preliminary analysis of dual-mode space nuclear fission solid core power and propulsion systems, NUROC3A. AMS report No. 1239b

    Energy Technology Data Exchange (ETDEWEB)

    Nichols, R.A.; Smith, W.W.

    1976-06-30

    The three-volume report describes a dual-mode nuclear space power and propulsion system concept that employs an advanced solid-core nuclear fission reactor coupled via heat pipes to one of several electric power conversion systems. The second volume describes the computer code and users' guide for the preliminary analysis of the system.

  2. 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.

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

    International Nuclear Information System (INIS)

    Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E.; Tills, J.

    1997-12-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  5. 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.

  6. Transient electromagnetic analysis in tokamaks using TYPHOON code

    International Nuclear Information System (INIS)

    Belov, A.V.; Duke, A.E.; Korolkov, M.D.; Kotov, V.L.; Kukhtin, V.P.; Lamzin, E.A.; Sytchevsky, S.E.

    1996-01-01

    The transient electromagnetic analysis of conducting structures in tokamaks is presented. This analysis is based on a three-dimensional thin conducting shell model. The finite element method has been used to solve the corresponding integrodifferential equation. The code TYPHOON has been developed to calculate transient processes in tokamaks. Calculation tests and the code verification have been carried out. The calculation results of eddy current and force distibution and a.c. losses for different construction elements for both ITER and TEXTOR tokamaks magnetic systems are presented. (orig.)

  7. Operational reactor physics analysis codes (ORPAC)

    International Nuclear Information System (INIS)

    Kumar, Jainendra; Singh, K.P.; Singh, Kanchhi

    2007-07-01

    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)

  8. Development of disruption thermal analysis code DREAM

    Energy Technology Data Exchange (ETDEWEB)

    Yamazaki, Seiichiro; Kobayahsi, Takeshi [Kawasaki Heavy Industries Ltd., Kobe (Japan); Seki, Masahiro

    1989-07-01

    When a plasma disruption takes place in a tokamak type fusion reactor, plasma facing componenets such as first wall and divertor/limiter are subjected to a intensse heat load in a short duration. At the surface of the wall, temperature rapidly rises, and melting and evaporation occurs. It causes reduction of wall thickness and crack initiation/propagation. As lifetime of the components is significantly affected by them, the transient analysis in consideration of phase changes and radiation heat loss in required in the design of these components. This paper describes the computer code DREAM, developed to perform the disruption thermal analysis, taking phase changes and radiation into account. (author).

  9. Development of disruption thermal analysis code DREAM

    International Nuclear Information System (INIS)

    Yamazaki, Seiichiro; Kobayahsi, Takeshi; Seki, Masahiro.

    1989-01-01

    When a plasma disruption takes place in a tokamak type fusion reactor, plasma facing componenets such as first wall and divertor/limiter are subjected to a intensse heat load in a short duration. At the surface of the wall, temperature rapidly rises, and melting and evaporation occurs. It causes reduction of wall thickness and crack initiation/propagation. As lifetime of the components is significantly affected by them, the transient analysis in consideration of phase changes and radiation heat loss in required in the design of these components. This paper describes the computer code DREAM, developed to perform the disruption thermal analysis, taking phase changes and radiation into account. (author)

  10. NORTICA—a new code for cyclotron analysis

    Science.gov (United States)

    Gorelov, D.; Johnson, D.; Marti, F.

    2001-12-01

    The new package NORTICA (Numerical ORbit Tracking In Cyclotrons with Analysis) of computer codes for beam dynamics simulations is under development at NSCL. The package was started as a replacement for the code MONSTER [1] developed in the laboratory in the past. The new codes are capable of beam dynamics simulations in both CCF (Coupled Cyclotron Facility) accelerators, the K500 and K1200 superconducting cyclotrons. The general purpose of this package is assisting in setting and tuning the cyclotrons taking into account the main field and extraction channel imperfections. The computer platform for the package is Alpha Station with UNIX operating system and X-Windows graphic interface. A multiple programming language approach was used in order to combine the reliability of the numerical algorithms developed over the long period of time in the laboratory and the friendliness of modern style user interface. This paper describes the capability and features of the codes in the present state.

  11. Two-phase flow characteristics analysis code: MINCS

    International Nuclear Information System (INIS)

    Watanabe, Tadashi; Hirano, Masashi; Akimoto, Masayuki; Tanabe, Fumiya; Kohsaka, Atsuo.

    1992-03-01

    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)

  12. Analog system for computing sparse codes

    Science.gov (United States)

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

    2010-08-24

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

  13. A Consistent System for Coding Laboratory Samples

    Science.gov (United States)

    Sih, John C.

    1996-07-01

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

  14. Burnup calculation code system COMRAD96

    International Nuclear Information System (INIS)

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

    1997-06-01

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

  15. Burnup calculation code system COMRAD96

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-06-01

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

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

    International Nuclear Information System (INIS)

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

    1985-04-01

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

  17. A mean field theory of coded CDMA systems

    International Nuclear Information System (INIS)

    Yano, Toru; Tanaka, Toshiyuki; Saad, David

    2008-01-01

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

  18. 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.

  19. Steady-state natural circulation analysis with computational fluid dynamic codes of a liquid metal-cooled accelerator driven system

    International Nuclear Information System (INIS)

    Abanades, A.; Pena, A.

    2009-01-01

    A new innovative nuclear installation is under research in the nuclear community for its potential application to nuclear waste management and, above all, for its capability to enhance the sustainability of nuclear energy in the future as component of a new nuclear fuel cycle in which its efficiency in terms of primary Uranium ore profit and radioactive waste generation will be improved. Such new nuclear installations are called accelerator driven system (ADS) and are the result of a profitable symbiosis between accelerator technology, high-energy physics and reactor technology. Many ADS concepts are based on the utilization of heavy liquid metal (HLM) coolants due to its neutronic and thermo-physical properties. Moreover, such coolants permit the operation in free circulation mode, one of the main aims of passive systems. In this paper, such operation regime is analysed in a proposed ADS design applying computational fluid dynamics (CFD)

  20. SRAC95; general purpose neutronics code system

    International Nuclear Information System (INIS)

    Okumura, Keisuke; Tsuchihashi, Keichiro; Kaneko, Kunio.

    1996-03-01

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

  1. 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).

  2. 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)

  3. Revised SWAT. The integrated burnup calculation code system

    International Nuclear Information System (INIS)

    Suyama, Kenya; Mochizuki, Hiroki; Kiyosumi, Takehide

    2000-07-01

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

  4. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-15

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

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

    International Nuclear Information System (INIS)

    Bratfisch, Christoph; Koch, Marco K.

    2017-01-01

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

  7. NACOM - a code for sodium spray fire analysis

    International Nuclear Information System (INIS)

    Rao, P.M.; Kannan, S.E.

