WorldWideScience

Sample records for uedge code comparisons

  1. UEDGE code comparisons with DIII-D bolometer data

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, J.M.

    1994-12-01

    This paper describes the work done to develop a bolometer post processor that converts volumetric radiated power values taken from a UEDGE solution, to a line integrated radiated power along chords of the bolometers in the DIII-D tokamak. The UEDGE code calculates plasma physics quantities, such as plasma density, radiated power, or electron temperature, and compares them to actual diagnostic measurements taken from the scrape off layer (SOL) and divertor regions of the DIII-D tokamak. Bolometers are devices measuring radiated power within the tokamak. The bolometer interceptors are made up of two complete arrays, an upper array with a vertical view and a lower array with a horizontal view, so that a two dimensional profile of the radiated power may be obtained. The bolometer post processor stores line integrated values taken from UEDGE solutions into a file in tabular format. Experimental data is then put into tabular form and placed in another file. Comparisons can be made between the UEDGE solutions and actual bolometer data. Analysis has been done to determine the accuracy of the plasma physics involved in producing UEDGE simulations.

  2. Models and applications of the UEDGE code

    International Nuclear Information System (INIS)

    Rensink, M.E.; Knoll, D.A.; Porter, G.D.; Rognlien, T.D.; Smith, G.R.; Wising, F.

    1996-09-01

    The transport of particles and energy from the core of a tokamak to nearby material surfaces is an important problem for understanding present experiments and for designing reactor-grade devices. A number of fluid transport codes have been developed to model the plasma in the edge and scrape-off layer (SOL) regions. This report will focus on recent model improvements and illustrative results from the UEDGE code. Some geometric and mesh considerations are introduced, followed by a general description of the plasma and neutral fluid models. A few comments on computational issues are given and then two important applications are illustrated concerning benchmarking and the ITER radiative divertor. Finally, we report on some recent work to improve the models in UEDGE by coupling to a Monte Carlo neutrals code and by utilizing an adaptive grid

  3. Status of parallel Python-based implementation of UEDGE

    Science.gov (United States)

    Umansky, M. V.; Pankin, A. Y.; Rognlien, T. D.; Dimits, A. M.; Friedman, A.; Joseph, I.

    2017-10-01

    The tokamak edge transport code UEDGE has long used the code-development and run-time framework Basis. However, with the support for Basis expected to terminate in the coming years, and with the advent of the modern numerical language Python, it has become desirable to move UEDGE to Python, to ensure its long-term viability. Our new Python-based UEDGE implementation takes advantage of the portable build system developed for FACETS. The new implementation gives access to Python's graphical libraries and numerical packages for pre- and post-processing, and support of HDF5 simplifies exchanging data. The older serial version of UEDGE has used for time-stepping the Newton-Krylov solver NKSOL. The renovated implementation uses backward Euler discretization with nonlinear solvers from PETSc, which has the promise to significantly improve the UEDGE parallel performance. We will report on assessment of some of the extended UEDGE capabilities emerging in the new implementation, and will discuss the future directions. Work performed for U.S. DOE by LLNL under contract DE-AC52-07NA27344.

  4. UEDGE simulations of He transport in DIII-D progress report FY 1996

    International Nuclear Information System (INIS)

    Fenstermacher, M. E.; Hill, D.N.

    1997-01-01

    In this report we present the status of numerical simulations of helium exhaust efficiency in the DIII-D tokamak. These computations are intended to serve eventually as a benchmark for simulations carried out for the ITER divertor geometry. Helium ash removal is an important issue for ITER since the helium ash can dilute the central fuel concentration and reduce the fusion power. Present experiments have shown that helium transport in the core plasma is sufficiently rapid to limit the ash buildup to acceptable levels if sufficient helium pumping can be maintained in the divertor. The question of pumping helium gas from the divertor has also been addressed in tokamak experiments, where it was found that the helium concentration in the divertor was about 5-10x lower than in the core plasma (deenrichment). Even so, the exhaust rate was adequate to meet the ITER requirements for central helium concentration. However, the experiments did not reproduce the anticipated ITER divertor geometry or operating conditions. Therefore, the predicted helium exhaust for ITER is still based on numerical simulation. In order to increase the confidence level in the simulations of helium exhaust in ITER, we decided to test the ability of the UEDGE code to simulate the measured enrichment of divertor helium in the DIII-D pumping plenum. Section II presents a description of the experimental discharge used for comparison with the present UEDGE simulations. The UEDGE runs which most closely match the data are presented in Section III including simulations with and without carbon impurity. Section IV presents UEDGE simulations of helium transport and comparison with the helium measurements for these discharges. Conclusions and plans for future work, to complete the detailed benchmarking of UEDGE helium transport models, are given in Section V. 6 refs., 26 figs., 1 tab

  5. Evaluation of an improved atomic data basis for carbon in UEDGE emission modeling for L-mode plasmas in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Muñoz Burgos, J.M., E-mail: munozj@fusion.gat.com [Oak Ridge Institute for Science and Education, Oak Ridge, TN 37831-0117 (United States); Leonard, A.W. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Loch, S.D.; Ballance, C.P. [Auburn University, Auburn, AL 36849 (United States)

    2013-07-15

    New scaled carbon atomic electron-impact excitation data is utilized to evaluate comparisons between experimental measurements and fluid emission modeling of detached plasmas at DIII-D. The C I and C II modeled emission lines for 909.8 and 514.7 nm were overestimated by a factor of 10–20 than observed experimentally for the inner leg, while the outer leg was within a factor of 2. Due to higher modeled emissions, a previous study using the UEDGE code predicted that a higher amount of carbon was required to achieve a detached outboard divertor plasma in L-mode at DIII-D. The line emission predicted by using the new scaled carbon data yields closer results when compared against experiment. We also compare modeling and measurements of D{sub α} emission from neutral deuterium against predictions from newly calculated R-Matrix with pseudostates data available at the ADAS database.

  6. UEDGE Simulations for Power and Particle Flow Analysis of FRC Rocket

    Science.gov (United States)

    Zheng, Fred; Evans, Eugene S.; McGreivy, Nick; Kaptanoglu, Alan; Izacard, Olivier; Cohen, Samuel A.

    2017-10-01

    The field-reversed configuration (FRC) is under consideration for use in a direct fusion drive (DFD) rocket propulsion system for future space missions. To achieve a rocket configuration, the FRC is embedded within an asymmetric magnetic mirror, in which one end is closed and contains a gas box, and the other end is open and incorporates a magnetic nozzle. Neutral deuterium is injected into the gas box, and flows through the scrape-off layer (SOL) around the core plasma and out the magnetic nozzle, both cooling the core and serving as propellant. Previous studies have examined a range of operating conditions for the SOL of a DFD using UEDGE, a 2D fluid code; discrepancies on the order of 5% were found during the analysis of overall power balance. This work extends the analysis of the previously-studied SOL geometry by updating boundary conditions and conducting a detailed study of power and particle flows within the simulation with the goals of modeling electrical power generation instead of thrust and achieving higher specific impulse. This work was supported, in part, by DOE Contract Number DE-AC02-09CH11466 and Princeton Environmental Institute.

  7. Quantitative comparisons between experimentally measured 2-D carbon radiation and Monte Carlo impurity (MCI) code simulations

    International Nuclear Information System (INIS)

    Evans, T.E.; Leonard, A.W.; West, W.P.; Finkenthal, D.F.; Fenstermacher, M.E.; Porter, G.D.

    1998-08-01

    Experimentally measured carbon line emissions and total radiated power distributions from the DIII-D divertor and Scrape-Off Layer (SOL) are compared to those calculated with the Monte Carlo Impurity (MCI) model. A UEDGE background plasma is used in MCI with the Roth and Garcia-Rosales (RG-R) chemical sputtering model and/or one of six physical sputtering models. While results from these simulations do not reproduce all of the features seen in the experimentally measured radiation patterns, the total radiated power calculated in MCI is in relatively good agreement with that measured by the DIII-D bolometric system when the Smith78 physical sputtering model is coupled to RG-R chemical sputtering in an unaltered UEDGE plasma. Alternatively, MCI simulations done with UEDGE background ion temperatures along the divertor target plates adjusted to better match those measured in the experiment resulted in three physical sputtering models which when coupled to the RG-R model gave a total radiated power that was within 10% of measured value

  8. Improved core-edge tokamak transport simulations with the CORSICA 2 code

    International Nuclear Information System (INIS)

    Tarditi, A.; Cohen, R.H.; Crotinger, J.A.

    1996-01-01

    The CORSICA 2 code models the nonlinear transport between the core and the edge of a tokamak plasma. The code couples a 2D axisymmetric edge/SOL model (UEDGE) to a 1D model for the radial core transport in toroidal flux coordinates (the transport module from the CORSICA 1 code). The core density and temperature profiles are joined to the flux-surface average profiles from the 2D code sufficiently inside the magnetic separatrix, at a flux surface on which the edge profiles are approximately constant. In the present version of the code, the deuterium density and electron and ion temperatures are coupled. The electron density is determined by imposing quasi-neutrality, both in the core and in the edge. The model allows the core-edge coupling of multiple ion densities while retaining a single temperature (corresponding to the equilibration value) for the all ion species. Applications of CORSICA 2 to modeling the DIII-D tokamak are discussed. This work will focus on the simulation of the L-H transition, coupling a single ion species (deuterium) and the two (electron and ion) temperatures. These simulations will employ a new self-consistent model for the L-H transition that is being implemented in the UEDGE code. Applications to the modeling of ITER ignition scenarios are also discussed. This will involve coupling a second density species (the thermal alphas), bringing the total number of coupled variables up to four. Finally, the progress in evolving the magnetic geometry is discussed. Currently, this geometry is calculated by CORSICA's MHD equilibrium module (TEQ) at the beginning of the run and fixed thereafter. However, CORSICA 1 can evolve this geometry quasistatically, and this quasistatic treatment is being extended to include the edge/SOL geometry. Recent improvements for code speed-up are also presented

  9. Simulation of experimentally achieved detached plasmas using the UEDGE code

    International Nuclear Information System (INIS)

    Porter, G.D.; Allen, S.; Fenstermacher, M.

    1995-01-01

    The introduction of a divertor Thomson scattering system in DIII-D has enabled accurate determination of the plasma properties in the divertor region. We identify two plasma regimes; detached and attached. The electron temperature in the detached regime is about 2 eV, much lower than 5 to 10 eV determined earlier. We show that fluid models of the DIII-D scrape-off layer plasma are able to reproduce many of the features of these two plasma regimes, including the boundaries for transition between them. Detailed comparison between the results obtained from the fluid models and experiment suggest the models underestimate the spatial extent of the low temperature region associated the detached plasma mode. We suggest that atomic physics processes at the low electron temperatures reported here may account for this discrepancy

  10. Numerical Tokamak Project code comparison

    International Nuclear Information System (INIS)

    Waltz, R.E.; Cohen, B.I.; Beer, M.A.

    1994-01-01

    The Numerical Tokamak Project undertook a code comparison using a set of TFTR tokamak parameters. Local radial annulus codes of both gyrokinetic and gyrofluid types were compared for both slab and toroidal case limits assuming ion temperature gradient mode turbulence in a pure plasma with adiabatic electrons. The heat diffusivities were found to be in good internal agreement within ± 50% of the group average over five codes

  11. Comparison of sodium aerosol codes

    International Nuclear Information System (INIS)

    Dunbar, I.H.; Fermandjian, J.; Bunz, H.; L'homme, A.; Lhiaubet, G.; Himeno, Y.; Kirby, C.R.; Mitsutsuka, N.

    1984-01-01

    Although hypothetical fast reactor accidents leading to severe core damage are very low probability events, their consequences are to be assessed. During such accidents, one can envisage the ejection of sodium, mixed with fuel and fission products, from the primary circuit into the secondary containment. Aerosols can be formed either by mechanical dispersion of the molten material or as a result of combustion of the sodium in the mixture. Therefore considerable effort has been devoted to study the different sodium aerosol phenomena. To ensure that the problems of describing the physical behaviour of sodium aerosols were adequately understood, a comparison of the codes being developed to describe their behaviour was undertaken. The comparison consists of two parts. The first is a comparative study of the computer codes used to predict aerosol behaviour during a hypothetical accident. It is a critical review of documentation available. The second part is an exercise in which code users have run their own codes with a pre-arranged input. For the critical comparative review of the computer models, documentation has been made available on the following codes: AEROSIM (UK), MAEROS (USA), HAARM-3 (USA), AEROSOLS/A2 (France), AEROSOLS/B1 (France), and PARDISEKO-IIIb (FRG)

  12. A CFD code comparison of wind turbine wakes

    DEFF Research Database (Denmark)

    Laan, van der, Paul Maarten; Storey, R. C.; Sørensen, Niels N.

    2014-01-01

    A comparison is made between the EllipSys3D and SnS CFD codes. Both codes are used to perform Large-Eddy Simulations (LES) of single wind turbine wakes, using the actuator disk method. The comparison shows that both LES models predict similar velocity deficits and stream-wise Reynolds-stresses fo...

  13. 24 CFR 200.926a - Residential building code comparison items.

    Science.gov (United States)

    2010-04-01

    ... 24 Housing and Urban Development 2 2010-04-01 2010-04-01 false Residential building code comparison items. 200.926a Section 200.926a Housing and Urban Development Regulations Relating to Housing and... § 200.926a Residential building code comparison items. HUD will review each local and State code...

  14. A photon dominated region code comparison study

    NARCIS (Netherlands)

    Roellig, M.; Abel, N. P.; Bell, T.; Bensch, F.; Black, J.; Ferland, G. J.; Jonkheid, B.; Kamp, I.; Kaufman, M. J.; Le Bourlot, J.; Le Petit, F.; Meijerink, R.; Morata, O.; Ossenkopf, Volker; Roueff, E.; Shaw, G.; Spaans, M.; Sternberg, A.; Stutzki, J.; Thi, W.-F.; van Dishoeck, E. F.; van Hoof, P. A. M.; Viti, S.; Wolfire, M. G.

    Aims. We present a comparison between independent computer codes, modeling the physics and chemistry of interstellar photon dominated regions (PDRs). Our goal was to understand the mutual differences in the PDR codes and their effects on the physical and chemical structure of the model clouds, and

  15. Comparison of beam deposition for three neutral beam injection codes

    International Nuclear Information System (INIS)

    Wieland, R.M.; Houlberg, W.A.; Mense, A.T.

    1979-03-01

    The three neutral beam injection codes BEAM (Houlberg, ORNL), HOFR (Howe, ORNL), and FREYA (Post, PPPL) are compared with respect to the calculation of the fast ion deposition profile H(r). Only plasmas of circular cross section are considered, with injection confined to the mid-plane of the torus. The approximations inherent in each code are pointed out, and a series of comparisons varying several parameters (beam energy and radius, machine size, and injection angle) shows excellent agreement among all the codes. A cost comparison (execution time and memory requirements) is made which points out the relative merits of each code within the context of incorporation into a plasma transport simulation code

  16. Review and comparison of WWER and LWR Codes and Standards

    International Nuclear Information System (INIS)

    Buckthorpe, D.; Tashkinov, A.; Brynda, J.; Davies, L.M.; Cueto-Felgeueroso, C.; Detroux, P.; Bieniussa, K.; Guinovart, J.

    2003-01-01

    The results of work on a collaborative project on comparison of Codes and Standards used for safety related components of the WWER and LWR type reactors is presented. This work was performed on behalf of the European Commission, Working Group Codes and Standards and considers areas such as rules, criteria and provisions, failure mechanisms , derivation and understanding behind the fatigue curves, piping, materials and aging, manufacturing and ISI. WWERs are essentially designed and constructed using the Russian PNAE Code together with special provisions in a few countries (e.g. Czech Republic) from national standards. The LWR Codes have a strong dependence on the ASME Code. Also within Western Europe other codes are used including RCC-M, KTA and British Standards. A comparison of procedures used in all these codes and standards have been made to investigate the potential for equivalencies between the codes and any grounds for future cooperation between eastern and western experts in this field. (author)

  17. Code comparison results for the loft LP-FP-2 experiment

    International Nuclear Information System (INIS)

    Merilo, M.; Mecham, D.C.

    1991-01-01

    Computer code calculations are compared with thermal hydraulic and fission product release, transport, and deposition data obtained from the OECD-LOFT LP-FP-2 experiment. Except for the MAAP code, which is a fully integrated severe accident code, the thermalhydraulic and fission product behavior were calculated with different codes. Six organizations participated in the thermal hydraulic portion of the code comparison exercise. These calculations were performed with RELAP 5, SCDAP/RELAP 5, and MAAP. The comparisons show generally well developed capabilities to determine the thermal-hydraulic conditions during the early stages of a severe core damage accident. Four participants submitted detailed fission product behavior calculations. Except for MAAP, as stated previously, the fission product inventory, core damage, fission product release, transport and deposition were calculated independently with different codes. Much larger differences than observed for the thermalhydraulic comparison were evident. The fission product inventory calculations were generally in good agreement with each other. Large differences were observed for release fractions and amounts of deposition. Net release calculations from the primary system were generally accurate within a factor of two or three for the more important fission products

  18. Progress report for FACETS (Framework Application for Core-Edge Transport Simulations): C.S. SAP

    Energy Technology Data Exchange (ETDEWEB)

    Epperly, T W

    2008-10-01

    FACETS framework to UEDGE uses a Forthon-generated Python layer that sits on top of the UEDGE Fortran code. To run on leadership class machines that do not support shared libraries such as Franklin (NERSCS) and Jaguar (ORNL), FACETS requires a more direct connection to the UEDGE Fortran that eliminates the Python layer. We designed and developed an extension for Forthon to provide a direct connection from C++ to the UEDGE Fortran using Babel. This also required added annotations to the Forthon '.v' input files to indicate how function parameters are used.

  19. Progress report for FACETS (Framework Application for Core-Edge Transport Simulations): C.S. SAP

    International Nuclear Information System (INIS)

    Epperly, T.W.

    2008-01-01

    framework to UEDGE uses a Forthon-generated Python layer that sits on top of the UEDGE Fortran code. To run on leadership class machines that do not support shared libraries such as Franklin (NERSCS) and Jaguar (ORNL), FACETS requires a more direct connection to the UEDGE Fortran that eliminates the Python layer. We designed and developed an extension for Forthon to provide a direct connection from C++ to the UEDGE Fortran using Babel. This also required added annotations to the Forthon '.v' input files to indicate how function parameters are used

  20. Optimization and Control of Burning Plasmas Through High Performance Computing

    Energy Technology Data Exchange (ETDEWEB)

    Pankin, Alexei [Tech-X Corporation, Boulder, CO (United States)

    2017-12-18

    This project has revived the FACETS code, that has been developed under SciDAC fund- ing in 2008-2012. The code has been dormant for a number of years after the SciDAC funding stopped. FACETS depends on external packages. The external packages and libraries such as PETSc, FFTW, HDF5 and NETCDF that are included in FACETS have evolved during these years. Some packages in FACETS are also parts of other codes such as PlasmaState, NUBEAM, GACODES, and UEDGE. These packages have been also evolved together with their host codes which include TRANSP, TGYRO and XPTOR. Finally, there is also a set of packages in FACETS that are being developed and maintained by Tech-X. These packages include BILDER, SciMake, and FcioWrappers. Many of these packages evolved significantly during the last several years and FACETS had to be updated to synchronize with the re- cent progress in the external packages. The PI has introduced new changes to the BILDER package to support the updated interfaces to the external modules. During the last year of the project, the FACETS version of the UEDGE code has been extracted from FACETS as a standalone package. The PI collaborates with the scientists from LLNL on the updated UEDGE model in FACETS. Drs. T. Rognlien, M. Umansky and A. Dimits from LLNL are contributing to this task.

  1. Pretest aerosol code comparisons for LWR aerosol containment tests LA1 and LA2

    International Nuclear Information System (INIS)

    Wright, A.L.; Wilson, J.H.; Arwood, P.C.