    2002-01-01

    Full text: In liquid metal fast breeder reactors (LMFBR), leakage of sodium can result in a spray fire. Because of higher burning rates in droplet form combustion of sodium in spray fire, thermal consequences are more severe than that in a sodium pool fire. The code NACOM was developed for the analysis of sodium spray fires in LMFBRs facilities. The code uses the validated model for estimating the falling droplet burning rates in pre-ignition and vapour phase combustion stages. It uses a distribution system to generate the droplet groups of different diameters that represent the spray. The code requires about 20 input parameters like sodium leak rates, sodium temperature, initial cell conditions like oxygen concentration, temperature and dimensions. NACOM is a validated code based on experiments with sodium inventory up to 650 kg in 0 to 21 % O 2 atmospheres. The paper brings out the salient features of the code along with the sensitivity analysis of the main input parameters like spray volume mean diameter, oxygen concentration etc. based on the results obtained. The limitations of the code and the confidence margins applicable to results obtained are also brought out

  8. Development of a code for the isotopic analysis of Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. H.; Kang, M. Y.; Kim, Jinhyeong; Choi, H. D. [Seoul National Univ., Seoul (Korea, Republic of)

    2013-05-15

    To strengthen the national nuclear nonproliferation regime by an establishment of nuclear forensic system, the techniques for nuclear material analysis and the categorization of important domestic nuclear materials are being developed. MGAU and FRAM are commercial software for the isotopic analysis of Uranium by using γ-spectroscopy, but the diversity of detection geometry and some effects - self attenuation, coincidence summing, etc. - suggest an analysis tool under continual improvement and modification. Hence, developing another code for HPGe γ- and x-ray spectrum analysis is started in this study. The analysis of the 87-101 keV region of Uranium spectrum is attempted based on the isotopic responses similar to those developed in MGAU. The code for isotopic analysis of Uranium is started from a fitting.

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  11. 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)

    Iwasaki, Takashi; Sakai, Takaaki; Enuma, Yasuhiro; Mizuno, Tomoyasu

    2002-03-01

    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)

  12. 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)

    Wu, T.; Cowan, C.L.; Lauer, A.; Schwiegk, H.J.

    1982-03-01

    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.)

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

    International Nuclear Information System (INIS)

    Brueckner-Foit, A.

    1987-10-01

    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.) [de

  14. OECD/DOE/CEA VVER-1000 coolant transient (V1000CT) benchmark for assessing coupled neutronics/thermal-hydraulics system codes for VVER-1000 RIA analysis

    International Nuclear Information System (INIS)

    Ivanov, B.; Ivanov, K.; Aniel, S.; Royer, E.; Kolev, N.; Groudev, P.

    2004-01-01

    The present paper describes the two phases of the OECD/DOE/CEA VVER-1000 coolant transient benchmark labeled as V1000CT. This benchmark is based on a data from the Bulgarian Kozloduy NPP Unit 6. The first phase of the benchmark was designed for the purpose of assessing neutron kinetics and thermal-hydraulic modeling for a VVER-1000 reactor, and specifically for their use in analyzing reactivity transients in a VVER-1000 reactor. Most of the results of Phase 1 will be compared against experimental data and the rest of the results will be used for code-to-code comparison. The second phase of the benchmark is planned for evaluation and improvement of the mixing computational models. Code-to-code and code-to-data comparisons will be done based on data of a mixing experiment conducted at Kozloduy-6. Main steam line break will be also analyzed in the second phase of the V1000CT benchmark. The results from it will be used for code-to-code comparison. The benchmark team has been involved in analyzing different aspects and performing sensitivity studies of the different benchmark exercises. The paper presents a comparison of selected results, obtained with two different system thermal-hydraulics codes, with the plant data for the Exercise 1 of Phase 1 of the benchmark as well as some results for Exercises 2 and 3. Overall, this benchmark has been well accepted internationally, with many organizations representing 11 countries participating in the first phase of the benchmark. (authors)

  15. Performance Analysis of DPSK Signals with Selection Combining and Convolutional Coding in Fading Channel

    National Research Council Canada - National Science Library

    Ong, Choon

    1998-01-01

    The performance analysis of a differential phase shift keyed (DPSK) communications system, operating in a Rayleigh fading environment, employing convolutional coding and diversity processing is presented...

  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. Integrated computer codes for nuclear power plant severe accident analysis

    International Nuclear Information System (INIS)

    Jordanov, I.; Khristov, Y.

    1995-01-01

    This overview contains a description of the Modular Accident Analysis Program (MAAP), ICARE computer code and Source Term Code Package (STCP). STCP is used to model TMLB sample problems for Zion Unit 1 and WWER-440/V-213 reactors. Comparison is made of STCP implementation on VAX and IBM systems. In order to improve accuracy, a double precision version of MARCH-3 component of STCP is created and the overall thermal hydraulics is modelled. Results of modelling the containment pressure, debris temperature, hydrogen mass are presented. 5 refs., 10 figs., 2 tabs

  18. Integrated computer codes for nuclear power plant severe accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Jordanov, I; Khristov, Y [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika

    1996-12-31

    This overview contains a description of the Modular Accident Analysis Program (MAAP), ICARE computer code and Source Term Code Package (STCP). STCP is used to model TMLB sample problems for Zion Unit 1 and WWER-440/V-213 reactors. Comparison is made of STCP implementation on VAX and IBM systems. In order to improve accuracy, a double precision version of MARCH-3 component of STCP is created and the overall thermal hydraulics is modelled. Results of modelling the containment pressure, debris temperature, hydrogen mass are presented. 5 refs., 10 figs., 2 tabs.

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    International Nuclear Information System (INIS)

    Chiba, Gou

    2003-09-01

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

  1. Performance analysis of spectral-phase-encoded optical code-division multiple-access system regarding the incorrectly decoded signal as a nonstationary random process

    Science.gov (United States)

    Yan, Meng; Yao, Minyu; Zhang, Hongming

    2005-11-01

    The performance of a spectral-phase-encoded (SPE) optical code-division multiple-access (OCDMA) system is analyzed. Regarding the incorrectly decoded signal (IDS) as a nonstationary random process, we derive a novel probability distribution for it. The probability distribution of the IDS is considered a chi-squared distribution with degrees of freedom r=1, which is more reasonable and accurate than in previous work. The bit error rate (BER) of an SPE OCDMA system under multiple-access interference is evaluated. Numerical results show that the system can sustain very low BER even when there are multiple simultaneous users, and as the code length becomes longer or the initial pulse becomes shorter, the system performs better.

  2. Code system BCG for gamma-ray skyshine calculation

    International Nuclear Information System (INIS)

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

    1979-03-01

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

  3. SWAAM-LT: The long-term, sodium/water reaction analysis method computer code

    International Nuclear Information System (INIS)

    Shin, Y.W.; Chung, H.H.; Wiedermann, A.H.; Tanabe, H.

    1993-01-01

    The SWAAM-LT Code, developed for analysis of long-term effects of sodium/water reactions, is discussed. The theoretical formulation of the code is described, including the introduction of system matrices for ease of computer programming as a general system code. Also, some typical results of the code predictions for available large scale tests are presented. Test data for the steam generator design with the cover-gas feature and without the cover-gas feature are available and analyzed. The capabilities and limitations of the code are then discussed in light of the comparison between the code prediction and the test data

  4. Concatenated coding system with iterated sequential inner decoding

    DEFF Research Database (Denmark)

    Jensen, Ole Riis; Paaske, Erik

    1995-01-01

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

  5. Performance testing of thermal analysis codes for nuclear fuel casks

    International Nuclear Information System (INIS)

    Sanchez, L.C.

    1987-01-01

    In 1982 Sandia National Laboratories held the First Industry/Government Joint Thermal and Structural Codes Information Exchange and presented the initial stages of an investigation of thermal analysis computer codes for use in the design of nuclear fuel shipping casks. The objective of the investigation was to (1) document publicly available computer codes, (2) assess code capabilities as determined from their user's manuals, and (3) assess code performance on cask-like model problems. Computer codes are required to handle the thermal phenomena of conduction, convection and radiation. Several of the available thermal computer codes were tested on a set of model problems to assess performance on cask-like problems. Solutions obtained with the computer codes for steady-state thermal analysis were in good agreement and the solutions for transient thermal analysis differed slightly among the computer codes due to modeling differences

  6. Analysis of ATLAS Cold Leg SBLOCA Using SPACE Code

    International Nuclear Information System (INIS)

    Kang, Doo Hyuk; Suh, Jae Seung; Kim, Se Yun

    2012-01-01

    SPACE Code has been developed to predict the thermal-hydraulic responses of nuclear steam supply system to the anticipated transients and postulated accidents and adopted advanced physical modeling of two-phase flows, mainly two-fluid, three-field models that comprise gas, continuous liquid, and droplet fields and has the capability to simulate 3D effects by the use of structured and/or non-structured meshes. In this paper, a cold-leg SBLOCA which is the experiment, SB-CL-09, of the ATLAS integral effect test facility during the second domestic stand problem (DSP-02) was analyzed. The results were compared with those of MARS-KS code simulations. The SPACE code with a 1.0 version was released by KHNP in 2012. The analysis has been performed in a desktop PC with Windows 7 environment

  7. The Vulnerability Assessment Code for Physical Protection System

    International Nuclear Information System (INIS)

    Jang, Sung Soon; Yoo, Ho Sik

    2007-01-01

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

  8. 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.

  9. 76 FR 64931 - Building Energy Codes Cost Analysis

    Science.gov (United States)

    2011-10-19

    ...-0046] Building Energy Codes Cost Analysis AGENCY: Office of Energy Efficiency and Renewable Energy... reopening of the time period for submitting comments on the request for information on Building Energy Codes... the request for information on Building Energy Code Cost Analysis and provide docket number EERE-2011...

  10. 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)

    Salazar C, J.H.; Nunez C, A.; Chavez M, C.

    2004-01-01

    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)

  11. Development of the criticality accident analysis code, AGNES

    International Nuclear Information System (INIS)

    Nakajima, Ken

    1989-01-01

    In the design works for the facilities which handle nuclear fuel, the evaluation of criticality accidents cannot be avoided even if their possibility is as small as negligible. In particular in the system using solution fuel like uranyl nitrate, solution has the property easily becoming dangerous form, and all the past criticality accidents occurred in the case of solution, therefore, the evaluation of criticality accidents becomes the most important item of safety analysis. When a criticality accident occurred in a solution fuel system, due to the generation and movement of radiolysis gas voids, the oscillation of power output and pressure pulses are observed. In order to evaluate the effect of criticality accidents, these output oscillation and pressure pulses must be calculated accurately. For this purpose, the development of the dynamic characteristic code AGNES (Accidentally Generated Nuclear Excursion Simulation code) was carried out. The AGNES is the reactor dynamic characteristic code having two independent void models. Modified energy model and pressure model, and as the benchmark calculation of the AGNES code, the results of the experimental analysis on the CRAC experiment are reported. (K.I.)

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

    Science.gov (United States)

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

    2004-01-01

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

  13. An assessment of the critical heat flux approaches of thermal-hydraulic system analysis codes using bundle data from the Heat Transfer Research Facility

    International Nuclear Information System (INIS)

    Min Lee

    1994-01-01

    Critical heat flux (CHF) bundle data from the Heat Transfer Research Facility of Columbia University are used to check the validity of the CHF approaches used in thermal-hydraulic system analysis codes for light water reactors. The CHF approaches assessed include the Biasi et al. correlation of TRAC, the Groeneveld et al. CHF table lookup approach of RELAP5/MOD3, the CHF table lookup approach of CATHARE, and the CHF approach of RETRAN. Depending on system pressure, RETRAN uses the B and W2, Barnett, and modified Barnett correlations and a linear interpolation scheme to predict CHF. Results show that among these CHF approaches, the Groeneveld et al. approach has the best prediction accuracy and the smallest uncertainty in the estimation of the HTRF bundle data. On the average, the Groeneveld et al. approach overpredicts the uniform axial heat flux distribution by 3.6% and the nonuniform axial heat flux distribution by 0.9%. The performance of the RETRAN approach is comparable with that of the Groenevel et al. Approach for uniform axial heat flux. In general, the accuracy and the uncertainty of all the approaches, except that of CATHARE, are worse under a nonuniform axial heat distribution than under a uniform axial heat distribution. All the CHF approaches assessed have a tendency to overpredict the HTRF bundle data at low pressure, low measured CHF, and high CHF quality. The performance of the Groenevel et al. approach is improved through a CHF table update and modification of the bundle correction factor using the HTRF bundle data

  14. 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.

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

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Lee, Young Jin; Hwang, Moon Kyu

    2007-03-01

    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. Health effects estimation code development for accident consequence analysis

    International Nuclear Information System (INIS)

    Togawa, O.; Homma, T.

    1992-01-01

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

  17. 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)

    Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.

    1998-03-01

    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.) [de

  18. Seismic Analysis Code (SAC): Development, porting, and maintenance within a legacy code base

    Science.gov (United States)

    Savage, B.; Snoke, J. A.

    2017-12-01

    The Seismic Analysis Code (SAC) is the result of toil of many developers over almost a 40-year history. Initially a Fortran-based code, it has undergone major transitions in underlying bit size from 16 to 32, in the 1980s, and 32 to 64 in 2009; as well as a change in language from Fortran to C in the late 1990s. Maintenance of SAC, the program and its associated libraries, have tracked changes in hardware and operating systems including the advent of Linux in the early 1990, the emergence and demise of Sun/Solaris, variants of OSX processors (PowerPC and x86), and Windows (Cygwin). Traces of these systems are still visible in source code and associated comments. A major concern while improving and maintaining a routinely used, legacy code is a fear of introducing bugs or inadvertently removing favorite features of long-time users. Prior to 2004, SAC was maintained and distributed by LLNL (Lawrence Livermore National Lab). In that year, the license was transferred from LLNL to IRIS (Incorporated Research Institutions for Seismology), but the license is not open source. However, there have been thousands of downloads a year of the package, either source code or binaries for specific system. Starting in 2004, the co-authors have maintained the SAC package for IRIS. In our updates, we fixed bugs, incorporated newly introduced seismic analysis procedures (such as EVALRESP), added new, accessible features (plotting and parsing), and improved the documentation (now in HTML and PDF formats). Moreover, we have added modern software engineering practices to the development of SAC including use of recent source control systems, high-level tests, and scripted, virtualized environments for rapid testing and building. Finally, a "sac-help" listserv (administered by IRIS) was setup for SAC-related issues and is the primary avenue for users seeking advice and reporting bugs. Attempts are always made to respond to issues and bugs in a timely fashion. For the past thirty-plus years

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

    International Nuclear Information System (INIS)

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

    1996-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-05-01

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

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

    Science.gov (United States)

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

    2009-04-01

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

  2. 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.