    1986-01-01

    The Light-Water-Reactor (LWR) Aerosol Containment Experiments (LACE) are being performed in Richland, Washington, at the Hanford Engineering Development Laboratory (HEDL) under the leadership of an international project board and the Electric Power Research Institute. These tests have two objectives: (1) to investigate, at large scale, the inherent aerosol retention behavior in LWR containments under simulated severe accident conditions, and (2) to provide an experimental data base for validating aerosol behavior and thermal-hydraulic computer codes. Aerosol computer-code comparison activities are being coordinated at the Oak Ridge National Laboratory. For each of the six LACE tests, ''pretest'' calculations (for code-to-code comparisons) and ''posttest'' calculations (for code-to-test data comparisons) are being performed. The overall goals of the comparison effort are (1) to provide code users with experience in applying their codes to LWR accident-sequence conditions and (2) to evaluate and improve the code models

  2. Comparison of neutron spectrum unfolding codes

    International Nuclear Information System (INIS)

    Zijp, W.

    1979-02-01

    This final report contains a set of four ECN-reports. The first is dealing with the comparison of the neutron spectrum unfolding codes CRYSTAL BALL, RFSP-JUL, SAND II and STAY'SL. The other three present the results of calculations about the influence of statistical weights in CRYSTAL BALL, SAND II and RFSP-JUL

  3. Comparison of computer code calculations with FEBA test data

    International Nuclear Information System (INIS)

    Zhu, Y.M.

    1988-06-01

    The FEBA forced feed reflood experiments included base line tests with unblocked geometry. The experiments consisted of separate effect tests on a full-length 5x5 rod bundle. Experimental cladding temperatures and heat transfer coefficients of FEBA test No. 216 are compared with the analytical data postcalculated utilizing the SSYST-3 computer code. The comparison indicates a satisfactory matching of the peak cladding temperatures, quench times and heat transfer coefficients for nearly all axial positions. This agreement was made possible by the use of an artificially adjusted value of the empirical code input parameter in the heat transfer for the dispersed flow regime. A limited comparison of test data and calculations using the RELAP4/MOD6 transient analysis code are also included. In this case the input data for the water entrainment fraction and the liquid weighting factor in the heat transfer for the dispersed flow regime were adjusted to match the experimental data. On the other hand, no fitting of the input parameters was made for the COBRA-TF calculations which are included in the data comparison. (orig.) [de

  4. 2-D fluid transport simulations of gaseous/radiative divertors

    International Nuclear Information System (INIS)

    Rognlien, T.D.; Brown, P.N.; Campbell, R.B.; Kaiser, T.B.; Knoll, D.A.; McHugh, P.R.; Porter, G.D.; Rensink, M.E.; Smith, G.R.

    1994-01-01

    The features of the fully implicit 2-D fluid code UEDGE are described. The utility of the code is demonstrated by showing bifurcations or multiple solutions of the tokamak edge plasma for both deuterium and impurity injection in the divertor. (orig.)

  5. ROLE OF NEUTRALS IN CORE FUELING AND PEDESTAL STRUCTURE IN H-MODE DIII-D DISCHARGES

    International Nuclear Information System (INIS)

    WOLF, NS; PETRIE, TW; PORTER, GD; ROGNLIEN, TD; GROEBNER, RJ; MAKOWSKI, MA

    2002-01-01

    OAK A271 ROLE OF NEUTRALS IN CORE FUELING AND PEDESTAL STRUCTURE IN H-MODE DIII-D DISCHARGES. The 2-D fluid code UEDGE was used to analyze DIII-D experiments to determine the role of neutrals in core fueling, core impurities, and also the H-mode pedestal structure. The authors compared the effects of divertor closure on the fueling rate and impurity density of high-triangularity, H-mode plasmas. UEDGE simulations indicate that the decrease in both deuterium core fueling (∼ 15%-20%) and core carbon density (∼ 15%-30%) with the closed divertor compared to the open divertor configuration is due to greater divertor screening of neutrals. They also compared UEDGE results with a simple analytic model of the H-mode pedestal structure. The model predicts both the width and gradient of the transport barrier in n e as a function of the pedestal density. The more sophisticated UEDGE simulations of H-mode discharges corroborate the simple analytic model, which is consistent with the hypothesis that fueling processes play a role in H-mode transport barrier formation

  6. THE EXPERIENCE OF COMPARISON OF STATIC SECURITY CODE ANALYZERS

    Directory of Open Access Journals (Sweden)

    Alexey Markov

    2015-09-01

    Full Text Available This work presents a methodological approach to comparison of static security code analyzers. It substantiates the comparison of the static analyzers as to efficiency and functionality indicators, which are stipulated in the international regulatory documents. The test data for assessment of static analyzers efficiency is represented by synthetic sets of open-source software, which contain vulnerabilities. We substantiated certain criteria for quality assessment of the static security code analyzers subject to standards NIST SP 500-268 and SATEC. We carried out experiments that allowed us to assess a number of the Russian proprietary software tools and open-source tools. We came to the conclusion that it is of paramount importance to develop Russian regulatory framework for testing software security (firstly, for controlling undocumented features and evaluating the quality of static security code analyzers.

  7. Comparison of Bit Error Rate of Line Codes in NG-PON2

    Directory of Open Access Journals (Sweden)

    Tomas Horvath

    2016-05-01

    Full Text Available This article focuses on simulation and comparison of line codes NRZ (Non Return to Zero, RZ (Return to Zero and Miller’s code for NG-PON2 (Next-Generation Passive Optical Network Stage 2 using. Our article provides solutions with Q-factor, BER (Bit Error Rate, and bandwidth comparison. Line codes are the most important part of communication over the optical fibre. The main role of these codes is digital signal representation. NG-PON2 networks use optical fibres for communication that is the reason why OptSim v5.2 is used for simulation.

  8. A comparison of LOCA analysis using SMOKIN and CERBERUS codes

    Energy Technology Data Exchange (ETDEWEB)

    Younis, M H [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Gaboury, G [Ontario Hydro, Toronto, ON (Canada)

    1996-12-31

    This paper presents the results of a comparison of the analyses of a postulated Loss of Coolant Accident (LOCA) in Pickering NGS A reactors using the two neutron kinetics codes SMOKIN and CERBERUS. Both codes have been used to simulate the space-time neutronic behaviour of CANDU-PHWR reactors. The main objective of the present study is to evaluate the accuracy with which SMOKIN can predict power transients compared to CERBERUS. The comparison shows that the two codes produce similar bulk power and reactivity transients. However, SMOKIN was found to overestimate the power transient (relative to CERBERUS) in some regions of the core, which is indicative of the spatial differences between the two codes. It was demonstrated that part of this overestimate is due to the use of reaction-rate averaged fuel properties in SMOKIN, compared to instantaneous fuel properties in CERBERUS. (author). 5 refs., 3 tabs., 6 figs.

  9. MOCCA Code for Star Cluster Simulation: Comparison with Optical Observations using COCOA

    OpenAIRE

    Askar, Abbas; Giersz, Mirek; Pych, Wojciech; Olech, Arkadiusz; Hypki, Arkadiusz

    2014-01-01

    We introduce and present preliminary results from COCOA (Cluster simulatiOn Comparison with ObservAtions) code for a star cluster after 12 Gyrs of evolution simulated using the MOCCA code. The COCOA code is being developed to quickly compare results of numerical simulations of star clusters with observational data. We use COCOA to obtain parameters of the projected cluster model. For comparison, a FITS file of the projected cluster was provided to observers so that they could use their observ...

  10. Data evaluation and code comparison activities

    International Nuclear Information System (INIS)

    Itikawa, Yukikazu; Takagi, Hidekazu; Nakamura, Yoshiharu; Imai, Makoto; Sasaki, Akira

    2013-01-01

    In atomic and molecular data base, intolerable numerical differences beyond error margin are found among some papers resulted from measurements or calculations even for the same collision processes. These differences spoil the reliability of the data base. This report describes the data evaluation for atomic and molecular data promoted by IAEA cooperated with other institutes, which Japanese researchers collaborate with. The reaction cross sections calculated numerically are evaluated for the collisions between electrons and molecular ions of H 2 + and HeH + . The application of an electron swarm parameter was shown for the evaluation and determination of the collision cross sections between electrons and molecules. In order to complete higher precision of atomic codes and a collisional-radiative model, IAEA held the workshop for the code comparison of the nonlocal thermodynamic equilibrium. (Y. Kazumata)

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

    Science.gov (United States)

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

    2002-08-01

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

  12. Comparison of elevated temperature design codes of ASME Subsection NH and RCC-MRx

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyeong-Yeon, E-mail: hylee@kaeri.re.kr

    2016-11-15

    Highlights: • Comparison of elevated temperature design (ETD) codes was made. • Material properties and evaluation procedures were compared. • Two heat-resistant materials of Grade 91 steel and austenitic stainless steel 316 are the target materials in the present study. • Application of the ETD codes to Generation IV reactor components and a comparison of the conservatism was conducted. - Abstract: The elevated temperature design (ETD) codes are used for the design evaluation of Generation IV (Gen IV) reactor systems such as sodium-cooled fast reactor (SFR), lead-cooled fast reactor (LFR), and very high temperature reactor (VHTR). In the present study, ETD code comparisons were made in terms of the material properties and design evaluation procedures for the recent versions of the two major ETD codes, ASME Section III Subsection NH and RCC-MRx. Conservatism in the design evaluation procedures was quantified and compared based on the evaluation results for SFR components as per the two ETD codes. The target materials are austenitic stainless steel 316 and Mod.9Cr-1Mo steel, which are the major two materials in a Gen IV SFR. The differences in the design evaluation procedures as well as the material properties in the two ETD codes are highlighted.

  13. WEC3: Wave Energy Converter Code Comparison Project: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Combourieu, Adrien; Lawson, Michael; Babarit, Aurelien; Ruehl, Kelley; Roy, Andre; Costello, Ronan; Laporte Weywada, Pauline; Bailey, Helen

    2017-01-01

    This paper describes the recently launched Wave Energy Converter Code Comparison (WEC3) project and present preliminary results from this effort. The objectives of WEC3 are to verify and validate numerical modelling tools that have been developed specifically to simulate wave energy conversion devices and to inform the upcoming IEA OES Annex VI Ocean Energy Modelling Verification and Validation project. WEC3 is divided into two phases. Phase 1 consists of a code-to-code verification and Phase II entails code-to-experiment validation. WEC3 focuses on mid-fidelity codes that simulate WECs using time-domain multibody dynamics methods to model device motions and hydrodynamic coefficients to model hydrodynamic forces. Consequently, high-fidelity numerical modelling tools, such as Navier-Stokes computational fluid dynamics simulation, and simple frequency domain modelling tools were not included in the WEC3 project.

  14. Simulation of single-phase rod bundle flow. Comparison between CFD-code ESTET, PWR core code THYC and experimental results

    International Nuclear Information System (INIS)

    Mur, J.; Larrauri, D.

    1998-07-01

    Computer simulation of flow in configurations close to pressurized water reactor (PWR) geometry is of great interest for Electricite de France (EDF). Although simulation of the flow through a whole PWR core with an all purpose CFD-code is not yet achievable, such a tool cna be quite useful to perform numerical experiments in order to try and improve the modeling introduced in computer codes devoted to reactor core thermal-hydraulic analysis. Further to simulation in small bare rod bundle configurations, the present study is focused on the simulation, with CFD-code ESTET and PWR core code THYC, of the flow in the experimental configuration VATICAN-1. ESTET simulation results are compared on the one hand to local velocity and concentration measurements, on the other hand with subchannel averaged values calculated by THYC. As far as the comparison with measurements is concerned, ESTET results are quite satisfactory relatively to available experimental data and their uncertainties. The effect of spacer grids and the prediction of the evolution of an unbalanced velocity profile seem to be correctly treated. As far as the comparison with THYC subchannel averaged values is concerned, the difficulty of a direct comparison between subchannel averaged and local values is pointed out. ESTET calculated local values are close to experimental local values. ESTET subchannel averaged values are also close to THYC calculation results. Thus, THYC results are satisfactory whereas their direct comparison to local measurements could show some disagreement. (author)

  15. Comparison of ANL containment codes with SNR-300 simulation experiments

    International Nuclear Information System (INIS)

    Marchertas, A.H.; Wang, C.Y.; Fistedis, S.H.

    1976-01-01

    A comparison of REXCO and ICECO code predictions is made with data obtained from experiments of LMFBR excursion models. The comparisons are based on published results of tests conducted for the safety analysis of the SNR-300 fast breeder. The test configurations consist of a centrally located spherical source immersed in a pool of water which is encased in a cylindrical container. The cylinical walls of the container are prestressed by holddown bolts which span the two rigid ends. The space above the surface of the water within the container is occupied by air. Although certain aspects of the tests could not be simulated by the analytical models exactly, the comparison of results shows quite close agreement. The fact that the REXCO and ICECO codes involve different analytical formulations, their own close correspondence of results lends added credence to the value of analytical predictions

  16. 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

  17. Extensions of the 3-dimensional plasma transport code E3D

    International Nuclear Information System (INIS)

    Runov, A.; Schneider, R.; Kasilov, S.; Reiter, D.

    2004-01-01

    One important aspect of modern fusion research is plasma edge physics. Fluid transport codes extending beyond the standard 2-D code packages like B2-Eirene or UEDGE are under development. A 3-dimensional plasma fluid code, E3D, based upon the Multiple Coordinate System Approach and a Monte Carlo integration procedure has been developed for general magnetic configurations including ergodic regions. These local magnetic coordinates lead to a full metric tensor which accurately accounts for all transport terms in the equations. Here, we discuss new computational aspects of the realization of the algorithm. The main limitation to the Monte Carlo code efficiency comes from the restriction on the parallel jump of advancing test particles which must be small compared to the gradient length of the diffusion coefficient. In our problems, the parallel diffusion coefficient depends on both plasma and magnetic field parameters. Usually, the second dependence is much more critical. In order to allow long parallel jumps, this dependence can be eliminated in two steps: first, the longitudinal coordinate x 3 of local magnetic coordinates is modified in such a way that in the new coordinate system the metric determinant and contra-variant components of the magnetic field scale along the magnetic field with powers of the magnetic field module (like in Boozer flux coordinates). Second, specific weights of the test particles are introduced. As a result of increased parallel jump length, the efficiency of the code is about two orders of magnitude better. (copyright 2004 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  18. MOCCA code for star cluster simulation: comparison with optical observations using COCOA

    Science.gov (United States)

    Askar, Abbas; Giersz, Mirek; Pych, Wojciech; Olech, Arkadiusz; Hypki, Arkadiusz

    2016-02-01

    We introduce and present preliminary results from COCOA (Cluster simulatiOn Comparison with ObservAtions) code for a star cluster after 12 Gyr of evolution simulated using the MOCCA code. The COCOA code is being developed to quickly compare results of numerical simulations of star clusters with observational data. We use COCOA to obtain parameters of the projected cluster model. For comparison, a FITS file of the projected cluster was provided to observers so that they could use their observational methods and techniques to obtain cluster parameters. The results show that the similarity of cluster parameters obtained through numerical simulations and observations depends significantly on the quality of observational data and photometric accuracy.

  19. A CFD code comparison of wind turbine wakes

    International Nuclear Information System (INIS)

    Van der Laan, M P; Sørensen, N N; Storey, R C; Cater, J E; Norris, S E

    2014-01-01

    A comparison is made between the EllipSys3D and SnS CFD codes. Both codes are used to perform Large-Eddy Simulations (LES) of single wind turbine wakes, using the actuator disk method. The comparison shows that both LES models predict similar velocity deficits and stream-wise Reynolds-stresses for four test cases. A grid resolution study, performed in EllipSys3D and SnS, shows that a minimal uniform cell spacing of 1/30 of the rotor diameter is necessary to resolve the wind turbine wake. In addition, the LES-predicted velocity deficits are also compared with Reynolds-Averaged Navier Stokes simulations using EllipSys3D for a test case that is based on field measurements. In these simulations, two eddy viscosity turbulence models are employed: the k-ε model and the k-ε-f p model. Where the k-ε model fails to predict the velocity deficit, the results of the k-ε-f P model show good agreement with both LES models and measurements

  20. The Second Workshop on Lineshape Code Comparison: Isolated Lines

    Directory of Open Access Journals (Sweden)

    Spiros Alexiou

    2014-05-01

    Full Text Available In this work, we briefly summarize the theoretical aspects of isolated line broadening. We present and discuss test run comparisons from different participating lineshape codes for the 2s-2p transition for LiI, B III and NV.

  1. The APS SASE FEL: modeling and code comparison

    International Nuclear Information System (INIS)

    Biedron, S. G.

    1999-01-01

    A self-amplified spontaneous emission (SASE) free-electron laser (FEL) is under construction at the Advanced Photon Source (APS). Five FEL simulation codes were used in the design phase: GENESIS, GINGER, MEDUSA, RON, and TDA3D. Initial comparisons between each of these independent formulations show good agreement for the parameters of the APS SASE FEL

  2. Comparison of DT neutron production codes MCUNED, ENEA-JSI source subroutine and DDT

    Energy Technology Data Exchange (ETDEWEB)

    Čufar, Aljaž, E-mail: aljaz.cufar@ijs.si [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Lengar, Igor; Kodeli, Ivan [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Milocco, Alberto [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Sauvan, Patrick [Departamento de Ingeniería Energética, E.T.S. Ingenieros Industriales, UNED, C/Juan del Rosal 12, 28040 Madrid (Spain); Conroy, Sean [VR Association, Uppsala University, Department of Physics and Astronomy, PO Box 516, SE-75120 Uppsala (Sweden); Snoj, Luka [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia)

    2016-11-01

    Highlights: • Results of three codes capable of simulating the accelerator based DT neutron generators were compared on a simple model where only a thin target made of mixture of titanium and tritium is present. Two typical deuteron beam energies, 100 keV and 250 keV, were used in the comparison. • Comparisons of the angular dependence of the total neutron flux and spectrum as well as the neutron spectrum of all the neutrons emitted from the target show general agreement of the results but also some noticeable differences. • A comparison of figures of merit of the calculations using different codes showed that the computational time necessary to achieve the same statistical uncertainty can vary for more than 30× when different codes for the simulation of the DT neutron generator are used. - Abstract: As the DT fusion reaction produces neutrons with energies significantly higher than in fission reactors, special fusion-relevant benchmark experiments are often performed using DT neutron generators. However, commonly used Monte Carlo particle transport codes such as MCNP or TRIPOLI cannot be directly used to analyze these experiments since they do not have the capabilities to model the production of DT neutrons. Three of the available approaches to model the DT neutron generator source are the MCUNED code, the ENEA-JSI DT source subroutine and the DDT code. The MCUNED code is an extension of the well-established and validated MCNPX Monte Carlo code. The ENEA-JSI source subroutine was originally prepared for the modelling of the FNG experiments using different versions of the MCNP code (−4, −5, −X) and was later extended to allow the modelling of both DT and DD neutron sources. The DDT code prepares the DT source definition file (SDEF card in MCNP) which can then be used in different versions of the MCNP code. In the paper the methods for the simulation of the DT neutron production used in the codes are briefly described and compared for the case of a

  3. Wind-induced transmission tower foundation loads. A field study-design code comparison

    Energy Technology Data Exchange (ETDEWEB)

    Savory, E. [Department of Mechanical and Materials Engineering, University of Western Ontario, London, Ont. (Canada); Parke, G.A.R.; Disney, P.; Toy, N. [School of Engineering, University of Surrey, Guildford, Surrey GU2 7XH (United Kingdom)

    2008-06-15

    This paper presents a comparison between the wind-induced foundation loads measured on a type L6 transmission line tower during a field study in the UK and those computed using the UK Code of Practice for lattice tower and transmission line design (BS8100). In this work, the Code provisions have been generalised to give the wind-induced strain in each of the tower legs immediately above the foundation as a function of wind direction and wind speed at the top of the tower. The complete data set from the field monitoring has been decomposed to provide a similar formulation for comparison purposes. The analysis shows excellent agreement between the Code calculations and the measured results, within the overall accuracy of the field data. This indicates that, at least for the tower type examined here, the existing design Code provides a reliable transformation of the local wind speed at the top of the tower into tension and compression loads on the foundations. (author)

  4. Pool swell sub-scale testing and code comparison

    International Nuclear Information System (INIS)

    Elisson, K.