  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. Implementation of the thermal-hydraulic transient analysis code RELAP4/MOD5 and MOD6 on the FACOM 230/75 computer system

    International Nuclear Information System (INIS)

    Kohsaka, Atsuo; Ishigai, Takahiro; Kumakura, Toshimasa; Naraoka, Ken-itsu

    1979-03-01

    Development efforts have continued on the extensively used LOCA analysis code RELAP-4, as seen in its history; that is, from the prototype version MOD2 to the latest one MOD6 which is capable of one-through calculations from blowdown to reflood phase of PWR-LOCA. Many improvements and refinements of the models have enlarged the scopes and extents of phenomena to treat. Correspondingly the size of program has increased version to version, and special programming techniques have continuously been introduced to manage the program within limited capacity of core memory. For example, the Dynamic Storage Allocation of MOD5 and the PRELOAD Preprocessor newly incorporated in MOD6 are those designed for the CDC computer with relatively small core size. Described are these programming techniques in detail and experiences on implementation of the codes on FACOM 230/75, together with some results of confirmatory calculations. (author)

  5. Computer codes and methods for simulating accelerator driven systems

    International Nuclear Information System (INIS)

    Sartori, E.; Byung Chan Na

    2003-01-01

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

  6. Performance analysis of WS-EWC coded optical CDMA networks with/without LDPC codes

    Science.gov (United States)

    Huang, Chun-Ming; Huang, Jen-Fa; Yang, Chao-Chin

    2010-10-01

    One extended Welch-Costas (EWC) code family for the wavelength-division-multiplexing/spectral-amplitude coding (WDM/SAC; WS) optical code-division multiple-access (OCDMA) networks is proposed. This system has a superior performance as compared to the previous modified quadratic congruence (MQC) coded OCDMA networks. However, since the performance of such a network is unsatisfactory when the data bit rate is higher, one class of quasi-cyclic low-density parity-check (QC-LDPC) code is adopted to improve that. Simulation results show that the performance of the high-speed WS-EWC coded OCDMA network can be greatly improved by using the LDPC codes.

  7. Genetic Code Analysis Toolkit: A novel tool to explore the coding properties of the genetic code and DNA sequences

    Science.gov (United States)

    Kraljić, K.; Strüngmann, L.; Fimmel, E.; Gumbel, M.

    2018-01-01

    The genetic code is degenerated and it is assumed that redundancy provides error detection and correction mechanisms in the translation process. However, the biological meaning of the code's structure is still under current research. This paper presents a Genetic Code Analysis Toolkit (GCAT) which provides workflows and algorithms for the analysis of the structure of nucleotide sequences. In particular, sets or sequences of codons can be transformed and tested for circularity, comma-freeness, dichotomic partitions and others. GCAT comes with a fertile editor custom-built to work with the genetic code and a batch mode for multi-sequence processing. With the ability to read FASTA files or load sequences from GenBank, the tool can be used for the mathematical and statistical analysis of existing sequence data. GCAT is Java-based and provides a plug-in concept for extensibility. Availability: Open source Homepage:http://www.gcat.bio/

  8. Validation of OPERA3D PCMI Analysis Code

    Energy Technology Data Exchange (ETDEWEB)

    Jeun, Ji Hoon; Choi, Jae Myung; Yoo, Jong Sung [KEPCO Nuclear Fuel Co., Daejeon (Korea, Republic of); Cheng, G.; Sim, K. S.; Chassie, Girma [Candu Energy INC.,Ontario (Canada)

    2013-10-15

    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.

  9. Analytical validation of the CACECO containment analysis code

    International Nuclear Information System (INIS)

    Peak, R.D.

    1979-08-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Peak, R.D.

    1979-06-01

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

  11. Uranium Isotopic Analysis with the FRAM Isotopic Analysis Code

    International Nuclear Information System (INIS)

    Vo, D.T.; Sampson, T.E.

    1999-01-01

    FRAM is the acronym for Fixed-Energy Response-Function Analysis with Multiple efficiency. This software was developed at Los Alamos National Laboratory originally for plutonium isotopic analysis. Later, it was adapted for uranium isotopic analysis in addition to plutonium. It is a code based on a self-calibration using several gamma-ray peaks for determining the isotopic ratios. The versatile-parameter database structure governs all facets of the data analysis. User editing of the parameter sets allows great flexibility in handling data with different isotopic distributions, interfering isotopes, and different acquisition parameters such as energy calibration and detector type

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  13. Codeword Structure Analysis for LDPC Convolutional Codes

    Directory of Open Access Journals (Sweden)

    Hua Zhou

    2015-12-01

    Full Text Available The codewords of a low-density parity-check (LDPC convolutional code (LDPC-CC are characterised into structured and non-structured. The number of the structured codewords is dominated by the size of the polynomial syndrome former matrix H T ( D , while the number of the non-structured ones depends on the particular monomials or polynomials in H T ( D . By evaluating the relationship of the codewords between the mother code and its super codes, the low weight non-structured codewords in the super codes can be eliminated by appropriately choosing the monomials or polynomials in H T ( D , resulting in improved distance spectrum of the mother code.

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, M H; Hur, C; Kim, D K; Cho, H J [POhang Univ., of Science and TECHnology, Pohang (Korea, Republic of)

    1996-07-15

    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.

  16. Performance Analysis of Faulty Gallager-B Decoding of QC-LDPC Codes with Applications

    Directory of Open Access Journals (Sweden)

    O. Al Rasheed

    2014-06-01

    Full Text Available In this paper we evaluate the performance of Gallager-B algorithm, used for decoding low-density parity-check (LDPC codes, under unreliable message computation. Our analysis is restricted to LDPC codes constructed from circular matrices (QC-LDPC codes. Using Monte Carlo simulation we investigate the effects of different code parameters on coding system performance, under a binary symmetric communication channel and independent transient faults model. One possible application of the presented analysis in designing memory architecture with unreliable components is considered.

  17. Next generation Zero-Code control system UI

    CERN Multimedia

    CERN. Geneva

    2017-01-01

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

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

    CSIR Research Space (South Africa)

    Modipa, TI

    2013-12-01

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

  19. Parallelization of Subchannel Analysis Code MATRA

    International Nuclear Information System (INIS)

    Kim, Seongjin; Hwang, Daehyun; Kwon, Hyouk

    2014-01-01

    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

  20. 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)

    Lapins, J.; Seubert, A.; Buck, M.; Bader, J.; Laurien, E.