    1981-01-01

    The main objective of the experiment was to investigate the pool swell dynamics in general and the forces on the lowered central part of the diaphragm between drywell and wetwell in particular. Apart from the high speed camera pressure transducers and strain gauges were used to monitor the transient. Data was recorded on a 14 channel FM recorder and then digitalised and plotted. In total more than one hundred tests were performed including parametric variations of for example geometry, break flow, initial drywell pressure and initial water level. In parallel to this experiment pool swell calculations have been performed with the computer codes COPTA and STEALTH. COPTA which is a lumped mass code for pressure suppression containment analysis has a slug pool swell mode. STEALTH which is a general purpose lagrangian hydrodynamics code has been used in a 2-D axisymmetric version. The STEALTH code has been used to calculate the radial variations in the vertical displacement and velocity of the pool surface and to predict the load on the lowered central part of the diaphragm. A comparison between the calculations and the experimental data indicates that both codes are sufficiently correct in their description of the pool swell transient. (orig.)

  5. Density effects on tokamak edge turbulence and transport with magnetic X-points

    International Nuclear Information System (INIS)

    Xu, X.Q.; Cohen, R.H.; Nevins, W.M.; Rognlien, T.D.; Ryutov, D.D.; Umansky, M.V.; Pearlstein, L.D.; Bulmer, R.H.; Russell, D.A.; Myra, J.R.; D'Ippolito, D.A.; Greenwald, M.; Snyder, P.B.; Mahdavi, M.A.

    2005-01-01

    Results are presented from the 3D electromagnetic turbulence code BOUT, the 2D transport code UEDGE, and theoretical analysis of boundary turbulence and transport in a real divertor-plasma geometry and its relationship to the density limit. Key results include: (1) a transition of the boundary turbulence from resistive X-point to resistive-ballooning as a critical plasma density is exceeded; (2) formation of an X-point MARFE in 2D UEDGE transport simulations for increasing outboard radial transport as found by BOUT for increasing density; (3) identification of convective transport by localized plasma 'blobs' in the SOL at high density during neutral fueling, and decorrelation of turbulence between the midplane and the divertor leg due to strong X-point magnetic shear; (4) a new divertor-leg instability driven at high plasma beta by a radial tilt of the divertor plate. (author)

  6. Comparison of design margin for core shroud in between design and construction code and fitness-for-service code

    International Nuclear Information System (INIS)

    Dozaki, Koji

    2007-01-01

    Structural design methods for core shroud of BWR are specified in JSME Design and Construction Code, like ASME Boiler and Pressure Vessel Code Sec. III, as a part of core support structure. Design margins are defined according to combination of the structural design method selected and service limit considered. Basically, those margins in JSME Code were determined after ASME Sec. III. Designers can select so-called twice-slope method for core shroud design among those design methods. On the other hand, flaw evaluation rules have been established for core shroud in JSME Fitness-for-Service Code. Twice-slope method is also adopted for fracture evaluation in that code even when the core shroud contains a flaw. Design margin was determined as structural factors separately from Design and Construction Code. As a natural consequence, there is a difference in those design margins between the two codes. In this paper, it is shown that the design margin in Fitness-for-Service Code is conservative by experimental evidences. Comparison of design margins between the two codes is discussed. (author)

  7. EG and G and NASA face seal codes comparison

    Science.gov (United States)

    Basu, Prit

    1994-01-01

    This viewgraph presentation presents the following results for the example comparison: EG&G code with face deformations suppressed and SPIRALG agree well with each other as well as with the experimental data; 0 rpm stiffness data calculated by EG&G code are about 70-100 percent lower than that by SPIRALG; there is no appreciable difference between 0 rpm and 16,000 rpm stiffness and damping coefficients calculated by SPIRALG; and the film damping above 500 psig calculated by SPIRALG is much higher than the O-Ring secondary seal damping (e.g. 50 lbf.s/in).

  8. Comparisons of 'Identical' Simulations by the Eulerian Gyrokinetic Codes GS2 and GYRO

    Science.gov (United States)

    Bravenec, R. V.; Ross, D. W.; Candy, J.; Dorland, W.; McKee, G. R.

    2003-10-01

    A major goal of the fusion program is to be able to predict tokamak transport from first-principles theory. To this end, the Eulerian gyrokinetic code GS2 was developed years ago and continues to be improved [1]. Recently, the Eulerian code GYRO was developed [2]. These codes are not subject to the statistical noise inherent to particle-in-cell (PIC) codes, and have been very successful in treating electromagnetic fluctuations. GS2 is fully spectral in the radial coordinate while GYRO uses finite-differences and ``banded" spectral schemes. To gain confidence in nonlinear simulations of experiment with these codes, ``apples-to-apples" comparisons (identical profile inputs, flux-tube geometry, two species, etc.) are first performed. We report on a series of linear and nonlinear comparisons (with overall agreement) including kinetic electrons, collisions, and shaped flux surfaces. We also compare nonlinear simulations of a DIII-D discharge to measurements of not only the fluxes but also the turbulence parameters. [1] F. Jenko, et al., Phys. Plasmas 7, 1904 (2000) and refs. therein. [2] J. Candy, J. Comput. Phys. 186, 545 (2003).

  9. Comparison of Two Commercial FE-Codes for Sheet Metal Forming

    International Nuclear Information System (INIS)

    Revuelta, A.; Larkiola, J.; Kanervo, K.; Korhonen, A. S.; Myllykoski, P.

    2007-01-01

    There is urgent need to develop new advanced fast and cost-effective mass-production methods for small sheet metal components. Traditionally progressive dies have been designed by using various CAD techniques. Recent results in mass production of small sheet metal parts using progressive dies and a transfer press showed that the tool design time may be cut in up to a half by using 3D finite element simulation of forming. In numerical simulation of sheet metal forming better constitutive models are required to obtain more accurate results, reduce the time for tool design and cut the production costs further. Accurate models are needed to describe the initial yielding, subsequent work hardening and to predict the formability. In this work two commercially available finite element simulation codes, PAM-STAMP and LS-DYNA, were compared in forming of small austenitic stainless steel sheet part for electronic industry. Several constitutive models were used in both codes and the results were compared. Comparisons were made between the same models in each of the codes and also between different models in the same code. Material models ranged from very simple to advanced ones, which took into account anisotropy and both isotropic and kinematic hardening behavior. In order to make a valid comparison we employed similar finite element meshes. The effects of the material models parameters were studied and the results were compared with experiments. The effects of the computational time were also studied

  10. Edge Plasma Simulations in NSTX and CTF: Synergy of Lithium Coating, Non-Diffusive Anomalous Transport and Drifts. Final Technical Report

    International Nuclear Information System (INIS)

    Pigarov, Alexander

    2012-01-01

    This is the final report for the Research Grant DE-FG02-08ER54989 'Edge Plasma Simulations in NSTX and CTF: Synergy of Lithium Coating, Non-Diffusive Anomalous Transport and Drifts'. The UCSD group including: A.Yu. Pigarov (PI), S.I. Krasheninnikov and R.D. Smirnov, was working on modeling of the impact of lithium coatings on edge plasma parameters in NSTX with the multi-species multi-fluid code UEDGE. The work was conducted in the following main areas: (i) improvements of UEDGE model for plasma-lithium interactions, (ii) understanding the physics of low-recycling divertor regime in NSTX caused by lithium pumping, (iii) study of synergistic effects with lithium coatings and non-diffusive ballooning-like cross-field transport, (iv) simulation of experimental multi-diagnostic data on edge plasma with lithium pumping in NSTX via self-consistent modeling of D-Li-C plasma with UEDGE, and (v) working-gas balance analysis. The accomplishments in these areas are given in the corresponding subsections in Section 2. Publications and presentations made under the Grant are listed in Section 3.

  11. Summary of aerosol code-comparison results for LWR aerosol containment tests LA1, LA2, and LA3

    International Nuclear Information System (INIS)

    Wright, A.L.; Wilson, J.H.; Arwood, P.C.

    1987-01-01

    The light-water reactor (LWR) aerosol containment experiments (LACE) are being performed in Richland, Washington, at the Hanford Engineering Development Laboratory under the leadership of an international project board and the Electric Power Research Institute. These tests have two objectives: (1) to investigate, at large scale, the inherent aerosol retention behavior in LWR containments under simulated severe accident conditions, and (2) to provide an experimental data base for validating aerosol behavior and thermal-hydraulic computer codes. Aerosol computer-code comparison activities for the LACE tests are being coordinated at the Oak Ridge National Laboratory. For each of the six experiments, pretest calculations (for code-to-code comparisons) and blind post-test calculations (for code-to-test data comparisons) are being performed. This paper presents a summary of the pretest aerosol-code results for tests LA1, LA2, and LA3

  12. Performance Comparison of Containment PT analysis between CAP and CONTEMPT Code

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Yeon Jun; Hong, Soon Joon; Hwang, Su Hyun; Kim, Min Ki; Lee, Byung Chul [FNC Tech., Seoul (Korea, Republic of); Ha, Sang Jun; Choi, Hoon [KHNP-CENTERAL RESEARCH INSTITUTE, Daejeon (Korea, Republic of)

    2013-10-15

    CAP, in the form that is linked with SPACE, computed the containment back-pressure during LOCA accident. In previous SAR (safety analysis report) report of Shin-Kori Units 3 and 4, the CONTEMPT series of codes(hereby referred to as just 'CONTEMPT') is used to evaluate the containment safety during the postulated loss-of-coolant accident (LOCA). In more detail, CONTEMPT-LT/028 was used to calculate the containment maximum PT, while CONTEMPT4/MOD5 to calculate the minimum PT. Actually, in minimum PT analysis, CONTEMPT4/MOD5, which provide back pressure condition of containment, was linked with RELAP5/MOD3.3 which calculate the amount of blowdown into containment. In this analysis, CONTEMPT4/MOD5 was modified based on KREM. CONTEMPT code was developed to predict the long term behavior of water-cooled nuclear reactor containment systems subjected to LOCA conditions. It calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and energy exchange with adjacent compartments, leakage on containment response. Models are provided for fan cooler and cooling spray as engineered safety systems. Any compartment may have both a liquid pool region and an air-vapor atmosphere region above the pool. Each region is assumed to have a uniform temperature, but the temperatures of the two regions may be different. As mentioned above, CONTEMP has the similar code features and it therefore is expected to show the similar analysis performance with CAP. In this study, the differences between CAP and two CONTEMPT code versions (CONTEMPT-LT/028 for maximum PT and CONTEMPT4/MOD5 for minimum PT) are, in detail, identified and the code performances were compared for the same problem. Code by code comparison was carried out to identify the difference of LOCA analysis between a series of COMTEMPT and CAP code. With regard to important factors that affect the transient behavior of compartment thermodynamic

  13. Performance Comparison of Containment PT analysis between CAP and CONTEMPT Code

    International Nuclear Information System (INIS)

    Choo, Yeon Jun; Hong, Soon Joon; Hwang, Su Hyun; Kim, Min Ki; Lee, Byung Chul; Ha, Sang Jun; Choi, Hoon

    2013-01-01

    CAP, in the form that is linked with SPACE, computed the containment back-pressure during LOCA accident. In previous SAR (safety analysis report) report of Shin-Kori Units 3 and 4, the CONTEMPT series of codes(hereby referred to as just 'CONTEMPT') is used to evaluate the containment safety during the postulated loss-of-coolant accident (LOCA). In more detail, CONTEMPT-LT/028 was used to calculate the containment maximum PT, while CONTEMPT4/MOD5 to calculate the minimum PT. Actually, in minimum PT analysis, CONTEMPT4/MOD5, which provide back pressure condition of containment, was linked with RELAP5/MOD3.3 which calculate the amount of blowdown into containment. In this analysis, CONTEMPT4/MOD5 was modified based on KREM. CONTEMPT code was developed to predict the long term behavior of water-cooled nuclear reactor containment systems subjected to LOCA conditions. It calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and energy exchange with adjacent compartments, leakage on containment response. Models are provided for fan cooler and cooling spray as engineered safety systems. Any compartment may have both a liquid pool region and an air-vapor atmosphere region above the pool. Each region is assumed to have a uniform temperature, but the temperatures of the two regions may be different. As mentioned above, CONTEMP has the similar code features and it therefore is expected to show the similar analysis performance with CAP. In this study, the differences between CAP and two CONTEMPT code versions (CONTEMPT-LT/028 for maximum PT and CONTEMPT4/MOD5 for minimum PT) are, in detail, identified and the code performances were compared for the same problem. Code by code comparison was carried out to identify the difference of LOCA analysis between a series of COMTEMPT and CAP code. With regard to important factors that affect the transient behavior of compartment thermodynamic state in

  14. Codification of LMFBR rules and comparison of codes

    International Nuclear Information System (INIS)

    Faure, O.; Debaene, J.P.

    1993-01-01

    The first part of this report presents the basic RCC-MR (regles de conception et de construction des materiels mecaniques des ilots nucleaires, reacteurs a neutrons rapides) design rules and their purpose. The second part is a qualitative comparison between RCC-MR, Code case N47 (ASME) and ETSDG Guide (MONJU Guide), made on the following topics: negligible creep test, ratcheting, creep fatigue, buckling, piping rules. An outline is given on improvements to RCC-MR rules now in progress

  15. The FRAM code: Description and some comparisons with MGA

    International Nuclear Information System (INIS)

    Sampson, T.E.; Kelley, T.A.

    1994-01-01

    The authors describe the initial development of the FRAM gamma-ray spectrometry code for analyzing plutonium isotopics, discuss its methodology, and present some comparisons with MGA on identical items. They also present some of the features of a new Windows 3.1-based version (PC/FRAM) and describe some current measurement problems. Development of the FRAM code began in about 1985, growing out of the need at the Los Alamos TA-55 Plutonium Facility for an isotopic analysis code to give accurate results for the effective specific power of heterogeneous (Am/Pu) pyrochemical residues. These residues present a difficult challenge because the americium is present mostly in a low-Z salt matrix (AmCl 3 ) with fines and small pieces of plutonium metal dispersed throughout the salt. Plutonium gamma rays suffer different attenuation than americium gamma rays of the same energy; this makes conventional analysis with a single relative efficiency function inaccurate for Am/Pu ratios and affects the analysis in other subtle ways

  16. 2D imaging of helium ion velocity in the DIII-D divertor

    Science.gov (United States)

    Samuell, C. M.; Porter, G. D.; Meyer, W. H.; Rognlien, T. D.; Allen, S. L.; Briesemeister, A.; Mclean, A. G.; Zeng, L.; Jaervinen, A. E.; Howard, J.

    2018-05-01

    Two-dimensional imaging of parallel ion velocities is compared to fluid modeling simulations to understand the role of ions in determining divertor conditions and benchmark the UEDGE fluid modeling code. Pure helium discharges are used so that spectroscopic He+ measurements represent the main-ion population at small electron temperatures. Electron temperatures and densities in the divertor match simulated values to within about 20%-30%, establishing the experiment/model match as being at least as good as those normally obtained in the more regularly simulated deuterium plasmas. He+ brightness (HeII) comparison indicates that the degree of detachment is captured well by UEDGE, principally due to the inclusion of E ×B drifts. Tomographically inverted Coherence Imaging Spectroscopy measurements are used to determine the He+ parallel velocities which display excellent agreement between the model and the experiment near the divertor target where He+ is predicted to be the main-ion species and where electron-dominated physics dictates the parallel momentum balance. Upstream near the X-point where He+ is a minority species and ion-dominated physics plays a more important role, there is an underestimation of the flow velocity magnitude by a factor of 2-3. These results indicate that more effort is required to be able to correctly predict ion momentum in these challenging regimes.

  17. A comparison of oxide thickness predictability from the perspective of codes

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo-Young; Shin, Hye-In; Kim, Kyung-Tae; Han, Hee-Tak; Kim, Hong-Jin; Kim, Yong-Hwan [KEPCO Nuclear Fuel Co. Ltd., Daejeon (Korea, Republic of)

    2016-10-15

    In Korea, OPR1000 and Westinghouse type nuclear power plant reactor fuel rods oxide thickness has been evaluated by imported code A. Because of this, there have been multiple constraints in operation and maintenance of fuel rod design system. For this reason, there has been a growing demand to establish an independent fuel rod design system. To meet this goal, KNF has recently developed its own code B for fuel rod design. The objective of this study is to compare oxide thickness prediction performance between code A and code B and to check the validity of predicting corrosion behaviors of newly developed code B. This study is based on Pool Side Examination (PSE) data for the performance confirmation. For the examination procedures, the oxide thickness measurement methods and equipment of PSE are described in detail. In this study, code B is confirmed conservatism and validity on evaluating cladding oxide thickness through the comparison with code A. Code prediction values show higher value than measured data from PSE. Throughout this study, the values by code B are evaluated and proved to be valid in a view point of the oxide thickness evaluation. However, the code B input for prediction has been made by designer's judgment with complex handwork that might be lead to excessive conservative result and ineffective design process with some possibility of errors.

  18. Comparisons of coded aperture imaging using various apertures and decoding methods

    International Nuclear Information System (INIS)

    Chang, L.T.; Macdonald, B.; Perez-Mendez, V.

    1976-07-01

    The utility of coded aperture γ camera imaging of radioisotope distributions in Nuclear Medicine is in its ability to give depth information about a three dimensional source. We have calculated imaging with Fresnel zone plate and multiple pinhole apertures to produce coded shadows and reconstruction of these shadows using correlation, Fresnel diffraction, and Fourier transform deconvolution. Comparisons of the coded apertures and decoding methods are made by evaluating their point response functions both for in-focus and out-of-focus image planes. Background averages and standard deviations were calculated. In some cases, background subtraction was made using combinations of two complementary apertures. Results using deconvolution reconstruction for finite numbers of events are also given

  19. Comparison of accelerator codes for a RHIC [Relativistic Heavy Ion Collider] lattice

    International Nuclear Information System (INIS)

    Milutinovic, J.; Ruggiero, A.G.

    1989-01-01

    We present the results of comparison of performances of several tracking or/and analysis codes. The basic purpose of this program was to assess reliability and accuracy of these codes, i.e., to determine the so-called ''error bars'' for the predicted values of tunes and other lattice functions as a minimum and, if possible, to discover potential difficulties with underlying physical models in these codes, inadequate algorithms, residual bugs and the like. Not only have we been able to determine the error bars, which for instance for the tunes at dp/p = +1% are Δν/sub ξ/ = 0.0027, Δν/sub y/ = 0.0010, but also our program has brought about improvements of several codes. 8 refs., 3 figs., 2 tabs

  20. Computerized Dental Comparison: A Critical Review of Dental Coding and Ranking Algorithms Used in Victim Identification.

    Science.gov (United States)

    Adams, Bradley J; Aschheim, Kenneth W

    2016-01-01

    Comparison of antemortem and postmortem dental records is a leading method of victim identification, especially for incidents involving a large number of decedents. This process may be expedited with computer software that provides a ranked list of best possible matches. This study provides a comparison of the most commonly used conventional coding and sorting algorithms used in the United States (WinID3) with a simplified coding format that utilizes an optimized sorting algorithm. The simplified system consists of seven basic codes and utilizes an optimized algorithm based largely on the percentage of matches. To perform this research, a large reference database of approximately 50,000 antemortem and postmortem records was created. For most disaster scenarios, the proposed simplified codes, paired with the optimized algorithm, performed better than WinID3 which uses more complex codes. The detailed coding system does show better performance with extremely large numbers of records and/or significant body fragmentation. © 2015 American Academy of Forensic Sciences.

  1. Automated Testing Infrastructure and Result Comparison for Geodynamics Codes

    Science.gov (United States)

    Heien, E. M.; Kellogg, L. H.

    2013-12-01

    The geodynamics community uses a wide variety of codes on a wide variety of both software and hardware platforms to simulate geophysical phenomenon. These codes are generally variants of finite difference or finite element calculations involving Stokes flow or wave propagation. A significant problem is that codes of even low complexity will return different results depending on the platform due to slight differences in hardware, software, compiler, and libraries. Furthermore, changes to the codes during development may affect solutions in unexpected ways such that previously validated results are altered. The Computational Infrastructure for Geodynamics (CIG) is funded by the NSF to enhance the capabilities of the geodynamics community through software development. CIG has recently done extensive work in setting up an automated testing and result validation system based on the BaTLab system developed at the University of Wisconsin, Madison. This system uses 16 variants of Linux and Mac platforms on both 32 and 64-bit processors to test several CIG codes, and has also recently been extended to support testing on the XSEDE TACC (Texas Advanced Computing Center) Stampede cluster. In this work we overview the system design and demonstrate how automated testing and validation occurs and results are reported. We also examine several results from the system from different codes and discuss how changes in compilers and libraries affect the results. Finally we detail some result comparison tools for different types of output (scalar fields, velocity fields, seismogram data), and discuss within what margins different results can be considered equivalent.

  2. Multi-dimensional free-electron laser simulation codes: a comparison study

    CERN Document Server

    Biedron, S G; Dejus, Roger J; Faatz, B; Freund, H P; Milton, S V; Nuhn, H D; Reiche, S

    2000-01-01

    A self-amplified spontaneous emission (SASE) free-electron laser (FEL) is under construction at the Advanced Photon Source (APS). Five FEL simulation codes were used in the design phase: GENESIS, GINGER, MEDUSA, RON, and TDA3D. Initial comparisons between each of these independent formulations show good agreement for the parameters of the APS SASE FEL.

  3. Multi-dimensional free-electron laser simulation codes: a comparison study

    International Nuclear Information System (INIS)

    Biedron, S. G.; Chae, Y. C.; Dejus, R. J.; Faatz, B.; Freund, H. P.; Milton, S. V.; Nuhn, H.-D.; Reiche, S.

    1999-01-01

    A self-amplified spontaneous emission (SASE) free-electron laser (FEL) is under construction at the Advanced Photon Source (APS). Five FEL simulation codes were used in the design phase: GENESIS, GINGER, MEDUSA, RON, and TDA3D. Initial comparisons between each of these independent formulations show good agreement for the parameters of the APS SASE FEL

  4. An assessment of ion temperature measurements in the boundary of the Alcator C-Mod tokamak and implications for ion fluid heat flux limiters

    International Nuclear Information System (INIS)

    Brunner, D; LaBombard, B; Churchill, R M; Hughes, J; Lipschultz, B; Ochoukov, R; Theiler, C; Walk, J; Rognlien, T D; Umansky, M V; Whyte, D

    2013-01-01

    The ion temperature is not frequently measured in the boundary of magnetic fusion devices. Comparisons among different ion temperature techniques and simulations are even rarer. Here we present a comparison of ion temperature measurements in the boundary of the Alcator C-Mod tokamak from three different diagnostics: charge exchange recombination spectroscopy (CXRS), an ion sensitive probe (ISP), and a retarding field analyzer (RFA). Comparison between CXRS and the ISP along with close examination of the ISP measurements reveals that the ISP is space charge limited. It is thus unable to measure ion temperature in the high density (>10 19 m −3 ) boundary plasma of C-Mod with its present geometry. Comparison of ion temperatures measured by CXRS and the RFA shows fair agreement. Ion and electron parallel heat flow is analyzed with a simple 1D fluid code. The code takes divertor measurements as input and results are compared to the measured ratios of upstream ion to electron temperature, as inferred respectively by CXRS and a Langmuir probe. The analysis reveals the limits of the fluid model at high Knudsen number. The upstream temperature ratio is under predicted by a factor of 2. Heat flux limiters (kinetic corrections) to the fluid model are necessary to match experimental data. The values required are found to be close to those reported in kinetic simulations. The 1D code is benchmarked against the 2D plasma fluid code UEDGE with good agreement. (paper)

  5. An assessment of ion temperature measurements in the boundary of the Alcator C-Mod tokamak and implications for ion fluid heat flux limiters

    Science.gov (United States)

    Brunner, D.; LaBombard, B.; Churchill, R. M.; Hughes, J.; Lipschultz, B.; Ochoukov, R.; Rognlien, T. D.; Theiler, C.; Walk, J.; Umansky, M. V.; Whyte, D.

    2013-09-01

    The ion temperature is not frequently measured in the boundary of magnetic fusion devices. Comparisons among different ion temperature techniques and simulations are even rarer. Here we present a comparison of ion temperature measurements in the boundary of the Alcator C-Mod tokamak from three different diagnostics: charge exchange recombination spectroscopy (CXRS), an ion sensitive probe (ISP), and a retarding field analyzer (RFA). Comparison between CXRS and the ISP along with close examination of the ISP measurements reveals that the ISP is space charge limited. It is thus unable to measure ion temperature in the high density (>1019 m-3) boundary plasma of C-Mod with its present geometry. Comparison of ion temperatures measured by CXRS and the RFA shows fair agreement. Ion and electron parallel heat flow is analyzed with a simple 1D fluid code. The code takes divertor measurements as input and results are compared to the measured ratios of upstream ion to electron temperature, as inferred respectively by CXRS and a Langmuir probe. The analysis reveals the limits of the fluid model at high Knudsen number. The upstream temperature ratio is under predicted by a factor of 2. Heat flux limiters (kinetic corrections) to the fluid model are necessary to match experimental data. The values required are found to be close to those reported in kinetic simulations. The 1D code is benchmarked against the 2D plasma fluid code UEDGE with good agreement.

  6. A comparison of two three-dimensional shell-element transient electromagnetics codes

    International Nuclear Information System (INIS)

    Yugo, J.J.; Williamson, D.E.

    1992-01-01

    Electromagnetic forces due to eddy currents strongly influence the design of components for the next generation of fusion devices. An effort has been made to benchmark two computer programs used to generate transient electromagnetic loads: SPARK and EddyCuFF. Two simple transient field problems were analyzed, both of which had been previously analyzed by the SPARK code with results recorded in the literature. A third problem that uses an ITER inboard blanket benchmark model was analyzed as well. This problem was driven with a self-consistent, distributed multifilament plasma model generated by an axisymmetric physics code. The benchmark problems showed good agreement between the two shell-element codes. Variations in calculated eddy currents of 1--3% have been found for similar, finely meshed models. A difference of 8% was found in induced current and 20% in force for a coarse mesh and complex, multifilament field driver. Because comparisons were made to results obtained from literature, model preparation and code execution times were not evaluated

  7. Comparison of secondary flows predicted by a viscous code and an inviscid code with experimental data for a turning duct

    Science.gov (United States)

    Schwab, J. R.; Povinelli, L. A.

    1984-01-01

    A comparison of the secondary flows computed by the viscous Kreskovsky-Briley-McDonald code and the inviscid Denton code with benchmark experimental data for turning duct is presented. The viscous code is a fully parabolized space-marching Navier-Stokes solver while the inviscid code is a time-marching Euler solver. The experimental data were collected by Taylor, Whitelaw, and Yianneskis with a laser Doppler velocimeter system in a 90 deg turning duct of square cross-section. The agreement between the viscous and inviscid computations was generally very good for the streamwise primary velocity and the radial secondary velocity, except at the walls, where slip conditions were specified for the inviscid code. The agreement between both the computations and the experimental data was not as close, especially at the 60.0 deg and 77.5 deg angular positions within the duct. This disagreement was attributed to incomplete modelling of the vortex development near the suction surface.

  8. Comparison of Zgoubi and S-Code regarding the FFAG muon acceleration

    International Nuclear Information System (INIS)

    Fourrier, J.; Machida, S.

    2006-06-01

    The high flux accelerator based neutrino source is foreseen as one of the next generation facilities of particle physics. Called Neutrino Factory (NuFact), it will be based on a muon storage ring where muons will decay, creating high flux neutrino beams. Muons are supposed to be accelerated from 5 to 20 GeV before being injected into the storage ring. In that purpose, Fixed Field Alternating Gradient accelerators (FFAG) are one of the possibilities. Cell designs have been done and tracking studies are on their way using codes such as MAD, S-Code or Zgoubi. In order to cross-check results so obtained, we have performed comparisons between S-Code and Zgoubi at Rutherford Appleton Laboratory. The present report will explain the different simulations done and the results. The paper has the following contents: 1. Introduction; 2. Time of Flight comparisons; 3. Particle acceleration with different emittances; 4. Bunch acceleration. In conclusion it is shown that the difference of time of flight between the particle 1 and the others is comparable from Zgoubi to S- Code. Nevertheless, further investigation is necessary to make sure that the same revolution time will be found with the same initial conditions. Concerning the acceleration it was found that that the larger the emittance the less efficient appears to be the acceleration. Particles on ellipses 10 to 40 π mm rad are accelerated beyond 9 GeV while particles on ellipse 50 π mm rad are only accelerated up to 7.5 GeV and those on 60 π mm rad ellipse just reach 6 GeV. Thus, a beam whom emittance would be larger than 30 π mm rad would not be accelerated enough to reach 10 GeV and to be injected into the second FFAG. In the same way as for the first FFAG the larger the emittance the less efficient the acceleration. Particles on ellipses 10 to 40 π mm rad are accelerated beyond 18 GeV while particles on ellipses 50 and 60 π mm rad are only accelerated up to 12 GeV. Thus, beams whom emittance is lower than 30 π mm

  9. Comparison of TITAN hybrid deterministic transport code and MCNP5 for simulation of SPECT

    International Nuclear Information System (INIS)

    Royston, K.; Haghighat, A.; Yi, C.

    2010-01-01

    Traditionally, Single Photon Emission Computed Tomography (SPECT) simulations use Monte Carlo methods. The hybrid deterministic transport code TITAN has recently been applied to the simulation of a SPECT myocardial perfusion study. The TITAN SPECT simulation uses the discrete ordinates formulation in the phantom region and a simplified ray-tracing formulation outside of the phantom. A SPECT model has been created in the Monte Carlo Neutral particle (MCNP)5 Monte Carlo code for comparison. In MCNP5 the collimator is directly modeled, but TITAN instead simulates the effect of collimator blur using a circular ordinate splitting technique. Projection images created using the TITAN code are compared to results using MCNP5 for three collimator acceptance angles. Normalized projection images for 2.97 deg, 1.42 deg and 0.98 deg collimator acceptance angles had maximum relative differences of 21.3%, 11.9% and 8.3%, respectively. Visually the images are in good agreement. Profiles through the projection images were plotted to find that the TITAN results followed the shape of the MCNP5 results with some differences in magnitude. A timing comparison on 16 processors found that the TITAN code completed the calculation 382 to 2787 times faster than MCNP5. Both codes exhibit good parallel performance. (author)

  10. Impact of dynamic rate coding aspects of mobile phone networks on forensic voice comparison.

    Science.gov (United States)

    Alzqhoul, Esam A S; Nair, Balamurali B T; Guillemin, Bernard J

    2015-09-01

    Previous studies have shown that landline and mobile phone networks are different in their ways of handling the speech signal, and therefore in their impact on it. But the same is also true of the different networks within the mobile phone arena. There are two major mobile phone technologies currently in use today, namely the global system for mobile communications (GSM) and code division multiple access (CDMA) and these are fundamentally different in their design. For example, the quality of the coded speech in the GSM network is a function of channel quality, whereas in the CDMA network it is determined by channel capacity (i.e., the number of users sharing a cell site). This paper examines the impact on the speech signal of a key feature of these networks, namely dynamic rate coding, and its subsequent impact on the task of likelihood-ratio-based forensic voice comparison (FVC). Surprisingly, both FVC accuracy and precision are found to be better for both GSM- and CDMA-coded speech than for uncoded. Intuitively one expects FVC accuracy to increase with increasing coded speech quality. This trend is shown to occur for the CDMA network, but, surprisingly, not for the GSM network. Further, in respect to comparisons between these two networks, FVC accuracy for CDMA-coded speech is shown to be slightly better than for GSM-coded speech, particularly when the coded-speech quality is high, but in terms of FVC precision the two networks are shown to be very similar. Copyright © 2015 The Chartered Society of Forensic Sciences. Published by Elsevier Ireland Ltd. All rights reserved.

  11. Combinatorial neural codes from a mathematical coding theory perspective.

    Science.gov (United States)

    Curto, Carina; Itskov, Vladimir; Morrison, Katherine; Roth, Zachary; Walker, Judy L

    2013-07-01

    Shannon's seminal 1948 work gave rise to two distinct areas of research: information theory and mathematical coding theory. While information theory has had a strong influence on theoretical neuroscience, ideas from mathematical coding theory have received considerably less attention. Here we take a new look at combinatorial neural codes from a mathematical coding theory perspective, examining the error correction capabilities of familiar receptive field codes (RF codes). We find, perhaps surprisingly, that the high levels of redundancy present in these codes do not support accurate error correction, although the error-correcting performance of receptive field codes catches up to that of random comparison codes when a small tolerance to error is introduced. However, receptive field codes are good at reflecting distances between represented stimuli, while the random comparison codes are not. We suggest that a compromise in error-correcting capability may be a necessary price to pay for a neural code whose structure serves not only error correction, but must also reflect relationships between stimuli.

  12. Comparison of computer codes related to the sodium oxide aerosol behavior in a containment building

    International Nuclear Information System (INIS)

    Fermandjian, J.

    1984-09-01

    In order to ensure that the problems of describing the physical behavior of sodium aerosols, during hypothetical fast reactor accidents, were adequately understood, a comparison of the computer codes (ABC/INTG, PNC, Japan; AEROSIM, UKAEA/SRD, United Kingdom; PARDISEKO IIIb, KfK, Germany; AEROSOLS/A2 and AEROSOLS/B1, CEA France) was undertaken in the frame of the CEC: exercise in which code users have run their own codes with a prearranged input

  13. Comparison of the calculations of the stability properties of a specific stellarator equilibrium with different MHD stability codes

    International Nuclear Information System (INIS)

    Nakamura, Y.; Matsumoto, T.; Wakatani, M.; Ichiguchi, K.; Garcia, L.; Carreras, B.A.

    1995-04-01

    A particular configuration of the LHD stellarator with an unusually flat pressure profile has been chosen to be a test case for comparison of the MHD stability property predictions of different three-dimensional and averaged codes for the purpose of code comparison and validation. In particular, two relatively localized instabilities, the fastest growing modes with toroidal mode number n = 2 and n = 3 were studied using several different codes, with the good agreement that has been found providing justification for the use of any of them for equilibria of the type considered

  14. Code-code comparisons of DIVIMP's 'onion-skin model' and the EDGE2D fluid code

    International Nuclear Information System (INIS)

    Stangeby, P.C.; Elder, J.D.; Horton, L.D.; Simonini, R.; Taroni, A.; Matthews, O.F.; Monk, R.D.

    1997-01-01

    In onion-skin modelling, O-SM, of the edge plasma, the cross-field power and particle flows are treated very simply e.g. as spatially uniform. The validity of O-S modelling requires demonstration that such approximations can still result in reasonable solutions for the edge plasma. This is demonstrated here by comparison of O-SM with full 2D fluid edge solutions generated by the EDGE2D code. The target boundary conditions for the O-SM are taken from the EDGE2D output and the complete O-SM solutions are then compared with the EDGE2D ones. Agreement is generally within 20% for n e , T e , T i and parallel particle flux density Γ for the medium and high recycling JET cases examined and somewhat less good for a strongly detached CMOD example. (orig.)

  15. Comparison of SISEC code simulations with earthquake data of ordinary and base-isolated buildings

    International Nuclear Information System (INIS)

    Wang, C.Y.; Gvildys, J.

    1991-01-01

    At Argonne National Laboratory (ANL), a 3-D computer program SISEC (Seismic Isolation System Evaluation Code) is being developed for simulating the system response of isolated and ordinary structures (Wang et al. 1991). This paper describes comparison of SISEC code simulations with building response data of actual earthquakes. To ensure the accuracy of analytical simulations, recorded data of full-size reinforced concrete structures located in Sendai, Japan are used in this benchmark comparison. The test structures consist of two three-story buildings, one base-isolated and the other one ordinary founded. They were constructed side by side to investigate the effect of base isolation on the acceleration response. Among 20 earthquakes observed since April 1989, complete records of three representative earthquakes, no.2, no.6, and no.17, are used for the code validation presented in this paper. Correlations of observed and calculated accelerations at all instrument locations are made. Also, relative response characteristics of ordinary and isolated building structures are investigated. (J.P.N.)

  16. Simulation of the scrape-off layer plasma during a disruption

    International Nuclear Information System (INIS)

    Rognlien, T.D.; Crotinger, J.A.; Porter, G.D.; Smith, G.R.; Kellman, A.G.; Taylor, P.L.

    1996-01-01

    The evolution of the scrape-off layer (SOL) during a disruption in the DIII-D tokamak is modeled using the 2-D UEDGE transport code. The focus is on the thermal quench phase when most of the energy content of the discharge is rapidly transported across the magnetic separatrix where it then flows to material surfaces or is radiated. Comparisons between the simulation and an experiment on the DIII-D tokamak are made with the heat flux to the divertor plate, and temperature and density profiles at the SOL midplane. The temporal response of the separate electron and ion heat-flux components to the divertor plate is calculated. The sensitivity of the solution to assumptions of electron heat-flux models and impurity radiation is investigated

  17. Plasma Interactions with Mixed Materials and Impurity Transport

    Energy Technology Data Exchange (ETDEWEB)

    Rognlien, T. D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Beiersdorfer, Peter [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Chernov, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Frolov, T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Magee, E. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Rudd, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Umansky, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-10-28

    The project brings together three discipline areas at LLNL to develop advanced capability to predict the impact of plasma/material interactions (PMI) on metallic surfaces in magnetic fusion energy (MFE) devices. These areas are (1) modeling transport of wall impurity ions through the edge plasma to the core plasma, (2) construction of a laser blow-off (LBO) system for injecting precise amounts of metallic atoms into a tokamak plasma, and (3) material science analysis of fundamental processes that modify metallic surfaces during plasma bombardment. The focus is on tungsten (W), which is being used for the ITER divertor and in designs of future MFE devices. In area (1), we have worked with the University of California, San Diego (UCSD) on applications of the UEDGE/DUSTT coupled codes to predict the influx of impurity ions from W dust through the edge plasma, including periodic edge-plasma oscillations, and revived a parallel version of UEDGE to speed up these simulations. In addition, the impurity transport model in the 2D UEDGE code has been implemented into the 3D BOUT++ turbulence/transport code to allow fundamental analysis of the impact of strong plasma turbulence on the impurity transport. In area (2), construction and testing of the LBO injection system has been completed. The original plan to install the LBO on the National Spherical Torus Experiment Upgrade (NSTX-U) at Princeton and its use to validate the impurity transport simulations is delayed owing to NSTX-U being offline for substantial magnetic coil repair period. In area (3), an analytic model has been developed to explain the growth of W tendrils (or fuzz) observed for helium-containing plasmas. Molecular dynamics calculations of W sputtering by W and deuterium (D) ions shows that a spatial blending of interatomic potentials is needed to describe the near-surface and deeper regions of the material.

  18. Plasma Interactions with Mixed Materials and Impurity Transport

    International Nuclear Information System (INIS)

    Rognlien, T. D.; Beiersdorfer, Peter; Chernov, A.; Frolov, T.; Magee, E.; Rudd, R.; Umansky, M.

    2016-01-01

    The project brings together three discipline areas at LLNL to develop advanced capability to predict the impact of plasma/material interactions (PMI) on metallic surfaces in magnetic fusion energy (MFE) devices. These areas are (1) modeling transport of wall impurity ions through the edge plasma to the core plasma, (2) construction of a laser blow-off (LBO) system for injecting precise amounts of metallic atoms into a tokamak plasma, and (3) material science analysis of fundamental processes that modify metallic surfaces during plasma bombardment. The focus is on tungsten (W), which is being used for the ITER divertor and in designs of future MFE devices. In area (1), we have worked with the University of California, San Diego (UCSD) on applications of the UEDGE/DUSTT coupled codes to predict the influx of impurity ions from W dust through the edge plasma, including periodic edge-plasma oscillations, and revived a parallel version of UEDGE to speed up these simulations. In addition, the impurity transport model in the 2D UEDGE code has been implemented into the 3D BOUT++ turbulence/transport code to allow fundamental analysis of the impact of strong plasma turbulence on the impurity transport. In area (2), construction and testing of the LBO injection system has been completed. The original plan to install the LBO on the National Spherical Torus Experiment Upgrade (NSTX-U) at Princeton and its use to validate the impurity transport simulations is delayed owing to NSTX-U being offline for substantial magnetic coil repair period. In area (3), an analytic model has been developed to explain the growth of W tendrils (or fuzz) observed for helium-containing plasmas. Molecular dynamics calculations of W sputtering by W and deuterium (D) ions shows that a spatial blending of interatomic potentials is needed to describe the near-surface and deeper regions of the material.

  19. ELECTRIC POTENIAL CELLS AT THE DIVERTED TOKAMAK SEPARATRIX

    International Nuclear Information System (INIS)

    SCHAFFER, M.J.; PORTER, G.D.; BOEDO, J.A.; BRAY, B.D.; HSIEH, C.L.; MOYER, R.A.; ROGNLIEN, T.D.; STANGEBY, P.C.; WATKINS, J.G.