    2011-01-01

    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)

  1. Code development of total sensitivity and uncertainty analysis for reactor physics calculations

    International Nuclear Information System (INIS)

    Wan, C.; Cao, L.; Wu, H.; Zu, T.; Shen, W.

    2015-01-01

    Sensitivity and uncertainty analysis are essential parts for reactor system to perform risk and policy analysis. In this study, total sensitivity and corresponding uncertainty analysis for responses of neutronics calculations have been accomplished and developed the S&U analysis code named UNICORN. The UNICORN code can consider the implicit effects of multigroup cross sections on the responses. The UNICORN code has been applied to typical pin-cell case in this paper, and can be proved correct by comparison the results with those of the TSUNAMI-1D code. (author)

  2. Code development of total sensitivity and uncertainty analysis for reactor physics calculations

    Energy Technology Data Exchange (ETDEWEB)

    Wan, C.; Cao, L.; Wu, H.; Zu, T., E-mail: chenghuiwan@stu.xjtu.edu.cn, E-mail: caolz@mail.xjtu.edu.cn, E-mail: hongchun@mail.xjtu.edu.cn, E-mail: tiejun@mail.xjtu.edu.cn [Xi' an Jiaotong Univ., School of Nuclear Science and Technology, Xi' an (China); Shen, W., E-mail: Wei.Shen@cnsc-ccsn.gc.ca [Xi' an Jiaotong Univ., School of Nuclear Science and Technology, Xi' an (China); Canadian Nuclear Safety Commission, Ottawa, ON (Canada)

    2015-07-01

    Sensitivity and uncertainty analysis are essential parts for reactor system to perform risk and policy analysis. In this study, total sensitivity and corresponding uncertainty analysis for responses of neutronics calculations have been accomplished and developed the S&U analysis code named UNICORN. The UNICORN code can consider the implicit effects of multigroup cross sections on the responses. The UNICORN code has been applied to typical pin-cell case in this paper, and can be proved correct by comparison the results with those of the TSUNAMI-1D code. (author)

  3. Analysis of visual coding variables on CRT generated displays

    International Nuclear Information System (INIS)

    Blackman, H.S.; Gilmore, W.E.

    1985-01-01

    Cathode ray tube generated safety parameter display systems in a nuclear power plant control room situation have been found to be improved in effectiveness when color coding is employed. Research has indicated strong support for graphic coding techniques particularly in redundant coding schemes. In addition, findings on pictographs, as applied in coding schemes, indicate the need for careful application and for further research in the development of a standardized set of symbols

  4. Development of a neutronic analysis code using data from Monju

    International Nuclear Information System (INIS)

    Rooijen, W.F.G. van; Yamano, N.; Shimazu, Y.

    2015-01-01

    In recent years three major sets of modern evaluated nuclear data have become available: JENDL-4.0, JEFF-3.1.2 and ENDF/B-VII.1. The authors were involved with a research project to establish analysis method for a future commercial-scale LMFBR. This project focused on JENDL-4.0 and conventional Japanese codes. As a cross check, we decided to also apply the fast reactor code ERANOS. This necessitated to produce nuclear data (cross sections, etc) for the ERANOS code system, discussed in this paper. We developed a nuclear data processing system to produce cross sections, probability tables, delayed neutron data, and covariance data from the evaluated nuclear data files for ERANOS. A benchmark calculation on the MZA/MZB benchmark showed very satisfying results. Subsequently, we analyzed the prototype LMFBR Monju with ERANOS and our own sets of nuclear data. The results are very satisfactory. The results from ERANOS indicate that the target accuracies for nuclear data have not been met, although the three sets of evaluated nuclear data all performed very well in our analysis. In the future, the covariance on nuclear data should be reduced to meet the target accuracies on criticality and feedback coefficients. (author)

  5. A Robust Cross Coding Scheme for OFDM Systems

    NARCIS (Netherlands)

    Shao, X.; Slump, Cornelis H.

    2010-01-01

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

  6. LACEwING: A New Moving Group Analysis Code

    Energy Technology Data Exchange (ETDEWEB)

    Riedel, Adric R. [Department of Astronomy, California Institute of Technology, Pasadena, CA 91125 (United States); Blunt, Sarah C.; Faherty, Jacqueline K. [Department of Astrophysics, American Museum of Natural History, New York, NY 10024 (United States); Lambrides, Erini L. [Department of Physics and Astronomy, Johns Hopkins University, Baltimore, MD 21218 (United States); Rice, Emily L. [Department of Engineering Science and Physics, The College of Staten Island, Staten Island, NY 10314 (United States); Cruz, Kelle L., E-mail: arr@astro.caltech.edu [Department of Physics and Astronomy, Hunter College, New York, NY 10065 (United States)

    2017-03-01

    We present a new nearby young moving group (NYMG) kinematic membership analysis code, LocAting Constituent mEmbers In Nearby Groups (LACEwING), a new Catalog of Suspected Nearby Young Stars, a new list of bona fide members of moving groups, and a kinematic traceback code. LACEwING is a convergence-style algorithm with carefully vetted membership statistics based on a large numerical simulation of the Solar Neighborhood. Given spatial and kinematic information on stars, LACEwING calculates membership probabilities in 13 NYMGs and three open clusters within 100 pc. In addition to describing the inputs, methods, and products of the code, we provide comparisons of LACEwING to other popular kinematic moving group membership identification codes. As a proof of concept, we use LACEwING to reconsider the membership of 930 stellar systems in the Solar Neighborhood (within 100 pc) that have reported measurable lithium equivalent widths. We quantify the evidence in support of a population of young stars not attached to any NYMGs, which is a possible sign of new as-yet-undiscovered groups or of a field population of young stars.

  7. Fast neutron analysis code SAD1

    International Nuclear Information System (INIS)

    Jung, M.; Ott, C.

    1985-01-01

    A listing and an example of outputs of the M.C. code SAD1 are given here. This code has been used many times to predict responses of fast neutrons in hydrogenic materials (in our case emulsions or plastics) towards the elastic n, p scattering. It can be easily extended to other kinds of such materials and to any kind of incident fast neutron spectrum

  8. Interface of RETRAN/MASTER Code System for APR1400

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  9. Variable code gamma ray imaging system

    International Nuclear Information System (INIS)

    Macovski, A.; Rosenfeld, D.

    1979-01-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    International Nuclear Information System (INIS)

    Yang, Yanhua; Zhang, Hao

    2016-01-01

    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.

  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. Preliminary Coupling of MATRA Code for Multi-physics Analysis

    International Nuclear Information System (INIS)

    Kim, Seongjin; Choi, Jinyoung; Yang, Yongsik; Kwon, Hyouk; Hwang, Daehyun

    2014-01-01

    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

  15. ZERBERUS - the code for reliability analysis of crack containing structures

    International Nuclear Information System (INIS)

    Cizelj, L.; Riesch-Oppermann, H.