    2000-01-01

    OAK-B135 Two-dimensional measurements by probes and Thomson scattering reveal unanticipated electric potential and electron pressure (p e ) maxima near the divertor X-point in L-mode plasmas in the DIII-D tokamak. The potential hill (∼ 50 V) drives E x B circulation (potential cell) of particles, energy and toroidal momentum around the X-point and in and out across the magnetic separatrix. Modeling by the UEDGE two-dimensional edge transport code with plasma drifts shows similar X-point potential and pressure hills. The code predicts additional drift-driven nonuniformity poloidally around the separatrix. Potential cells in UEDGE arise from parallel (to B) viscous stress acting on the Pfirsch-Schlueter ion return flow of the (del)B drift. These experimental and theoretical results demonstrate that the boundary layer just inside the separatrix of low power tokamak plasmas can be far from poloidal uniformity. They speculate that separatrix potential cells might be a major feature of L-mode edge transport and their suppression an important feature of H-mode

  20. Electric potential cells at the diverted tokamak separatrix

    Energy Technology Data Exchange (ETDEWEB)

    Schaffer, Michael J.; Bray, Bruce D.; Hsieh, Chung-Lih [General Atomics, San Diego, California (United States); Porter, Gary D.; Rognlien, Thomas D. [Lawrence Livermore National Laboratory, Livermore, California (United States); Boedo, Jose A.; Moyer, Richard A. [University of California, San Diego, California (United States); Stangeby, Peter C. [Univ. of Toronto, Toronto (Canada); Watkins, Jonathan G. [Sandia National Laboratories, Albuquerque, New Mexico (United States)

    2001-07-01

    Two-dimensional measurements by probes and Thomson scattering reveal unanticipated electric potential and electron pressure (p{sub e}) maxima near the divertor X-point in L-mode plasmas in the DIII-D tokamak. The potential hill ({approx}100 V) drives ExB circulation ('potential cell') of particles, energy and toroidal momentum around the X-point and in and out across the magnetic separatrix. Modeling by the UEDGE two-dimensional edge transport code with plasma drifts shows similar X-point potential and pressure hills. The code predicts additional drift-driven nonuniformity poloidally around the separatrix. Potential cells in UEDGE arises from parallel (to B) viscous stress acting on the Pfirsch-Schlueter ion return flow of the {nabla}B drift. These experimental and theoretical results demonstrate that the boundary layer just inside the separatrix of low power tokamak plasmas can be far from poloidal uniformity. We speculate that separatrix potential cells might be a major feature of L-mode edge transport and their suppression an important feature of H-mode. (author)

  1. Electric potential cells at the diverted tokamak separatrix

    International Nuclear Information System (INIS)

    Schaffer, Michael J.; Bray, Bruce D.; Hsieh, Chung-Lih; Porter, Gary D.; Rognlien, Thomas D.; Boedo, Jose A.; Moyer, Richard A.; Stangeby, Peter C.; Watkins, Jonathan G.

    2001-01-01

    Two-dimensional measurements by probes and Thomson scattering reveal unanticipated electric potential and electron pressure (p e ) maxima near the divertor X-point in L-mode plasmas in the DIII-D tokamak. The potential hill (∼100 V) drives ExB circulation ('potential cell') of particles, energy and toroidal momentum around the X-point and in and out across the magnetic separatrix. Modeling by the UEDGE two-dimensional edge transport code with plasma drifts shows similar X-point potential and pressure hills. The code predicts additional drift-driven nonuniformity poloidally around the separatrix. Potential cells in UEDGE arises from parallel (to B) viscous stress acting on the Pfirsch-Schlueter ion return flow of the ∇B drift. These experimental and theoretical results demonstrate that the boundary layer just inside the separatrix of low power tokamak plasmas can be far from poloidal uniformity. We speculate that separatrix potential cells might be a major feature of L-mode edge transport and their suppression an important feature of H-mode. (author)

  2. A comparison of the steady-state facility in the RELAP-UK code with the CUSH code and with collant flow in the Winfrith SGHWR

    International Nuclear Information System (INIS)

    Roberts, H.A.; Smith, C.P.

    1976-02-01

    Provision of capability for performing steady-state calculations in RELAP-UK has led to the possibility of the wider use of this code for steady-state assessments of the behaviour of commercial systems with complicated circuit arrangements. In the studies discussed in this report, the first objective is to demonstrate the performance of the RELAP-UK code in a steady-state role, and to make comparisons with the CUSH code, and with measurements obtained on the Winfrith Steam-Generating Heavy Water Reactor. (U.K.)

  3. Verification of three dimensional triangular prismatic discrete ordinates transport code ENSEMBLE-TRIZ by comparison with Monte Carlo code GMVP

    International Nuclear Information System (INIS)

    Homma, Y.; Moriwaki, H.; Ikeda, K.; Ohdi, S.

    2013-01-01

    This paper deals with the verification of the 3 dimensional triangular prismatic discrete ordinates transport calculation code ENSEMBLE-TRIZ by comparison with the multi-group Monte Carlo calculation code GMVP in a large fast breeder reactor. The reactor is a 750 MWe electric power sodium cooled reactor. Nuclear characteristics are calculated at the beginning of cycle of an initial core and at the beginning and the end of cycle of an equilibrium core. According to the calculations, the differences between the two methodologies are smaller than 0.0002 Δk in the multiplication factor, relatively about 1% in the control rod reactivity, and 1% in the sodium void reactivity. (authors)

  4. Comparison of results between different precision MAFIA codes

    International Nuclear Information System (INIS)

    Farkas, D.; Tice, B.

    1990-01-01

    In order to satisfy the inquiries of the Mafia Code users at SLAC, an evaluation of these codes was done. This consisted of running a cavity with known solutions. This study considered only the time independent solutions. No wake-field calculations were tried. The two machines involved were the NMFECC Cray (e-machine) at LLNL and the IBM/3081 at SLAC. The primary difference between the implementation of the codes on these machines is that the Cray has 64-bit accuracy while the IBM version has 32-bit accuracy. Unfortunately this study is incomplete as the Post-processor (P3) could not be made to work properly on the SLAC machine. This meant that no q's were calculated and no field patterns were generated. A certain amount of guessing had to be done when constructing the comparison tables. This problem aside, the probable conclusions that may be drawn are: (1) thirty-two bit precision is adequate for frequency determination; (2) sixty-four bit precision is desirable for field determination. This conclusion is deduced from the accuracy statistics. The cavity selected for study was a rectangular one with the dimensions (4,3,5) in centimeters. Only half of this cavity was used (2,3,5) with the x dimension being the one that was halved. The boundary conditions (B.C.) on the plane of symmetry were varied between Neumann and Dirichlet so as to cover all possible modes. Ten (10) modes were ran for each boundary condition

  5. Coding in pigeons: Multiple-coding versus single-code/default strategies.

    Science.gov (United States)

    Pinto, Carlos; Machado, Armando

    2015-05-01

    To investigate the coding strategies that pigeons may use in a temporal discrimination tasks, pigeons were trained on a matching-to-sample procedure with three sample durations (2s, 6s and 18s) and two comparisons (red and green hues). One comparison was correct following 2-s samples and the other was correct following both 6-s and 18-s samples. Tests were then run to contrast the predictions of two hypotheses concerning the pigeons' coding strategies, the multiple-coding and the single-code/default. According to the multiple-coding hypothesis, three response rules are acquired, one for each sample. According to the single-code/default hypothesis, only two response rules are acquired, one for the 2-s sample and a "default" rule for any other duration. In retention interval tests, pigeons preferred the "default" key, a result predicted by the single-code/default hypothesis. In no-sample tests, pigeons preferred the key associated with the 2-s sample, a result predicted by multiple-coding. Finally, in generalization tests, when the sample duration equaled 3.5s, the geometric mean of 2s and 6s, pigeons preferred the key associated with the 6-s and 18-s samples, a result predicted by the single-code/default hypothesis. The pattern of results suggests the need for models that take into account multiple sources of stimulus control. © Society for the Experimental Analysis of Behavior.

  6. Offshore Code Comparison Collaboration within IEA Wind Task 23: Phase IV Results Regarding Floating Wind Turbine Modeling; Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Jonkman, J.; Larsen, T.; Hansen, A.; Nygaard, T.; Maus, K.; Karimirad, M.; Gao, Z.; Moan, T.; Fylling, I.

    2010-04-01

    Offshore wind turbines are designed and analyzed using comprehensive simulation codes that account for the coupled dynamics of the wind inflow, aerodynamics, elasticity, and controls of the turbine, along with the incident waves, sea current, hydrodynamics, and foundation dynamics of the support structure. This paper describes the latest findings of the code-to-code verification activities of the Offshore Code Comparison Collaboration, which operates under Subtask 2 of the International Energy Agency Wind Task 23. In the latest phase of the project, participants used an assortment of codes to model the coupled dynamic response of a 5-MW wind turbine installed on a floating spar buoy in 320 m of water. Code predictions were compared from load-case simulations selected to test different model features. The comparisons have resulted in a greater understanding of offshore floating wind turbine dynamics and modeling techniques, and better knowledge of the validity of various approximations. The lessons learned from this exercise have improved the participants' codes, thus improving the standard of offshore wind turbine modeling.

  7. Comparison of LIFE-4 and TEMECH code predictions with TREAT transient test data

    International Nuclear Information System (INIS)

    Gneiting, B.C.; Bard, F.E.; Hunter, C.W.

    1984-09-01

    Transient tests in the TREAT reactor were performed on FFTF Reference design mixed-oxide fuel pins, most of which had received prior steady-state irradiation in the EBR-II reactor. These transient test results provide a data base for calibration and verification of fuel performance codes and for evaluation of processes that affect pin damage during transient events. This paper presents a comparison of the LIFE-4 and TEMECH fuel pin thermal/mechanical analysis codes with the results from 20 HEDL TREAT experiments, ten of which resulted in pin failure. Both the LIFE-4 and TEMECH codes provided an adequate representation of the thermal and mechanical data from the TREAT experiments. Also, a criterion for 50% probability of pin failure was developed for each code using an average cumulative damage fraction value calculated for the pins that failed. Both codes employ the two major cladding loading mechanisms of differential thermal expansion and central cavity pressurization which were demonstrated by the test results. However, a detailed evaluation of the code predictions shows that the two code systems weigh the loading mechanism differently to reach the same end points of the TREAT transient results

  8. A comparison of two nodal codes : Advanced nodal code (ANC) and analytic function expansion nodal (AFEN) code

    International Nuclear Information System (INIS)

    Chung, S.K.; Hah, C.J.; Lee, H.C.; Kim, Y.H.; Cho, N.Z.

    1996-01-01

    Modern nodal methods usually employs the transverse integration technique in order to reduce a multi-dimensional diffusion equation to one-dimensional diffusion equations. The use of the transverse integration technique requires two major approximations such as a transverse leakage approximation and a one-dimensional flux approximation. Both the transverse leakage and the one-dimensional flux are approximated by polynomials. ANC (Advanced Nodal Code) developed by Westinghouse employs a modern nodal expansion method for the flux calculation, the equivalence theory for the homogenization error reduction and a group theory for pin power recovery. Unlike the conventional modern nodal methods, AFEN (Analytic Function Expansion Nodal) method expands homogeneous flux distributions within a node into non-separable analytic basis functions, which eliminate two major approximations of the modern nodal methods. A comparison study of AFEN with ANC has been performed to see the applicability of AFEN to commercial PWR and different types of reactors such as MOX fueled reactor. The qualification comparison results demonstrate that AFEN methodology is accurate enough to apply for commercial PWR analysis. The results show that AFEN provides very accurate results (core multiplication factor and assembly power distribution) for cores that exhibit strong flux gradients as in a MOX loaded core. (author)

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

    International Nuclear Information System (INIS)

    Rossi, J.

    1994-02-01

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

  10. Offshore code comparison collaboration continuation (OC4), phase I - Results of coupled simulations of an offshore wind turbine with jacket support structure

    DEFF Research Database (Denmark)

    Popko, Wojciech; Vorpahl, Fabian; Zuga, Adam

    2012-01-01

    In this paper, the exemplary results of the IEA Wind Task 30 "Offshore Code Comparison Collaboration Continuation" (OC4) Project - Phase I, focused on the coupled simulation of an offshore wind turbine (OWT) with a jacket support structure, are presented. The focus of this task has been the verif......In this paper, the exemplary results of the IEA Wind Task 30 "Offshore Code Comparison Collaboration Continuation" (OC4) Project - Phase I, focused on the coupled simulation of an offshore wind turbine (OWT) with a jacket support structure, are presented. The focus of this task has been...... the verification of OWT modeling codes through code-to-code comparisons. The discrepancies between the results are shown and the sources of the differences are discussed. The importance of the local dynamics of the structure is depicted in the simulation results. Furthermore, attention is given to aspects...

  11. Offshore code comparison collaboration continuation within IEA Wind Task 30: Phase II results regarding a floating semisubmersible wind system

    DEFF Research Database (Denmark)

    Robertson, Amy; Jonkman, Jason M.; Vorpahl, Fabian

    2014-01-01

    Offshore wind turbines are designed and analyzed using comprehensive simulation tools (or codes) that account for the coupled dynamics of the wind inflow, aerodynamics, elasticity, and controls of the turbine, along with the incident waves, sea current, hydrodynamics, mooring dynamics, and founda......Offshore wind turbines are designed and analyzed using comprehensive simulation tools (or codes) that account for the coupled dynamics of the wind inflow, aerodynamics, elasticity, and controls of the turbine, along with the incident waves, sea current, hydrodynamics, mooring dynamics......, and foundation dynamics of the support structure. This paper describes the latest findings of the code-to-code verification activities of the Offshore Code Comparison Collaboration Continuation project, which operates under the International Energy Agency Wind Task 30. In the latest phase of the project......, participants used an assortment of simulation codes to model the coupled dynamic response of a 5-MW wind turbine installed on a floating semisubmersible in 200 m of water. Code predictions were compared from load case simulations selected to test different model features. The comparisons have resulted...

  12. Comparison of neutron diffusion theory codes in two and three space dimensions using a sodium cooled fast reactor benchmark

    International Nuclear Information System (INIS)

    Butland, A.T.D.; Putney, J.; Sweet, D.W.

    1980-04-01

    This report describes work performed to compare two UK neutron diffusion theory codes, TIGAR and SNAP, with published results for eight other codes available abroad. Both mesh edge and mesh centred finite difference diffusion theory codes as well as one axial synthesis code are included in the comparison and a range of iteration procedures are used by them. Comparison is made of calculations for a model of the sodium cooled fast reactor SNR-300 in both triangular and rectangular geometry and for a range of spatial meshes, enabling extrapolations to infinite mesh to be made. Calculated values of the effective multiplication constant, keff, for all the codes, agree very well when extrapolated to infinite mesh, indicating that no significant errors arising from the finite difference approximation but independent of mesh spacing are present in the calculations. The variation of keff with mesh area is found to be linear for the small meshes considered here, with the gradients for the mesh centred and mesh edged codes being of opposite sign. The results obtained using the mesh centred codes TIGAR, SNAP and CITATION agree closely with one another for all the meshes considered; the mesh edge codes agree less closely. (author)

  13. Results of aerosol code comparisons with releases from ACE MCCI tests

    International Nuclear Information System (INIS)

    Fink, J.K.; Corradini, M.; Hidaka, A.; Hontanon, E.; Mignanelli, M.A.; Schroedl, E.; Strizhov, V.

    1992-01-01

    Results of aerosol release calculations by six groups from six countries are compared with the releases from ACE MCCI Test L6. The codes used for these calculations included: SOLGASMIX-PV, SOLGASMIX Reactor 1986, CORCON.UW, VANESA 1.01, and CORCON mod2.04/VANESA 1.01. Calculations were performed with the standard VANESA 1.01 code and with modifications to the VANESA code such as the inclusion of various zirconium-silica chemical reactions. Comparisons of results from these calculations were made with Test L6 release fractions for U, Zr, Si, the fission-product elements Te, Ba, Sr, Ce, La, Mo and control materials Ag, In, and Ru. Reasonable agreement was obtained between calculations and Test L6 results for the volatile elements Ag, In and Te. Calculated releases of the low volatility fission products ranged from within an order of magnitude to five orders of magnitude of Test L6 values. Releases were over and underestimated by calculations. Poorest agreements were obtained for Mo and Si

  14. A comparison of two fully coupled codes for integrated dynamic analysis of floating vertical axis wind turbines

    NARCIS (Netherlands)

    Koppenol, Boy; Cheng, Zhengshun; Gao, Zhen; Simao Ferreira, C.; Moan, T; Tande, John Olav Giæver; Kvamsdal, Trond; Muskulus, Michael

    2017-01-01

    This paper presents a comparison of two state-of-the-art codes that are capable of modelling floating vertical axis wind turbines (VAWTs) in fully coupled time-domain simulations, being the HAWC2 by DTU and the SIMO-RIFLEX-AC code by NTNU/MARINTEK. The comparative study focusses on the way

  15. Fuel model studies. Comparison of our present version of GAPCON-THERMAL-2 with results from the EPRI code comparison study. Partial report

    International Nuclear Information System (INIS)

    Malen, K.; Jansson, L.

    1978-08-01

    Runs with our present version of GAPCON-THERMAL-2 have been compared to results from the EPRI code comparison study. Usually also our version of GAPCON predicts high temperatures, 100-300 K or 10-15% higher than average code predictions and experimental results. The well-known temperaturegas release instablility is found also with GAPCON. In this case one identifies the gas release limits 1400 deg C and 1700 deg C as instablility points. (author)

  16. Comparisons of the simulation results using different codes for ADS spallation target

    International Nuclear Information System (INIS)

    Yu Hongwei; Fan Sheng; Shen Qingbiao; Zhao Zhixiang; Wan Junsheng

    2002-01-01

    The calculations to the standard thick target were made by using different codes. The simulation of the thick Pb target with length of 60 cm, diameter of 20 cm bombarded with 800, 1000, 1500 and 2000 MeV energetic proton beam was carried out. The yields and the spectra of emitted neutron were studied. The spallation target was simulated by SNSP, SHIELD, DCM/CEM (Dubna Cascade Model /Cascade Evaporation Mode) and LAHET codes. The Simulation Results were compared with experiments. The comparisons show good agreement between the experiments and the SNSP simulated leakage neutron yield. The SHIELD simulated leakage neutron spectra are in good agreement with the LAHET and the DCM/CEM simulated leakage neutron spectra

  17. Reconciliation of international administrative coding systems for comparison of colorectal surgery outcome.

    Science.gov (United States)

    Munasinghe, A; Chang, D; Mamidanna, R; Middleton, S; Joy, M; Penninckx, F; Darzi, A; Livingston, E; Faiz, O

    2014-07-01

    Significant variation in colorectal surgery outcomes exists between different countries. Better understanding of the sources of variable outcomes using administrative data requires alignment of differing clinical coding systems. We aimed to map similar diagnoses and procedures across administrative coding systems used in different countries. Administrative data were collected in a central database as part of the Global Comparators (GC) Project. In order to unify these data, a systematic translation of diagnostic and procedural codes was undertaken. Codes for colorectal diagnoses, resections, operative complications and reoperative interventions were mapped across the respective national healthcare administrative coding systems. Discharge data from January 2006 to June 2011 for patients who had undergone colorectal surgical resections were analysed to generate risk-adjusted models for mortality, length of stay, readmissions and reoperations. In all, 52 544 case records were collated from 31 institutions in five countries. Mapping of all the coding systems was achieved so that diagnosis and procedures from the participant countries could be compared. Using the aligned coding systems to develop risk-adjusted models, the 30-day mortality rate for colorectal surgery was 3.95% (95% CI 0.86-7.54), the 30-day readmission rate was 11.05% (5.67-17.61), the 28-day reoperation rate was 6.13% (3.68-9.66) and the mean length of stay was 14 (7.65-46.76) days. The linkage of international hospital administrative data that we developed enabled comparison of documented surgical outcomes between countries. This methodology may facilitate international benchmarking. Colorectal Disease © 2014 The Association of Coloproctology of Great Britain and Ireland.

  18. Scrape-off layer plasma modeling for the DIII-D tokamak

    International Nuclear Information System (INIS)

    Porter, G.D.; Rognlien, T.D.; Allen, S.L.

    1994-09-01

    The behavior of the scrape-off layer (SOL) region in tokamaks is believed to play an important role determining the overall device performance. In addition, control of the exhaust power has become one of the most important issues in the design of future devices such as ITER and TPX. This paper presents the results of application of 2-D fluid models to the DII-D tokamak, and research into the importance of processes which are inadequately treated in the fluid models. Comparison of measured and simulated profiles of SOL plasma parameters suggest the physics model contained in the UEDGE code is sufficient to simulate plasmas which are attached to the divertor plates. Experimental evidence suggests the presence of enhanced plasma recombination and momentum removal leading to the existence of detached plasma states. UEDGE simulation of these plasmas obtains a bifurcation to a low temperature plasma at the divertor, but the plasma remains attached. Understanding the physics of this detachment is important for the design of future devices. Analytic studies of the behavior of SOL plasmas enhance our understanding beyond that achieved with fluid modeling. Analysis of the effect of drifts on sheath structure suggest these drifts may play a role in the detachment process. Analysis of the turbulent-transport equations indicate a bifurcation which is qualitatively similar to the experimentally different behavior of the L- and H-mode SOL. Electrostatic simulations of conducting wall modes suggest possible control of the SOL width by biasing

  19. Dual Coding, Reasoning and Fallacies.

    Science.gov (United States)

    Hample, Dale

    1982-01-01

    Develops the theory that a fallacy is not a comparison of a rhetorical text to a set of definitions but a comparison of one person's cognition with another's. Reviews Paivio's dual coding theory, relates nonverbal coding to reasoning processes, and generates a limited fallacy theory based on dual coding theory. (PD)

  20. Fracture flow code

    International Nuclear Information System (INIS)

    Dershowitz, W; Herbert, A.; Long, J.

    1989-03-01

    The hydrology of the SCV site will be modelled utilizing discrete fracture flow models. These models are complex, and can not be fully cerified by comparison to analytical solutions. The best approach for verification of these codes is therefore cross-verification between different codes. This is complicated by the variation in assumptions and solution techniques utilized in different codes. Cross-verification procedures are defined which allow comparison of the codes developed by Harwell Laboratory, Lawrence Berkeley Laboratory, and Golder Associates Inc. Six cross-verification datasets are defined for deterministic and stochastic verification of geometric and flow features of the codes. Additional datasets for verification of transport features will be documented in a future report. (13 figs., 7 tabs., 10 refs.) (authors)

  1. Comparison of ATF and TJ-II stellarator equilibria as computed by the 3-D VMEC and PIES codes

    International Nuclear Information System (INIS)

    Johnson, J.L.; Monticello, D.A.; Reiman, A.H.; Hirshman, S.P.

    1992-01-01

    A comparison is made of results from the PIES code, which determines the equilibrium properties of three-dimensional toroidal configurations by direct integration along the magnetic field lines, with those from the VMEC code, which uses an energy minimization in a flux representation to determine the equilibrium configuration, for two devices: the ATF stellarator at Oak Ridge and the TJ-11 heliac which is being built in Madrid. The results obtained from the two codes are in good agreement, providing additional validation for the codes

  2. A Comparison of Source Code Plagiarism Detection Engines

    Science.gov (United States)

    Lancaster, Thomas; Culwin, Fintan

    2004-06-01

    Automated techniques for finding plagiarism in student source code submissions have been in use for over 20 years and there are many available engines and services. This paper reviews the literature on the major modern detection engines, providing a comparison of them based upon the metrics and techniques they deploy. Generally the most common and effective techniques are seen to involve tokenising student submissions then searching pairs of submissions for long common substrings, an example of what is defined to be a paired structural metric. Computing academics are recommended to use one of the two Web-based detection engines, MOSS and JPlag. It is shown that whilst detection is well established there are still places where further research would be useful, particularly where visual support of the investigation process is possible.

  3. Comparison of aerosol behavior codes with experimental results from a sodium fire in a containment

    International Nuclear Information System (INIS)

    Lhiaubet, G.; Kissane, M.P.; Seino, H.; Miyake, O.; Himeno, Y.

    1990-01-01

    The containment expert group (CONT), a subgroup of the CEC fast reactor Safety Working Group (SWG), has carried out several studies on the behavior of sodium aerosols which might form in a severe fast reactor accident during which primary sodium leaks into the secondary containment. These studies comprise an intercalibration of measurement devices used to determine the aerosol particle size spectrum, and the analysis and comparison of codes applied to the determination of aerosol behavior in a reactor containment. The paper outlines the results of measurements of typical data made for aerosols produced in a sodium fire and their comparison with results from different codes (PARDISEKO, AEROSIM, CONTAIN, AEROSOLS/B2). The sodium fire experiment took place at CEN-Cadarache (France) in a 400 m 3 vessel. The fire lasted 90 minutes and the aerosol measurements were made over 10 hours at different locations inside the vessel. The results showed that the suspended mass calculated along the time with different codes was in good agreement with the experiment. However, the calculated aerosol deposition on the walls was diverging and always significantly lower than the measured values

  4. 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

  5. Experimental research and comparison of LDPC and RS channel coding in ultraviolet communication systems.

    Science.gov (United States)

    Wu, Menglong; Han, Dahai; Zhang, Xiang; Zhang, Feng; Zhang, Min; Yue, Guangxin

    2014-03-10

    We have implemented a modified Low-Density Parity-Check (LDPC) codec algorithm in ultraviolet (UV) communication system. Simulations are conducted with measured parameters to evaluate the LDPC-based UV system performance. Moreover, LDPC (960, 480) and RS (18, 10) are implemented and experimented via a non-line-of-sight (NLOS) UV test bed. The experimental results are in agreement with the simulation and suggest that based on the given power and 10(-3)bit error rate (BER), in comparison with an uncoded system, average communication distance increases 32% with RS code, while 78% with LDPC code.

  6. SPACE CHARGE SIMULATION METHODS INCORPORATED IN SOME MULTI - PARTICLE TRACKING CODES AND THEIR RESULTS COMPARISON

    International Nuclear Information System (INIS)

    BEEBE - WANG, J.; LUCCIO, A.U.; D IMPERIO, N.; MACHIDA, S.

    2002-01-01

    Space charge in high intensity beams is an important issue in accelerator physics. Due to the complicity of the problems, the most effective way of investigating its effect is by computer simulations. In the resent years, many space charge simulation methods have been developed and incorporated in various 2D or 3D multi-particle-tracking codes. It has becoming necessary to benchmark these methods against each other, and against experimental results. As a part of global effort, we present our initial comparison of the space charge methods incorporated in simulation codes ORBIT++, ORBIT and SIMPSONS. In this paper, the methods included in these codes are overviewed. The simulation results are presented and compared. Finally, from this study, the advantages and disadvantages of each method are discussed

  7. SPACE CHARGE SIMULATION METHODS INCORPORATED IN SOME MULTI - PARTICLE TRACKING CODES AND THEIR RESULTS COMPARISON.

    Energy Technology Data Exchange (ETDEWEB)

    BEEBE - WANG,J.; LUCCIO,A.U.; D IMPERIO,N.; MACHIDA,S.

    2002-06-03

    Space charge in high intensity beams is an important issue in accelerator physics. Due to the complicity of the problems, the most effective way of investigating its effect is by computer simulations. In the resent years, many space charge simulation methods have been developed and incorporated in various 2D or 3D multi-particle-tracking codes. It has becoming necessary to benchmark these methods against each other, and against experimental results. As a part of global effort, we present our initial comparison of the space charge methods incorporated in simulation codes ORBIT++, ORBIT and SIMPSONS. In this paper, the methods included in these codes are overviewed. The simulation results are presented and compared. Finally, from this study, the advantages and disadvantages of each method are discussed.

  8. Development of integrated SOL/Divertor code and simulation study of the JT-60U/JT-60SA tokamaks

    International Nuclear Information System (INIS)

    Kawashima, H.; Shimizu, K.; Takizuka, T.

    2007-01-01

    To predict the particle and heat controllability in the divertor of tokamak reactors such as ITER and to optimize the divertor design, comprehensive simulations by integrated modelling with taking in various physical processes are indispensable. For the design study of ITER divertor, the modelling codes such as B2, UEDGE and EDGE2D have been developed, and their results have contributed to the evolution of the divertor concept. In Japan Atomic Energy Agency (JAEA), SOL/divertor codes have also been developed for the interpretation and the prediction on behaviours of plasmas, neutrals and impurities in the SOL/divertor regions. The code development is originally carried out since physics models can be verified quickly and flexibly under the circumstance of close collaboration with JT-60 team. Figure 1 shows our code system, which consists of the 2 dimensional fluid code SOLDOR, the neutral Monte Carlo (MC) code NEUT2D, and the impurity MC code IMPMC. The particle simulation code PARASOL has also been developed in order to establish the physics modelling used in fluid simulations. Integration of SOLDOR, NEUT2D and IMPMC, called the '' SONIC '' code, is being carried out to simulate self-consistently the SOL/divertor plasmas in present tokamaks and in future devices. Combination of the SOLDOR and NEUT2D was completed, which has the features such as 1) high-resolution oscillation-free scheme in solving fluid equations, 2) neutral transport calculation under the fine meshes, 3) success in reduction of MC noise, 4) optimization on the massive parallel computer, etc. The simulation reproduces the X-point MARFE in the JT-60U experiment. It is found that the chemically sputtered carbon at the dome causes the radiation peaking near the X-point. The performance of divertor pumping in JT-60U is evaluated from the particle balances. We also present the divertor designing of JT-60SA, which is the modification program of JT-60U to establish high beta steady-state operation. To

  9. Comparison of two LES codes for wind turbine wake studies

    International Nuclear Information System (INIS)

    Sarlak, H; Mikkelsen, R; Sørensen, J N; Pierella, F

    2014-01-01

    For the third time a blind test comparison in Norway 2013, was conducted comparing numerical simulations for the rotor C p and C t and wake profiles with the experimental results. As the only large eddy simulation study among participants, results of the Technical University of Denmark (DTU) using their in-house CFD solver, EllipSys3D, proved to be more reliable among the other models for capturing the wake profiles and the turbulence intensities downstream the turbine. It was therefore remarked in the workshop to investigate other LES codes to compare their performance with EllipSys3D. The aim of this paper is to investigate on two CFD solvers, the DTU's in-house code, EllipSys3D and the open-sourse toolbox, OpenFoam, for a set of actuator line based LES computations. Two types of simulations are performed: the wake behind a signle rotor and the wake behind a cluster of three inline rotors. Results are compared in terms of velocity deficit, turbulence kinetic energy and eddy viscosity. It is seen that both codes predict similar near-wake flow structures with the exception of OpenFoam's simulations without the subgrid-scale model. The differences begin to increase with increasing the distance from the upstream rotor. From the single rotor simulations, EllipSys3D is found to predict a slower wake recovery in the case of uniform laminar flow. From the 3-rotor computations, it is seen that the difference between the codes is smaller as the disturbance created by the downstream rotors causes break down of the wake structures and more homogenuous flow structures. It is finally observed that OpenFoam computations are more sensitive to the SGS models

  10. The fast code

    Energy Technology Data Exchange (ETDEWEB)

    Freeman, L.N.; Wilson, R.E. [Oregon State Univ., Dept. of Mechanical Engineering, Corvallis, OR (United States)

    1996-09-01

    The FAST Code which is capable of determining structural loads on a flexible, teetering, horizontal axis wind turbine is described and comparisons of calculated loads with test data are given at two wind speeds for the ESI-80. The FAST Code models a two-bladed HAWT with degrees of freedom for blade bending, teeter, drive train flexibility, yaw, and windwise and crosswind tower motion. The code allows blade dimensions, stiffnesses, and weights to differ and models tower shadow, wind shear, and turbulence. Additionally, dynamic stall is included as are delta-3 and an underslung rotor. Load comparisons are made with ESI-80 test data in the form of power spectral density, rainflow counting, occurrence histograms, and azimuth averaged bin plots. It is concluded that agreement between the FAST Code and test results is good. (au)

  11. German nuclear codes revised: comparison with approaches used in other countries

    International Nuclear Information System (INIS)

    Raetzke, C.; Micklinghoff, M.

    2005-01-01

    The article deals with the plan of the German Federal Ministry for the Environment (BMU) to revise the German set of nuclear codes, and draws a comparison with approaches pursued in other countries in formulating and implementing new requirements imposed upon existing plants. A striking feature of the BMU project is the intention to have the codes reflect the state of the art in an entirely abstract way irrespective of existing plants. This implies new requirements imposed on plant design, among other things. However, the state authorities, which establish the licensing conditions for individual plants in concrete terms, will not be able to apply these new codes for legal reasons (protection of vested rights) to the extent in which they incorporate changes in safety philosophy. Also the procedure adopted has raised considerable concern. The processing time of two years is inordinately short, and participation of the public and of industry does not go beyond the strictly formal framework of general public participation. In the light of this absence of quality assurance, it would be surprising if this new set of codes did not suffer from considerable deficits in its contents. Other countries show that the BMU is embarking on an isolated approach in every respect. Elsewhere, backfitting requirements are developed carefully and over long periods of time; they are discussed in detail with the operators; costs and benefits are weighted, and the consequences are evaluated. These elements are in common to procedures in all countries, irrespective of very different steps in detail. (orig.)

  12. Code Glosses in Academic Writing: The Comparison of Iranian and Native Authors

    Directory of Open Access Journals (Sweden)

    Mahdi Dehghan

    2016-01-01

    Full Text Available One of the challenges of academic writing is the creation of a structurally and communicatively well-organized and coherent text. Metadiscourse enables authors of journal articles to achieve this goal by raising the writers‘ awareness about discourse features which can contribute to a better academic content production. Also, L1 background has always been a hot topic in applied linguistics and native versus non-native comparisons have been of particular interest in this field. The purpose of this study was to see whether native English speakers and Iranians use code glosses as a sub-category of metadiscourse similarly in their academic writings. To this end, the introduction section of 30 journal articles written by native and Iranian non-native English authors were investigated and the number of code-glosses in each group was counted and analyzed. Both descriptive and inferential statistics were used to answer the research question of this study. The result of data analysis revealed that there was no significant difference between the frequency count of code glosses used by Iranian and native English authors. This study can have pedagogical implications for EAP course designers as well as academic writing instructors and students.

  13. Analysis of SCARABEE BE+3 experiment with ASTEC-Na and comparison with other SFR safety analysis codes

    International Nuclear Information System (INIS)

    Bandini, Giacomino; Ederli, Stefano; Perez-Martin, Sara; Pfrang, Werner; Girault, Nathalie; Cloarec, Laure

    2017-01-01

    The ASTEC-Na code was further developed and assessed in the frame of JASMIN project of the 7th EU Framework Program to extend the original capability of ASTEC, dealing with severe accident analysis in LWR to Sodium-cooled Fast Reactors (SFR). The in-pile BE+3 experiment from the SCARABEE-N program has been simulated with ASTEC-Na for thermal-hydraulic models validation purpose. The adequacy of ASTEC-Na thermal-hydraulic models has been also investigated through the comparison with other safety analysis codes. The analysis of SCARABEE BE+3 test confirms the good performance of ASTEC-Na code in the calculation of single-phase conditions and boiling onset, while larger deviations are encountered in the analysis of the two-phase conditions, mainly regarding the propagation of the boiling front. Furthermore, reasonable agreement was found with other code results. (author)

  14. Comparison of Heavy Water Reactor Thermalhydraulic Code Predictions with Small Break LOCA Experimental Data

    International Nuclear Information System (INIS)

    2012-08-01

    Activities within the frame of the IAEA's Technical Working Group on Advanced Technologies for HWRs (TWG-HWR) are conducted in a project within the IAEA's subprogramme on nuclear power reactor technology development. The objective of the activities on HWRs is to foster, within the frame of the TWG-HWR, information exchange and cooperative research on technology development for current and future HWRs, with an emphasis on safety, economics and fuel resource sustainability. One of the activities recommended by the TWG-HWR was an international standard problem exercise entitled Intercomparison and Validation of Computer Codes for Thermalhydraulics Safety Analyses. Intercomparison and validation of computer codes used in different countries for thermalhydraulics safety analyses will enhance the confidence in the predictions made by these codes. However, the intercomparison and validation exercise needs a set of reliable experimental data. Two RD-14M small break loss of coolant accident (SBLOCA) tests, simulating HWR LOCA behaviour, conducted by Atomic Energy of Canada Ltd (AECL), were selected for this validation project. This report provides a comparison of the results obtained from eight participating organizations from six countries (Argentina, Canada, China, India, Republic of Korea, and Romania), utilizing four different computer codes (ATMIKA, CATHENA, MARS-KS, and RELAP5). General conclusions are reached and recommendations made.

  15. Comparison of the thermal neutron scattering treatment in MCNP6 and GEANT4 codes

    Science.gov (United States)

    Tran, H. N.; Marchix, A.; Letourneau, A.; Darpentigny, J.; Menelle, A.; Ott, F.; Schwindling, J.; Chauvin, N.

    2018-06-01

    To ensure the reliability of simulation tools, verification and comparison should be made regularly. This paper describes the work performed in order to compare the neutron transport treatment in MCNP6.1 and GEANT4-10.3 in the thermal energy range. This work focuses on the thermal neutron scattering processes for several potential materials which would be involved in the neutron source designs of Compact Accelerator-based Neutrons Sources (CANS), such as beryllium metal, beryllium oxide, polyethylene, graphite, para-hydrogen, light water, heavy water, aluminium and iron. Both thermal scattering law and free gas model, coming from the evaluated data library ENDF/B-VII, were considered. It was observed that the GEANT4.10.03-patch2 version was not able to account properly the coherent elastic process occurring in crystal lattice. This bug is treated in this work and it should be included in the next release of the code. Cross section sampling and integral tests have been performed for both simulation codes showing a fair agreement between the two codes for most of the materials except for iron and aluminium.

  16. Comparison calculations of WWER-1000 fuel assemblies by using the MCNP 4.2 a KASSETA codes

    International Nuclear Information System (INIS)

    Trgina, M.

    1993-12-01

    The power multiplication and distribution factors are compared for various geometries and material configurations of WWER-1000 fuel assemblies. The calculations were performed in 2 ways: (i) using nuclear data, employing older and current data collections, and (ii) using the author's own model based on the KASSETA code. The comparison code MCNP 4.2 is described, intended for computerized simulation of the transport of neutrons, photons and electrons. This code uses its own cross section library. The methodology is outlined and a specification of the Monte Carlo method employed is given. The use of the refined data library gave rise to appreciable deviations of the multiplication factors in all variants. The use of the older data library led to identical criticality results for the variant with water holes. For inserted absorbers the discrepancies in criticality and in power distribution data are appreciable. The marked disagreement between the results of application of the MCNP 4.2 and KASSETA codes for the variants with inserted control elements is indicative of inappropriateness of the approximation procedure in the latter code. (J.B.). 2 tabs., 11 figs., 11 refs

  17. Comparison of the THYC and FLICA-3M codes by the pseudo-cubic thin-plate method

    International Nuclear Information System (INIS)

    Banner, D.; Crecy, F. de.