    1992-04-01

    Brief description of the First- and Second Order Reliability Methods, being the theoretical background of the code, is given. The code structure is described in detail, with special emphasis to the new application fields. The numerical example investigates failure probability of steam generator tubing affected by stress corrosion cracking. The changes necessary to accommodate this analysis within the ZERBERUS code are explained. Analysis results are compared to different Monte Carlo techniques. (orig./HP) [de

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

    Science.gov (United States)

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

    2018-03-01

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

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

    International Nuclear Information System (INIS)

    Langenbuch, S.; Velkov, K.

    1999-01-01

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

  18. 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…

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

    DEFF Research Database (Denmark)

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

    2015-01-01

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

  20. Content Analysis Coding Schemes for Online Asynchronous Discussion

    Science.gov (United States)

    Weltzer-Ward, Lisa

    2011-01-01

    Purpose: Researchers commonly utilize coding-based analysis of classroom asynchronous discussion contributions as part of studies of online learning and instruction. However, this analysis is inconsistent from study to study with over 50 coding schemes and procedures applied in the last eight years. The aim of this article is to provide a basis…

  1. A Semantic Analysis Method for Scientific and Engineering Code

    Science.gov (United States)

    Stewart, Mark E. M.

    1998-01-01

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

  2. 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.

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

    International Nuclear Information System (INIS)

    2004-03-01

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

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  5. Subchannel analysis code development for CANDU fuel channel

    International Nuclear Information System (INIS)

    Park, J. H.; Suk, H. C.; Jun, J. S.; Oh, D. J.; Hwang, D. H.; Yoo, Y. J.

    1998-07-01

    Since there are several subchannel codes such as COBRA and TORC codes for a PWR fuel channel but not for a CANDU fuel channel in our country, the subchannel analysis code for a CANDU fuel channel was developed for the prediction of flow conditions on the subchannels, for the accurate assessment of the thermal margin, the effect of appendages, and radial/axial power profile of fuel bundles on flow conditions and CHF and so on. In order to develop the subchannel analysis code for a CANDU fuel channel, subchannel analysis methodology and its applicability/pertinence for a fuel channel were reviewed from the CANDU fuel channel point of view. Several thermalhydraulic and numerical models for the subchannel analysis on a CANDU fuel channel were developed. The experimental data of the CANDU fuel channel were collected, analyzed and used for validation of a subchannel analysis code developed in this work. (author). 11 refs., 3 tabs., 50 figs

  6. Pipe whip analysis using the TEDEL code

    International Nuclear Information System (INIS)

    Millard, D.; Hoffmann, A.

    1985-02-01

    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

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

    International Nuclear Information System (INIS)

    Cesari, F.; Ferranti, P.; Gasparrini, M.; Labanti, L.

    1975-01-01

    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

  8. Development status of Severe Accident Analysis Code SAMPSON

    International Nuclear Information System (INIS)

    Iwashita, Tsuyoshi; Ujita, Hiroshi

    2000-01-01

    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)

  9. Available computer codes and data for radiation transport analysis

    International Nuclear Information System (INIS)

    Trubey, D.K.; Maskewitz, B.F.; Roussin, R.W.

    1975-01-01

    The Radiation Shielding Information Center (RSIC), sponsored and supported by the Energy Research and Development Administration (ERDA) and the Defense Nuclear Agency (DNA), is a technical institute serving the radiation transport and shielding community. It acquires, selects, stores, retrieves, evaluates, analyzes, synthesizes, and disseminates information on shielding and ionizing radiation transport. The major activities include: (1) operating a computer-based information system and answering inquiries on radiation analysis, (2) collecting, checking out, packaging, and distributing large computer codes, and evaluated and processed data libraries. The data packages include multigroup coupled neutron-gamma-ray cross sections and kerma coefficients, other nuclear data, and radiation transport benchmark problem results

  10. Development validation and use of computer codes for inelastic analysis

    International Nuclear Information System (INIS)

    Jobson, D.A.

    1983-01-01

    A finite element scheme is a system which provides routines so carry out the operations which are common to all finite element programs. The list of items that can be provided as standard by the finite element scheme is surprisingly large and the list provided by the UNCLE finite element scheme is unusually comprehensive. This presentation covers the following: construction of the program, setting up a finite element mesh, generation of coordinates, incorporating boundary and load conditions. Program validation was done by creep calculations performed using CAUSE code. Program use is illustrated by calculating a typical inelastic analysis problem. This includes computer model of the PFR intermediate heat exchanger

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

    Science.gov (United States)

    Pinilla, Samuel; Poveda, Juan; Arguello, Henry

    2018-03-01

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

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

    Directory of Open Access Journals (Sweden)

    Bih-Chyun Yeh

    2016-01-01

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

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

    Science.gov (United States)

    Kumawat, Soma; Ravi Kumar, M.

    2016-07-01

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

  14. Severe accident analysis using MARCH 1.0 code

    International Nuclear Information System (INIS)

    Guimaraes, A.C.F.

    1987-09-01

    The description and utilization of the MARCH 1.0 computer code, which aim to analyse physical phenomena associated with core meltdown accidents in PWR type reactors, are presented. The primary system is modeled as a single volume which is partitioned into a gas (steam and hydrogen) region and a water region. March predicts blowdown from the primary system in single phase. Based on results of the probabilistic safety analysis for the Zion and Indian Point Nuclear Power Plants, the S 2 HFX sequence accident for Angra-1 reactor is studied. The S 2 HFX sequence means that the loss of coolant accident occurs through small break in primary system with bot total failures of the reactor safety system and containment in yours recirculation modes, leading the core melt and the containment failure due to overpressurization. The obtained results were considered reasonable if compared with the results obtained for the Zion and Indian Point nuclear power plants. (Author) [pt

  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. Evaluation of the DRAGON code for VHTR design analysis

    International Nuclear Information System (INIS)

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

    2006-01-01

    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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

    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)

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

    International Nuclear Information System (INIS)

    Vertes, P.

    1992-09-01

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

  19. The Analysis of SBWR Critical Power Bundle Using Cobrag Code

    Directory of Open Access Journals (Sweden)

    Yohannes Sardjono

    2013-03-01

    Full Text Available The coolant mechanism of SBWR is similar with the Dodewaard Nuclear Power Plant (NPP in the Netherlands that first went critical in 1968. The similarity of both NPP is cooled by natural convection system. These coolant concept is very related with same parameters on fuel bundle design especially fuel bundle length, core pressure drop and core flow rate as well as critical power bundle. The analysis was carried out by using COBRAG computer code. COBRAG computer code is GE Company proprietary. Basically COBRAG computer code is a tool to solve compressible three-dimensional, two fluid, three field equations for two phase flow. The three fields are the vapor field, the continuous liquid field, and the liquid drop field. This code has been applied to analyses model flow and heat transfer within the reactor core. This volume describes the finitevolume equations and the numerical solution methods used to solve these equations. This analysis of same parameters has been done i.e.; inlet sub cooling 20 BTU/lbm and 40 BTU/lbm, 1000 psi pressure and R-factor is 1.038, mass flux are 0.5 Mlb/hr.ft2, 0.75 Mlb/hr.ft2, 1.00 Mlb/hr.ft2 and 1.25 Mlb/hr.ft2. Those conditions based on history operation of some type of the cell fuel bundle line at GE Nuclear Energy. According to the results, it can be concluded that SBWR critical power bundle is 10.5 % less than current BWR critical power bundle with length reduction of 12 ft to 9 ft.