    1993-06-01

    The pseudo cubic Spline method (PCSM) is a statistical tool developed by the CEA. It is designed to analyse experimental points and in particular thermalhydraulic data. Predictors of the occurrence of critical heat flux are obtained by using Spline functions. In this paper, predictors have been computed from the same CHF databases by using two different flow analyses to derive local thermal-hydraulic variables at the CHF location. In fact, CEA's FLICA-3M represents rod bundles by interconnected subchannels whereas EDF's THYC code uses a porous 3D approach. In a first step, the PCSM is briefly presented as well as the two codes studied here. Then, the comparison methodology is explained in order to prove that advanced analysis of thermalhydraulic codes can be achieved with the PCSM. (authors). 6 figs., 2 tabs., 5 refs

  18. Benchmark Problems of the Geothermal Technologies Office Code Comparison Study

    Energy Technology Data Exchange (ETDEWEB)

    White, Mark D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Podgorney, Robert [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Kelkar, Sharad M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); McClure, Mark W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Danko, George [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ghassemi, Ahmad [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fu, Pengcheng [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bahrami, Davood [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Barbier, Charlotte [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cheng, Qinglu [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Chiu, Kit-Kwan [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Detournay, Christine [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Elsworth, Derek [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fang, Yi [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Furtney, Jason K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Gan, Quan [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Gao, Qian [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Guo, Bin [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hao, Yue [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Horne, Roland N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Huang, Kai [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Im, Kyungjae [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Norbeck, Jack [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rutqvist, Jonny [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Safari, M. R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sesetty, Varahanaresh [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sonnenthal, Eric [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Tao, Qingfeng [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); White, Signe K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wong, Yang [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Xia, Yidong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-12-02

    A diverse suite of numerical simulators is currently being applied to predict or understand the performance of enhanced geothermal systems (EGS). To build confidence and identify critical development needs for these analytical tools, the United States Department of Energy, Geothermal Technologies Office has sponsored a Code Comparison Study (GTO-CCS), with participants from universities, industry, and national laboratories. A principal objective for the study was to create a community forum for improvement and verification of numerical simulators for EGS modeling. Teams participating in the study were those representing U.S. national laboratories, universities, and industries, and each team brought unique numerical simulation capabilities to bear on the problems. Two classes of problems were developed during the study, benchmark problems and challenge problems. The benchmark problems were structured to test the ability of the collection of numerical simulators to solve various combinations of coupled thermal, hydrologic, geomechanical, and geochemical processes. This class of problems was strictly defined in terms of properties, driving forces, initial conditions, and boundary conditions. Study participants submitted solutions to problems for which their simulation tools were deemed capable or nearly capable. Some participating codes were originally developed for EGS applications whereas some others were designed for different applications but can simulate processes similar to those in EGS. Solution submissions from both were encouraged. In some cases, participants made small incremental changes to their numerical simulation codes to address specific elements of the problem, and in other cases participants submitted solutions with existing simulation tools, acknowledging the limitations of the code. The challenge problems were based on the enhanced geothermal systems research conducted at Fenton Hill, near Los Alamos, New Mexico, between 1974 and 1995. The problems

  19. Inter-comparison of Computer Codes for TRISO-based Fuel Micro-Modeling and Performance Assessment

    International Nuclear Information System (INIS)

    Boer, Brian; Keun Jo, Chang; Wu, Wen; Ougouag, Abderrafi M.; McEachren, Donald; Venneri, Francesco

    2010-01-01

    The Next Generation Nuclear Plant (NGNP), the Deep Burn Pebble Bed Reactor (DB-PBR) and the Deep Burn Prismatic Block Reactor (DB-PMR) are all based on fuels that use TRISO particles as their fundamental constituent. The TRISO particle properties include very high durability in radiation environments, hence the designs reliance on the TRISO to form the principal barrier to radioactive materials release. This durability forms the basis for the selection of this fuel type for applications such as Deep Bun (DB), which require exposures up to four times those expected for light water reactors. It follows that the study and prediction of the durability of TRISO particles must be carried as part of the safety and overall performance characterization of all the designs mentioned above. Such evaluations have been carried out independently by the performers of the DB project using independently developed codes. These codes, PASTA, PISA and COPA, incorporate models for stress analysis on the various layers of the TRISO particle (and of the intervening matrix material for some of them), model for fission products release and migration then accumulation within the SiC layer of the TRISO particle, just next to the layer, models for free oxygen and CO formation and migration to the same location, models for temperature field modeling within the various layers of the TRISO particle and models for the prediction of failure rates. All these models may be either internal to the code or external. This large number of models and the possibility of different constitutive data and model formulations and the possibility of a variety of solution techniques makes it highly unlikely that the model would give identical results in the modeling of identical situations. The purpose of this paper is to present the results of an inter-comparison between the codes and to identify areas of agreement and areas that need reconciliation. The inter-comparison has been carried out by the cooperating

  20. A Comparison Among Plastic Deformation Capacities of RC Members According to International Codes

    International Nuclear Information System (INIS)

    Tripepi, C.; Failla, G.; Santini, A.; Nucera, F.

    2008-01-01

    The aim is to compare plastic deformation capacities of flexure-controlled reinforced concrete members, as predicted by the Italian Seismic Code, Eurocode 8 and FEMA356. For completeness, recent studies in the literature are also referred to. The comparison is pursued in context with a nonlinear static analysis run on 2D frame structures. This allows to assess whether and to which extent plastic deformation capacities may be affected by variations in those quantities, such as shear span and/or axial load, depending on which plastic deformation capacities are generally given

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  2. Comparison of Engine Cycle Codes for Rocket-Based Combined Cycle Engines

    Science.gov (United States)

    Waltrup, Paul J.; Auslender, Aaron H.; Bradford, John E.; Carreiro, Louis R.; Gettinger, Christopher; Komar, D. R.; McDonald, J.; Snyder, Christopher A.

    2002-01-01

    This paper summarizes the results from a one day workshop on Rocket-Based Combined Cycle (RBCC) Engine Cycle Codes held in Monterey CA in November of 2000 at the 2000 JANNAF JPM with the authors as primary participants. The objectives of the workshop were to discuss and compare the merits of existing Rocket-Based Combined Cycle (RBCC) engine cycle codes being used by government and industry to predict RBCC engine performance and interpret experimental results. These merits included physical and chemical modeling, accuracy and user friendliness. The ultimate purpose of the workshop was to identify the best codes for analyzing RBCC engines and to document any potential shortcomings, not to demonstrate the merits or deficiencies of any particular engine design. Five cases representative of the operating regimes of typical RBCC engines were used as the basis of these comparisons. These included Mach 0 sea level static and Mach 1.0 and Mach 2.5 Air-Augmented-Rocket (AAR), Mach 4 subsonic combustion ramjet or dual-mode scramjet, and Mach 8 scramjet operating modes. Specification of a generic RBCC engine geometry and concomitant component operating efficiencies, bypass ratios, fuel/oxidizer/air equivalence ratios and flight dynamic pressures were provided. The engine included an air inlet, isolator duct, axial rocket motor/injector, axial wall fuel injectors, diverging combustor, and exit nozzle. Gaseous hydrogen was used as the fuel with the rocket portion of the system using a gaseous H2/O2 propellant system to avoid cryogenic issues. The results of the workshop, even after post-workshop adjudication of differences, were surprising. They showed that the codes predicted essentially the same performance at the Mach 0 and I conditions, but progressively diverged from a common value (for example, for fuel specific impulse, Isp) as the flight Mach number increased, with the largest differences at Mach 8. The example cases and results are compared and discussed in this paper.

  3. Comparison of computer code calculations with experimental results obtained in the NSPP series of experiments

    International Nuclear Information System (INIS)

    Tobias, M.L.

    1987-01-01

    Experiments were done on several aerosols in air atmospheres at varying temperatures and humidity conditions of interest in forming a data base for testing aerosol behavior models used as part of the process of evaluating the ''source term'' in light water reactor accidents. This paper deals with the problems of predicting the observed experimental data for suspended aerosol concentration with aerosol calculational codes. Comparisons of measured versus predicted data are provided

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

    International Nuclear Information System (INIS)

    Neymotin, L.

    1994-04-01

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

  5. Computer code SICHTA-85/MOD 1 for thermohydraulic and mechanical modelling of WWER fuel channel behaviour during LOCA and comparison with original version of the SICHTA code

    International Nuclear Information System (INIS)

    Bujan, A.; Adamik, V.; Misak, J.

    1986-01-01

    A brief description is presented of the expansion of the SICHTA-83 computer code for the analysis of the thermal history of the fuel channel for large LOCAs by modelling the mechanical behaviour of fuel element cladding. The new version of the code has a more detailed treatment of heat transfer in the fuel-cladding gap because it also respects the mechanical (plastic) deformations of the cladding and the fuel-cladding interaction (magnitude of contact pressure). Also respected is the change in pressure of the gas filling of the fuel element, the mechanical criterion is considered of a failure of the cladding and the degree is considered of the blockage of the through-flow cross section for coolant flow in the fuel channel. The LOCA WWER-440 model computation provides a comparison of the new SICHTA-85/MOD 1 code with the results of the original 83 version of SICHTA. (author)

  6. Simulation of guided-wave ultrasound propagation in composite laminates: Benchmark comparisons of numerical codes and experiment.

    Science.gov (United States)

    Leckey, Cara A C; Wheeler, Kevin R; Hafiychuk, Vasyl N; Hafiychuk, Halyna; Timuçin, Doğan A

    2018-03-01

    Ultrasonic wave methods constitute the leading physical mechanism for nondestructive evaluation (NDE) and structural health monitoring (SHM) of solid composite materials, such as carbon fiber reinforced polymer (CFRP) laminates. Computational models of ultrasonic wave excitation, propagation, and scattering in CFRP composites can be extremely valuable in designing practicable NDE and SHM hardware, software, and methodologies that accomplish the desired accuracy, reliability, efficiency, and coverage. The development and application of ultrasonic simulation approaches for composite materials is an active area of research in the field of NDE. This paper presents comparisons of guided wave simulations for CFRP composites implemented using four different simulation codes: the commercial finite element modeling (FEM) packages ABAQUS, ANSYS, and COMSOL, and a custom code executing the Elastodynamic Finite Integration Technique (EFIT). Benchmark comparisons are made between the simulation tools and both experimental laser Doppler vibrometry data and theoretical dispersion curves. A pristine and a delamination type case (Teflon insert in the experimental specimen) is studied. A summary is given of the accuracy of simulation results and the respective computational performance of the four different simulation tools. Published by Elsevier B.V.

  7. A class of Sudan-decodable codes

    DEFF Research Database (Denmark)

    Nielsen, Rasmus Refslund

    2000-01-01

    In this article, Sudan's algorithm is modified into an efficient method to list-decode a class of codes which can be seen as a generalization of Reed-Solomon codes. The algorithm is specialized into a very efficient method for unique decoding. The code construction can be generalized based...... on algebraic-geometry codes and the decoding algorithms are generalized accordingly. Comparisons with Reed-Solomon and Hermitian codes are made....

  8. Update on comparison of the particle production using Mars simulation code

    CERN Document Server

    Prior, G; Kirk, H G; Souchlas, N; Ding, X

    2011-01-01

    In the International Design Study for the Neutrino Factory (IDS-NF), a 5-15 GeV (kinetic energy) proton beam impinges a Hg jet target, in order to produce pions that will decay into muons. The muons are captured and transformed into a beam, then passed to the downstream acceleration system. The target sits in a solenoid eld tapering from 20 T down to below 2 T, over several meters, permitting an optimized capture of the pions that will produce useful muons for the machine. The target and pion capture systems have been simulated using MARS. This paper presents an updated comparison of the particles production using the MARS code versions m1507 and m1510 on different machines located at the European Organization for Nuclear Research (CERN) and Brookhaven National Laboratory (BNL).

  9. NEACRP comparison of codes for the radiation protection assessment of transportation packages. Solutions to problems 1 - 4

    International Nuclear Information System (INIS)

    Avery, A.F.; Locke, H.F.

    1992-03-01

    In 1985 the Reactor Physics Committee of the Nuclear Energy Agency initiated an intercomparison of codes for the calculation of the performance of shielding for the transportation of spent reactor fuel. The results of the application of a range of codes to the prediction of the dose-rates in the four theoretical benchmarks set to examine the attenuation of radiation through a variety of cask geometries are presented in this report. The contributions from neutrons, fission product gamma-rays and secondary gamma-rays are tabulated separately, and grouped according to the type of method of calculation employed. A brief discussion is included for each set of results, and overall comparisons of the methods, codes, and nuclear data are made. A number of conclusions are drawn on the advantages and disadvantages of the various methods of calculation, based upon the results of their application to these four benchmark problems

  10. Validation of one-dimensional module of MARS 2.1 computer code by comparison with the RELAP5/MOD3.3 developmental assessment results

    International Nuclear Information System (INIS)

    Lee, Y. J.; Bae, S. W.; Chung, B. D.

    2003-02-01

    This report records the results of the code validation for the one-dimensional module of the MARS 2.1 thermal hydraulics analysis code by means of result-comparison with the RELAP5/MOD3.3 computer code. For the validation calculations, simulations of the RELAP5 code development assessment problem, which consists of 22 simulation problems in 3 categories, have been selected. The results of the 3 categories of simulations demonstrate that the one-dimensional module of the MARS 2.1 code and the RELAP5/MOD3.3 code are essentially the same code. This is expected as the two codes have basically the same set of field equations, constitutive equations and main thermal hydraulic models. The results suggests that the high level of code validity of the RELAP5/MOD3.3 can be directly applied to the MARS one-dimensional module

  11. NEACRP comparison of source term codes for the radiation protection assessment of transportation packages

    International Nuclear Information System (INIS)

    Broadhead, B.L.; Locke, H.F.; Avery, A.F.

    1994-01-01

    The results for Problems 5 and 6 of the NEACRP code comparison as submitted by six participating countries are presented in summary. These problems concentrate on the prediction of the neutron and gamma-ray sources arising in fuel after a specified irradiation, the fuel being uranium oxide for problem 5 and a mixture of uranium and plutonium oxides for problem 6. In both problems the predicted neutron sources are in good agreement for all participants. For gamma rays, however, there are differences, largely due to the omission of bremsstrahlung in some calculations

  12. The aeroelastic code FLEXLAST

    Energy Technology Data Exchange (ETDEWEB)

    Visser, B. [Stork Product Eng., Amsterdam (Netherlands)

    1996-09-01

    To support the discussion on aeroelastic codes, a description of the code FLEXLAST was given and experiences within benchmarks and measurement programmes were summarized. The code FLEXLAST has been developed since 1982 at Stork Product Engineering (SPE). Since 1992 FLEXLAST has been used by Dutch industries for wind turbine and rotor design. Based on the comparison with measurements, it can be concluded that the main shortcomings of wind turbine modelling lie in the field of aerodynamics, wind field and wake modelling. (au)

  13. Validation of thermalhydraulic codes

    International Nuclear Information System (INIS)

    Wilkie, D.

    1992-01-01

    Thermalhydraulic codes require to be validated against experimental data collected over a wide range of situations if they are to be relied upon. A good example is provided by the nuclear industry where codes are used for safety studies and for determining operating conditions. Errors in the codes could lead to financial penalties, to the incorrect estimation of the consequences of accidents and even to the accidents themselves. Comparison between prediction and experiment is often described qualitatively or in approximate terms, e.g. ''agreement is within 10%''. A quantitative method is preferable, especially when several competing codes are available. The codes can then be ranked in order of merit. Such a method is described. (Author)

  14. The Coding Process and Its Challenges

    Directory of Open Access Journals (Sweden)

    Judith A. Holton, Ph.D.

    2010-02-01

    Full Text Available Coding is the core process in classic grounded theory methodology. It is through coding that the conceptual abstraction of data and its reintegration as theory takes place. There are two types of coding in a classic grounded theory study: substantive coding, which includes both open and selective coding procedures, and theoretical coding. In substantive coding, the researcher works with the data directly, fracturing and analysing it, initially through open coding for the emergence of a core category and related concepts and then subsequently through theoretical sampling and selective coding of data to theoretically saturate the core and related concepts. Theoretical saturation is achieved through constant comparison of incidents (indicators in the data to elicit the properties and dimensions of each category (code. This constant comparing of incidents continues until the process yields the interchangeability of indicators, meaning that no new properties or dimensions are emerging from continued coding and comparison. At this point, the concepts have achieved theoretical saturation and the theorist shifts attention to exploring the emergent fit of potential theoretical codes that enable the conceptual integration of the core and related concepts to produce hypotheses that account for relationships between the concepts thereby explaining the latent pattern of social behaviour that forms the basis of the emergent theory. The coding of data in grounded theory occurs in conjunction with analysis through a process of conceptual memoing, capturing the theorist’s ideation of the emerging theory. Memoing occurs initially at the substantive coding level and proceeds to higher levels of conceptual abstraction as coding proceeds to theoretical saturation and the theorist begins to explore conceptual reintegration through theoretical coding.

  15. Code-experiment comparison on wall condensation tests in the presence of non-condensable gases-Numerical calculations for containment studies

    Energy Technology Data Exchange (ETDEWEB)

    Malet, J., E-mail: jeanne.malet@irsn.fr [Institut de Radioprotection et de Surete Nucleaire (IRSN), PSN-RES, SCA, BP 68, 91192 Gif-sur-Yvette (France); Porcheron, E.; Dumay, F.; Vendel, J. [Institut de Radioprotection et de Surete Nucleaire (IRSN), PSN-RES, SCA, BP 68, 91192 Gif-sur-Yvette (France)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer Steam condensation on walls has been investigated in the TOSQAN vessel. Black-Right-Pointing-Pointer Experiments on 7 different tests have been performed. Black-Right-Pointing-Pointer Different steam injections and wall temperatures are used. Black-Right-Pointing-Pointer Simulations are performed in 2D using the TONUS code. Black-Right-Pointing-Pointer Code-experiments comparisons at many different locations show a good agreement. - Abstract: During the course of a severe Pressurized Water Reactor accident, pressurization of the containment occurs and hydrogen can be produced by the reactor core oxidation and distributed in the containment according to convection flows and wall condensation. Filmwise wall condensation in the presence of non-condensable gases is a subject of many interests and extensive studies have been performed in the past. Some empirical correlations have demonstrated their limit for extrapolation under different thermal-hydraulic conditions and at different geometries/scales. The French Institute for Radiological Protection and Nuclear Safety (IRSN) has developed a numerical tool and an experimental facility in order to investigate free convection flows in the presence of condensation. The objective of this paper is to present numerical results obtained on different wall condensation tests in 7 m{sup 3} volume vessel (TOSQAN facility), and to compare them with the experimental ones. Over eight tests are considered here, and code-experiment comparison is performed on many different locations, giving an extensive insight of the code assessment for air-steam mixture flows involving wall condensation in the presence of non-condensable gases.

  16. Intracoin - International Nuclide Transport Code Intercomparison Study

    International Nuclear Information System (INIS)

    1984-09-01

    The purpose of the project is to obtain improved knowledge of the influence of various strategies for radionuclide transport modelling for the safety assessment of final repositories for nuclear waste. This is a report of the first phase of the project which was devoted to a comparison of the numerical accuracy of the computer codes used in the study. The codes can be divided into five groups, namely advection-dispersion models, models including matrix diffusion and chemical effects and finally combined models. The results are presented as comparisons of calculations since the objective of level 1 was code verification. (G.B.)

  17. An Efficient Method for Verifying Gyrokinetic Microstability Codes

    Science.gov (United States)

    Bravenec, R.; Candy, J.; Dorland, W.; Holland, C.

    2009-11-01

    Benchmarks for gyrokinetic microstability codes can be developed through successful ``apples-to-apples'' comparisons among them. Unlike previous efforts, we perform the comparisons for actual discharges, rendering the verification efforts relevant to existing experiments and future devices (ITER). The process requires i) assembling the experimental analyses at multiple times, radii, discharges, and devices, ii) creating the input files ensuring that the input parameters are faithfully translated code-to-code, iii) running the codes, and iv) comparing the results, all in an organized fashion. The purpose of this work is to automate this process as much as possible: At present, a python routine is used to generate and organize GYRO input files from TRANSP or ONETWO analyses. Another routine translates the GYRO input files into GS2 input files. (Translation software for other codes has not yet been written.) Other python codes submit the multiple GYRO and GS2 jobs, organize the results, and collect them into a table suitable for plotting. (These separate python routines could easily be consolidated.) An example of the process -- a linear comparison between GYRO and GS2 for a DIII-D discharge at multiple radii -- will be presented.

  18. Final Report for the "Fusion Application for Core-Edge Transport Simulations (FACETS)"

    Energy Technology Data Exchange (ETDEWEB)

    Cary, John R; Kruger, Scott

    2014-10-02

    The FACETS project over its lifetime developed the first self-consistent core-edge coupled capabilities, a new transport solver for modeling core transport in tokamak cores, developed a new code for modeling wall physics over long time scales, and significantly improved the capabilities and performance of legacy components, UEDGE, NUBEAM, GLF23, GYRO, and BOUT++. These improved capabilities leveraged the team’s expertise in applied mathematics (solvers and algorithms) and computer science (performance improvements and language interoperability). The project pioneered new methods for tackling the complexity of simulating the concomitant complexity of tokamak experiments.

  19. Techniques and results of tokamak-edge simulation

    International Nuclear Information System (INIS)

    Smith, G.R.; Brown, P.N.; Campbell, R.B.; Knoll, D.A.; McHugh, P.R.; Rensink, M.E.; Rognlien, T.D.