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

    International Nuclear Information System (INIS)

    Liu, X.J.; Cheng, X.

    2015-01-01

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

  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. Post test analysis of TEPSS tests -P2-, -P3-, -P5- and -P7- using the system code RELAP5/MOD 3.2

    International Nuclear Information System (INIS)

    Luebbesmeyer, D.

    2000-01-01

    For the PANDA-Test-Facility (TEPSS configuration) post-test calculations and analyses have been performed for experiment -P2- (Early Start), -P3- (PCC start up), -P5- (Symmetric case, Two PCCs only) and -P7- (Severe Accident). Post test calculations have been performed with the system code RELAP5/Mod 3.2 using two different nodalization of the PANDA facility namely a basis nodalization and a much reduced one. The general trend of the calculations can be summarised: RELAP5/Mod3.2 calculated the general trends of the experiments sufficiently accurate; Using the reduced nodalization the results seem to be slightly more accurate than for the basic nodalization; On the other hand, calculations based on the reduced nodalization are not significantly faster than those with basic nodalization; The mass error is in the order of 200 to 900 kg. (author)

  3. 14 CFR Sec. 1-4 - System of accounts coding.

    Science.gov (United States)

    2010-01-01

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

  4. Status of the CONTAIN computer code for LWR containment analysis

    International Nuclear Information System (INIS)

    Bergeron, K.D.; Murata, K.K.; Rexroth, P.E.; Clauser, M.J.; Senglaub, M.E.; Sciacca, F.W.; Trebilcock, W.

    1983-01-01

    The current status of the CONTAIN code for LWR safety analysis is reviewed. Three example calculations are discussed as illustrations of the code's capabilities: (1) a demonstration of the spray model in a realistic PWR problem, and a comparison with CONTEMPT results; (2) a comparison of CONTAIN results for a major aerosol experiment against experimental results and predictions of the HAARM aerosol code; and (3) an LWR sample problem, involving a TMLB' sequence for the Zion reactor containment

  5. Status of the CONTAIN computer code for LWR containment analysis

    International Nuclear Information System (INIS)

    Bergeron, K.D.; Murata, K.K.; Rexroth, P.E.; Clauser, M.J.; Senglaub, M.E.; Sciacca, F.W.; Trebilcock, W.

    1982-01-01

    The current status of the CONTAIN code for LWR safety analysis is reviewed. Three example calculations are discussed as illustrations of the code's capabilities: (1) a demonstration of the spray model in a realistic PWR problem, and a comparison with CONTEMPT results; (2) a comparison of CONTAIN results for a major aerosol experiment against experimental results and predictions of the HAARM aerosol code; and (3) an LWR sample problem, involving a TMLB' sequence for the Zion reactor containment

  6. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

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

    International Nuclear Information System (INIS)

    1995-08-01

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

  9. A Comparative Study on Seismic Analysis of Bangladesh National Building Code (BNBC) with Other Building Codes

    Science.gov (United States)

    Bari, Md. S.; Das, T.

    2013-09-01

    Tectonic framework of Bangladesh and adjoining areas indicate that Bangladesh lies well within an active seismic zone. The after effect of earthquake is more severe in an underdeveloped and a densely populated country like ours than any other developed countries. Bangladesh National Building Code (BNBC) was first established in 1993 to provide guidelines for design and construction of new structure subject to earthquake ground motions in order to minimize the risk to life for all structures. A revision of BNBC 1993 is undergoing to make this up to date with other international building codes. This paper aims at the comparison of various provisions of seismic analysis as given in building codes of different countries. This comparison will give an idea regarding where our country stands when it comes to safety against earth quake. Primarily, various seismic parameters in BNBC 2010 (draft) have been studied and compared with that of BNBC 1993. Later, both 1993 and 2010 edition of BNBC codes have been compared graphically with building codes of other countries such as National Building Code of India 2005 (NBC-India 2005), American Society of Civil Engineering 7-05 (ASCE 7-05). The base shear/weight ratios have been plotted against the height of the building. The investigation in this paper reveals that BNBC 1993 has the least base shear among all the codes. Factored Base shear values of BNBC 2010 are found to have increased significantly than that of BNBC 1993 for low rise buildings (≤20 m) around the country than its predecessor. Despite revision of the code, BNBC 2010 (draft) still suggests less base shear values when compared to the Indian and American code. Therefore, this increase in factor of safety against the earthquake imposed by the proposed BNBC 2010 code by suggesting higher values of base shear is appreciable.

  10. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-07-01

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

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    Mochizuki, Hiroki; Suyama, Kenya; Okuno, Hiroshi

    2003-01-01

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

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

    International Nuclear Information System (INIS)

    Catsaros, N.; Gaveau, B.; Jaekel, M.; Maillard, J.; Maurel, G.; Savva, P.; Silva, J.; Varvayanni, M.; Zisis, Th.

    2009-01-01

    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.

  15. Simplified diagnostic coding sheet for computerized data storage and analysis in ophthalmology.

    Science.gov (United States)

    Tauber, J; Lahav, M

    1987-11-01

    A review of currently-available diagnostic coding systems revealed that most are either too abbreviated or too detailed. We have compiled a simplified diagnostic coding sheet based on the International Coding and Diagnosis (ICD-9), which is both complete and easy to use in a general practice. The information is transferred to a computer, which uses the relevant (ICD-9) diagnoses as database and can be retrieved later for display of patients' problems or analysis of clinical data.

  16. A code for structural analysis of fuel assemblies

    International Nuclear Information System (INIS)

    Hayashi, I.M.V.; Perrotta, J.A.

    1988-08-01

    It's presented the code ELCOM for the matrix analysis of tubular structures coupled by rigid spacers, typical of PWR's fuel elements. The code ELCOM makes a static structural analysis, where the displacements and internal forces are obtained for each tubular structure at the joints with the spacers, and also, the natural frequencies and vibrational modes of an equilavent integrated structure are obtained. The ELCOM result is compared to a PWR fuel element structural analysis obtained in published paper. (author) [pt

  17. Multidisciplinary System Reliability Analysis

    Science.gov (United States)

    Mahadevan, Sankaran; Han, Song; Chamis, Christos C. (Technical Monitor)

    2001-01-01

    The objective of this study is to develop a new methodology for estimating the reliability of engineering systems that encompass multiple disciplines. The methodology is formulated in the context of the NESSUS probabilistic structural analysis code, developed under the leadership of NASA Glenn Research Center. The NESSUS code has been successfully applied to the reliability estimation of a variety of structural engineering systems. This study examines whether the features of NESSUS could be used to investigate the reliability of systems in other disciplines such as heat transfer, fluid mechanics, electrical circuits etc., without considerable programming effort specific to each discipline. In this study, the mechanical equivalence between system behavior models in different disciplines are investigated to achieve this objective. A new methodology is presented for the analysis of heat transfer, fluid flow, and electrical circuit problems using the structural analysis routines within NESSUS, by utilizing the equivalence between the computational quantities in different disciplines. This technique is integrated with the fast probability integration and system reliability techniques within the NESSUS code, to successfully compute the system reliability of multidisciplinary systems. Traditional as well as progressive failure analysis methods for system reliability estimation are demonstrated, through a numerical example of a heat exchanger system involving failure modes in structural, heat transfer and fluid flow disciplines.

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

    Energy Technology Data Exchange (ETDEWEB)

    Shahedi, S. [Department of Energy Engineering, Sharif University of Technology, Azadi Street, Tehran (Iran, Islamic Republic of); Jafari, J., E-mail: jalil_jafari@yahoo.co [Reactors and Accelerators R and D School, Nuclear Science and Technology Research Institute, North Kargar Street, Tehran (Iran, Islamic Republic of); Boroushaki, M. [Department of Energy Engineering, Sharif University of Technology, Azadi Street, Tehran (Iran, Islamic Republic of); D' Auria, F. [DIMNP, University of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy)

    2010-10-15

    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.

  19. PAPIRUS - a computer code for FBR fuel performance analysis

    International Nuclear Information System (INIS)

    Kobayashi, Y.; Tsuboi, Y.; Sogame, M.

    1991-01-01

    The FBR fuel performance analysis code PAPIRUS has been developed to design fuels for demonstration and future commercial reactors. A pellet structural model was developed to describe the generation, depletion and transport of vacancies and atomic elements in unified fashion. PAPIRUS results in comparison with the power - to - melt test data from HEDL showed validity of the code at the initial reactor startup. (author)

  20. 76 FR 57982 - Building Energy Codes Cost Analysis

    Science.gov (United States)

    2011-09-19

    ... DEPARTMENT OF ENERGY Office of Energy Efficiency and Renewable Energy [Docket No. EERE-2011-BT-BC-0046] Building Energy Codes Cost Analysis Correction In notice document 2011-23236 beginning on page...-23236 Filed 9-16-11; 8:45 am] BILLING CODE 1505-01-P ...

  1. An economic analysis code used for PWR fuel cycle

    International Nuclear Information System (INIS)

    Liu Dingqin

    1989-01-01

    An economic analysis code used for PWR fuel cycle is developed. This economic code includes 12 subroutines representing vavious processes for entire PWR fuel cycle, and indicates the influence of the fuel cost on the cost of the electricity generation and the influence of individual process on the sensitivity of the fuel cycle cost

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

    International Nuclear Information System (INIS)

    Holowach, M.J.; Hochreiter, L.E.; Cheung, F.B.

    2002-01-01

    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)

  3. Simulation of LOCA power transients of CANDU6 by SCAN/RELAP-CANDU coupled code system

    International Nuclear Information System (INIS)

    Hong, In Seob; Kim, Chang Hyo; Hwang, Su Hyun; Kim, Man Woong; Chung, Bub Dong

    2004-01-01

    As can be seen in the standalone application of RELAP-CANDU for LOCA analysis of CANDU-PHWR, the system thermal-hydraulic code alone cannot predict the transient behavior accurately. Therefore, best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. The purpose of this research is to develop and test a coupled neutronics and thermal-hydraulics analysis code, SCAN (SNU CANDU-PHWR Neutronics) and RELAP-CANDU, for transient analysis of CANDU-PHWR's. For this purpose, a spatial kinetics calculation module of SCAN, a 3-D CANDU-PHWR neutronics design and analysis code, is dynamically coupled with RELAP-CANDU, the system thermal-hydraulic code for CANDU-PHWR. The performance of the coupled code system is examined by simulation of reactor power transients caused by a hypothetical Loss Of Coolant Accident (LOCA) in Wolsong units, which involves the insertion of positive void reactivity into the core in the course of transients. Specifically, a 40% Reactor Inlet Header (RIH) break LOCA was assumed for the test of the SCAN/RELAP-CANDU coupled code system analysis

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

    International Nuclear Information System (INIS)

    Murao, Yoshio; Okubo, Tsutomu; Sugimoto, Jun; Iguchi, Tadashi; Sudoh, Takashi.

    1985-02-01

    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)

  5. Safety analysis code SCTRAN development for SCWR and its application to CGNPC SCWR

    International Nuclear Information System (INIS)

    Wu, Pan; Gou, Junli; Shan, Jianqiang; Jiang, Yang; Yang, Jue; Zhang, Bo

    2013-01-01

    Highlights: ► A new safety analysis code named SCTRAN is developed for SCWRs. ► Capability of SCTRAN is verified by comparing with code APROS and RELAP5-3D. ► A new passive safety system is proposed for CGNPC SCWR and analyzed with SCTRAN. ► CGNPC SCWR is able to cope with two critical accidents for SCWRs, LOFA and LOCA. - Abstract: Design analysis is one of the main difficulties during the research and design of SCWRs. Currently, the development of safety analysis code for SCWR is still in its infancy all around the world, and very few computer codes could carry out the trans-critical calculations where significant changes in water properties would take place. In this paper, a safety analysis code SCTRAN for SCWRs has been developed based on code RETRAN-02, the best estimate code used for safety analysis of light water reactors. The ability of SCTRAN code to simulate transients where both supercritical and subcritical regimes are encountered has been verified by comparing with APROS and RELAP5-3D codes. Furthermore, the LOFA and LOCA transients for the CGNPC SCWR design were analyzed with SCTRAN code. The characteristics and performance of the passive safety systems applied to CGNPC SCWR were evaluated. The results show that: (1) The SCTRAN computer code developed in this study is capable to perform design analysis for SCWRs; (2) During LOFA and LOCA accidents in a CGNPC SCWR, the passive safety systems would significantly mitigate the consequences of these transients and enhance the inherent safety

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

    International Nuclear Information System (INIS)

    Sasaki, Makoto; Nakazawa, Masaharu.

    1981-01-01

    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)

  7. Thermohydraulic analysis of nuclear power plant accidents by computer codes

    International Nuclear Information System (INIS)

    Petelin, S.; Stritar, A.; Istenic, R.; Gregoric, M.; Jerele, A.; Mavko, B.

    1982-01-01

    RELAP4/MOD6, BRUCH-D-06, CONTEMPT-LT-28, RELAP5/MOD1 and COBRA-4-1 codes were successful y implemented at the CYBER 172 computer in Ljubljana. Input models of NPP Krsko for the first three codes were prepared. Because of the high computer cost only one analysis of double ended guillotine break of the cold leg of NPP Krsko by RELAP4 code has been done. BRUCH code is easier and cheaper for use. Several analysis have been done. Sensitivity study was performed with CONTEMPT-LT-28 for double ended pump suction break. These codes are intended to be used as a basis for independent safety analyses. (author)

  8. 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.

  9. 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