    1995-01-01

    This paper describes recent development of the UEDGE code in three important areas. (1) Non-orthogonal grids allow accurate treatment of experimental geometries in which divertor plates intersect flux surfaces at oblique angles. (2) Radiating impurities are included by means of one or more continuity equations that describe transport and sources and sinks due to ionization and recombination processes. (3) Advanced iterative methods that reduce storage and execution time allow us to find fully converged solutions of larger problems (i.e., finer grids). Sample calculations are presented to illustrate these developments. ((orig.))

  20. Techniques and results of tokamak-edge simulation

    International Nuclear Information System (INIS)

    Smith, G.R.; Brown, P.N.; Rensink, M.E.; Rognlien, T.D.; Campbell, R.B.; Knoll, D.A.; McHugh, P.R.

    1994-01-01

    This paper describes recent development of the UEDGE code in three important areas. (1) Non-orthogonal grids allow accurate treatment of experimental geometries in which divertor plates intersect flux surfaces at oblique angles. (2) Radating impurities are included by means of one or more continuity equations that describe transport and sources, and sinks due to ionization and recombination processes. (3) Advanced iterative methods that reduce storage and execution time allow us to find fully converged solutions of larger problems (i.e., finer grids). Sample calculations are presented to illustrate these development

  1. Optix: A Monte Carlo scintillation light transport code

    Energy Technology Data Exchange (ETDEWEB)

    Safari, M.J., E-mail: mjsafari@aut.ac.ir [Department of Energy Engineering and Physics, Amir Kabir University of Technology, PO Box 15875-4413, Tehran (Iran, Islamic Republic of); Afarideh, H. [Department of Energy Engineering and Physics, Amir Kabir University of Technology, PO Box 15875-4413, Tehran (Iran, Islamic Republic of); Ghal-Eh, N. [School of Physics, Damghan University, PO Box 36716-41167, Damghan (Iran, Islamic Republic of); Davani, F. Abbasi [Nuclear Engineering Department, Shahid Beheshti University, PO Box 1983963113, Tehran (Iran, Islamic Republic of)

    2014-02-11

    The paper reports on the capabilities of Monte Carlo scintillation light transport code Optix, which is an extended version of previously introduced code Optics. Optix provides the user a variety of both numerical and graphical outputs with a very simple and user-friendly input structure. A benchmarking strategy has been adopted based on the comparison with experimental results, semi-analytical solutions, and other Monte Carlo simulation codes to verify various aspects of the developed code. Besides, some extensive comparisons have been made against the tracking abilities of general-purpose MCNPX and FLUKA codes. The presented benchmark results for the Optix code exhibit promising agreements. -- Highlights: • Monte Carlo simulation of scintillation light transport in 3D geometry. • Evaluation of angular distribution of detected photons. • Benchmark studies to check the accuracy of Monte Carlo simulations.

  2. A GFR benchmark comparison of transient analysis codes based on the ETDR concept

    International Nuclear Information System (INIS)

    Bubelis, E.; Coddington, P.; Castelliti, D.; Dor, I.; Fouillet, C.; Geus, E. de; Marshall, T.D.; Van Rooijen, W.; Schikorr, M.; Stainsby, R.

    2007-01-01

    A GFR (Gas-cooled Fast Reactor) transient benchmark study was performed to investigate the ability of different code systems to calculate the transition in the core heat removal from the main circuit forced flow to natural circulation cooling using the Decay Heat Removal (DHR) system. This benchmark is based on a main blower failure in the Experimental Technology Demonstration Reactor (ETDR) with reactor scram. The codes taking part into the benchmark are: RELAP5, TRAC/AAA, CATHARE, SIM-ADS, MANTA and SPECTRA. For comparison purposes the benchmark was divided into several stages: the initial steady-state solution, the main blower flow run-down, the opening of the DHR loop and the transition to natural circulation and finally the 'quasi' steady heat removal from the core by the DHR system. The results submitted by the participants showed that all the codes gave consistent results for all four stages of the benchmark. In the steady-state the calculations revealed some differences in the clad and fuel temperatures, the core and main loop pressure drops and in the total Helium mass inventory. Also some disagreements were observed in the Helium and water flow rates in the DHR loop during the final natural circulation stage. Good agreement was observed for the total main blower flow rate and Helium temperature rise in the core, as well as for the Helium inlet temperature into the core. In order to understand the reason for the differences in the initial 'blind' calculations a second round of calculations was performed using a more precise set of boundary conditions

  3. Reliability of cause of death coding: an international comparison.

    Science.gov (United States)

    Antini, Carmen; Rajs, Danuta; Muñoz-Quezada, María Teresa; Mondaca, Boris Andrés Lucero; Heiss, Gerardo

    2015-07-01

    This study evaluates the agreement of nosologic coding of cardiovascular causes of death between a Chilean coder and one in the United States, in a stratified random sample of death certificates of persons aged ≥ 60, issued in 2008 in the Valparaíso and Metropolitan regions, Chile. All causes of death were converted to ICD-10 codes in parallel by both coders. Concordance was analyzed with inter-coder agreement and Cohen's kappa coefficient by level of specification ICD-10 code for the underlying cause and the total causes of death coding. Inter-coder agreement was 76.4% for all causes of death and 80.6% for the underlying cause (agreement at the four-digit level), with differences by the level of specification of the ICD-10 code, by line of the death certificate, and by number of causes of death per certificate. Cohen's kappa coefficient was 0.76 (95%CI: 0.68-0.84) for the underlying cause and 0.75 (95%CI: 0.74-0.77) for the total causes of death. In conclusion, causes of death coding and inter-coder agreement for cardiovascular diseases in two regions of Chile are comparable to an external benchmark and with reports from other countries.

  4. Upgrades to the WIMS-ANL code

    International Nuclear Information System (INIS)

    Woodruff, W. L.

    1998-01-01

    The dusty old source code in WIMS-D4M has been completely rewritten to conform more closely with current FORTRAN coding practices. The revised code contains many improvements in appearance, error checking and in control of the output. The output is now tabulated to fit the typical 80 column window or terminal screen. The Segev method for resonance integral interpolation is now an option. Most of the dimension limitations have been removed and replaced with variable dimensions within a compile-time fixed container. The library is no longer restricted to the 69 energy group structure, and two new libraries have been generated for use with the code. The new libraries are both based on ENDF/B-VI data with one having the original 69 energy group structure and the second with a 172 group structure. The common source code can be used with PCs using both Windows 95 and NT, with a Linux based operating system and with UNIX based workstations. Comparisons of this version of the code to earlier evaluations with ENDF/B-V are provided, as well as, comparisons with the new libraries

  5. Upgrades to the WIMS-ANL code

    International Nuclear Information System (INIS)

    Woodruff, W.L.; Leopando, L.S.

    1998-01-01

    The dusty old source code in WIMS-D4M has been completely rewritten to conform more closely with current FORTRAN coding practices. The revised code contains many improvements in appearance, error checking and in control of the output. The output is now tabulated to fit the typical 80 column window or terminal screen. The Segev method for resonance integral interpolation is now an option. Most of the dimension limitations have been removed and replaced with variable dimensions within a compile-time fixed container. The library is no longer restricted to the 69 energy group structure, and two new libraries have been generated for use with the code. The new libraries are both based on ENDF/B-VI data with one having the original 69 energy group structure and the second with a 172 group structure. The common source code can be used with PCs using both Windows 95 and NT, with a Linux based operating system and with UNIX based workstations. Comparisons of this version of the code to earlier evaluations with ENDF/B-V are provided, as well as, comparisons with the new libraries. (author)

  6. Genome-wide identification of coding and non-coding conserved sequence tags in human and mouse genomes

    Directory of Open Access Journals (Sweden)

    Maggi Giorgio P

    2008-06-01

    Full Text Available Abstract Background The accurate detection of genes and the identification of functional regions is still an open issue in the annotation of genomic sequences. This problem affects new genomes but also those of very well studied organisms such as human and mouse where, despite the great efforts, the inventory of genes and regulatory regions is far from complete. Comparative genomics is an effective approach to address this problem. Unfortunately it is limited by the computational requirements needed to perform genome-wide comparisons and by the problem of discriminating between conserved coding and non-coding sequences. This discrimination is often based (thus dependent on the availability of annotated proteins. Results In this paper we present the results of a comprehensive comparison of human and mouse genomes performed with a new high throughput grid-based system which allows the rapid detection of conserved sequences and accurate assessment of their coding potential. By detecting clusters of coding conserved sequences the system is also suitable to accurately identify potential gene loci. Following this analysis we created a collection of human-mouse conserved sequence tags and carefully compared our results to reliable annotations in order to benchmark the reliability of our classifications. Strikingly we were able to detect several potential gene loci supported by EST sequences but not corresponding to as yet annotated genes. Conclusion Here we present a new system which allows comprehensive comparison of genomes to detect conserved coding and non-coding sequences and the identification of potential gene loci. Our system does not require the availability of any annotated sequence thus is suitable for the analysis of new or poorly annotated genomes.

  7. International assessment of PCA codes

    International Nuclear Information System (INIS)

    Neymotin, L.; Lui, C.; Glynn, J.; Archarya, S.

    1993-11-01

    Over the past three years (1991-1993), an extensive international exercise for intercomparison of a group of six Probabilistic Consequence Assessment (PCA) codes was undertaken. The exercise was jointly sponsored by the Commission of European Communities (CEC) and OECD Nuclear Energy Agency. This exercise was a logical continuation of a similar effort undertaken by OECD/NEA/CSNI in 1979-1981. The PCA codes are currently used by different countries for predicting radiological health and economic consequences of severe accidents at nuclear power plants (and certain types of non-reactor nuclear facilities) resulting in releases of radioactive materials into the atmosphere. The codes participating in the exercise were: ARANO (Finland), CONDOR (UK), COSYMA (CEC), LENA (Sweden), MACCS (USA), and OSCAAR (Japan). In parallel with this inter-code comparison effort, two separate groups performed a similar set of calculations using two of the participating codes, MACCS and COSYMA. Results of the intercode and inter-MACCS comparisons are presented in this paper. The MACCS group included four participants: GREECE: Institute of Nuclear Technology and Radiation Protection, NCSR Demokritos; ITALY: ENEL, ENEA/DISP, and ENEA/NUC-RIN; SPAIN: Universidad Politecnica de Madrid (UPM) and Consejo de Seguridad Nuclear; USA: Brookhaven National Laboratory, US NRC and DOE

  8. Translation of ARAC computer codes

    International Nuclear Information System (INIS)

    Takahashi, Kunio; Chino, Masamichi; Honma, Toshimitsu; Ishikawa, Hirohiko; Kai, Michiaki; Imai, Kazuhiko; Asai, Kiyoshi

    1982-05-01

    In 1981 we have translated the famous MATHEW, ADPIC and their auxiliary computer codes for CDC 7600 computer version to FACOM M-200's. The codes consist of a part of the Atmospheric Release Advisory Capability (ARAC) system of Lawrence Livermore National Laboratory (LLNL). The MATHEW is a code for three-dimensional wind field analysis. Using observed data, it calculates the mass-consistent wind field of grid cells by a variational method. The ADPIC is a code for three-dimensional concentration prediction of gases and particulates released to the atmosphere. It calculates concentrations in grid cells by the particle-in-cell method. They are written in LLLTRAN, i.e., LLNL Fortran language and are implemented on the CDC 7600 computers of LLNL. In this report, i) the computational methods of the MATHEW/ADPIC and their auxiliary codes, ii) comparisons of the calculated results with our JAERI particle-in-cell, and gaussian plume models, iii) translation procedures from the CDC version to FACOM M-200's, are described. Under the permission of LLNL G-Division, this report is published to keep the track of the translation procedures and to serve our JAERI researchers for comparisons and references of their works. (author)

  9. Validation of One-Dimensional Module of MARS-KS1.2 Computer Code By Comparison with the RELAP5/MOD3.3/patch3 Developmental Assessment Results

    International Nuclear Information System (INIS)

    Bae, S. W.; Chung, B. D.

    2010-07-01

    This report records the results of the code validation for the one-dimensional module of the MARS-KS thermal hydraulics analysis code by means of result-comparison with the RELAP5/MOD3.3 computer code. For the validation calculations, simulations of the RELAP5 Code Developmental Assessment Problem, which consists of 22 simulation problems in 3 categories, have been selected. The results of the 3 categories of simulations demonstrate that the one-dimensional module of the MARS code and the RELAP5/MOD3.3 code are essentially the same code. This is expected as the two codes have basically the same set of field equations, constitutive equations and main thermal hydraulic models. The result suggests that the high level of code validity of the RELAP5/MOD3.3 can be directly applied to the MARS one-dimensional module

  10. Fast comparison of IS radar code sequences for lag profile inversion

    Directory of Open Access Journals (Sweden)

    M. S. Lehtinen

    2008-08-01

    Full Text Available A fast method for theoretically comparing the posteriori variances produced by different phase code sequences in incoherent scatter radar (ISR experiments is introduced. Alternating codes of types 1 and 2 are known to be optimal for selected range resolutions, but the code sets are inconveniently long for many purposes like ground clutter estimation and in cases where coherent echoes from lower ionospheric layers are to be analyzed in addition to standard F-layer spectra.

    The method is used in practice for searching binary code quads that have estimation accuracy almost equal to that of much longer alternating code sets. Though the code sequences can consist of as few as four different transmission envelopes, the lag profile estimation variances are near to the theoretical minimum. Thus the short code sequence is equally good as a full cycle of alternating codes with the same pulse length and bit length. The short code groups cannot be directly decoded, but the decoding is done in connection with more computationally expensive lag profile inversion in data analysis.

    The actual code searches as well as the analysis and real data results from the found short code searches are explained in other papers sent to the same issue of this journal. We also discuss interesting subtle differences found between the different alternating codes by this method. We assume that thermal noise dominates the incoherent scatter signal.

  11. 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

  12. 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

  13. OPAL reactor calculations using the Monte Carlo code serpent

    Energy Technology Data Exchange (ETDEWEB)

    Ferraro, Diego; Villarino, Eduardo [Nuclear Engineering Dept., INVAP S.E., Rio Negro (Argentina)

    2012-03-15

    In the present work the Monte Carlo cell code developed by VTT Serpent v1.1.14 is used to model the MTR fuel assemblies (FA) and control rods (CR) from OPAL (Open Pool Australian Light-water) reactor in order to obtain few-group constants with burnup dependence to be used in the already developed reactor core models. These core calculations are performed using CITVAP 3-D diffusion code, which is well-known reactor code based on CITATION. Subsequently the results are compared with those obtained by the deterministic calculation line used by INVAP, which uses the Collision Probability Condor cell-code to obtain few-group constants. Finally the results are compared with the experimental data obtained from the reactor information for several operation cycles. As a result several evaluations are performed, including a code to code cell comparison at cell and core level and calculation-experiment comparison at core level in order to evaluate the Serpent code actual capabilities. (author)

  14. Gap Conductance model Validation in the TASS/SMR-S code using MARS code

    International Nuclear Information System (INIS)

    Ahn, Sang Jun; Yang, Soo Hyung; Chung, Young Jong; Lee, Won Jae

    2010-01-01

    Korea Atomic Energy Research Institute (KAERI) has been developing the TASS/SMR-S (Transient and Setpoint Simulation/Small and Medium Reactor) code, which is a thermal hydraulic code for the safety analysis of the advanced integral reactor. An appropriate work to validate the applicability of the thermal hydraulic models within the code should be demanded. Among the models, the gap conductance model which is describes the thermal gap conductivity between fuel and cladding was validated through the comparison with MARS code. The validation of the gap conductance model was performed by evaluating the variation of the gap temperature and gap width as the changed with the power fraction. In this paper, a brief description of the gap conductance model in the TASS/SMR-S code is presented. In addition, calculated results to validate the gap conductance model are demonstrated by comparing with the results of the MARS code with the test case

  15. A Comparison of Nuclear Power Plant Simulator with RELAP5/MOD3 code about Steam Generator Tube Rupture

    International Nuclear Information System (INIS)

    Kim, Sung Hyun; Moon, Chan Ki; Park, Sung Baek; Na, Man Gyun

    2013-01-01

    The RELAP5/MOD3 code introduced in cooperation with U. S. NRC has been utilized mainly for validation calculation of accident analysis submitted by licensee in Korea. The Korea Institute of Nuclear Safety has built a verification system of LWR accident analysis with RELAP5/MOD3 code engine. Therefore, the simulator replicates the design basis accident and its results are compared with RELAP5/MOD3 code results that will have important implications in the verification of the simulator in the future. The SGTR simulations were performed by the simulator and its results were compared with ones by RELAP5/MOD3 code in this study. Thus, the results of this study can be used as materials to build the verification system of the nuclear power plant simulator. We tried to compare with RELAP5/MOD3 verification code by replicating major parameters of steam generator tube rupture using the simulator for OPR-1000 in Yonggwang training center. By comparing the changes in temperature, pressure and inventory of the reactor coolant system and main steam system during the SGTR, it was confirmed that the main behaviors of SGTR which the simulator and RELAP5/MOD3 code showed are similar. However, the behavior of SG pressure and level that are important parameters to diagnose the accident were a little different. We estimated that RELAP5/MOD3 code was not reflected the major control systems in detail, such as FWCS, SBCS and PPCS. The different behaviors of SG level and pressure in this study should be needed an additional review. As a result of the comparison, the major simulation parameters behavior by RELAP5/MOD3 code agreed well with the one by the simulator. Therefore, it is thought that RELAP5/MOD3 code is used as a tool for validation of NPP simulator in the near future through this study

  16. 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.

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

    Science.gov (United States)

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

    2017-07-01

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

  18. Validating the BISON fuel performance code to integral LWR experiments

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, R.L., E-mail: Richard.Williamson@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Gamble, K.A., E-mail: Kyle.Gamble@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Perez, D.M., E-mail: Danielle.Perez@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Novascone, S.R., E-mail: Stephen.Novascone@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Pastore, G., E-mail: Giovanni.Pastore@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Gardner, R.J., E-mail: Russell.Gardner@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Hales, J.D., E-mail: Jason.Hales@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Liu, W., E-mail: Wenfeng.Liu@anatech.com [ANATECH Corporation, 5435 Oberlin Dr., San Diego, CA 92121 (United States); Mai, A., E-mail: Anh.Mai@anatech.com [ANATECH Corporation, 5435 Oberlin Dr., San Diego, CA 92121 (United States)

    2016-05-15

    Highlights: • The BISON multidimensional fuel performance code is being validated to integral LWR experiments. • Code and solution verification are necessary prerequisites to validation. • Fuel centerline temperature comparisons through all phases of fuel life are very reasonable. • Accuracy in predicting fission gas release is consistent with state-of-the-art modeling and the involved uncertainties. • Rod diameter comparisons are not satisfactory and further investigation is underway. - Abstract: BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON's computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to date for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Results demonstrate that (1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, (2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and (3) comparison

  19. An experimental comparison of coded modulation strategies for 100 Gb/s transceivers

    NARCIS (Netherlands)

    Sillekens, E.; Alvarado, A.; Okonkwo, C.; Thomsen, B.C.

    2016-01-01

    Coded modulation is a key technique to increase the spectral efficiency of coherent optical communication systems. Two popular strategies for coded modulation are turbo trellis-coded modulation (TTCM) and bit-interleaved coded modulation (BICM) based on low-density parity-check (LDPC) codes.

  20. MCNP and GADRAS Comparisons

    International Nuclear Information System (INIS)

    Klasky, Marc Louis; Myers, Steven Charles; James, Michael R.; Mayo, Douglas R.

    2016-01-01

    To facilitate the timely execution of System Threat Reviews (STRs) for DNDO, and also to develop a methodology for performing STRs, LANL performed comparisons of several radiation transport codes (MCNP, GADRAS, and Gamma-Designer) that have been previously utilized to compute radiation signatures. While each of these codes has strengths, it is of paramount interest to determine the limitations of each of the respective codes and also to identify the most time efficient means by which to produce computational results, given the large number of parametric cases that are anticipated in performing STR's. These comparisons serve to identify regions of applicability for each code and provide estimates of uncertainty that may be anticipated. Furthermore, while performing these comparisons, examination of the sensitivity of the results to modeling assumptions was also examined. These investigations serve to enable the creation of the LANL methodology for performing STRs. Given the wide variety of radiation test sources, scenarios, and detectors, LANL calculated comparisons of the following parameters: decay data, multiplicity, device (n,γ) leakages, and radiation transport through representative scenes and shielding. This investigation was performed to understand potential limitations utilizing specific codes for different aspects of the STR challenges.

  1. Comparisons with measured data of the simulated local core parameters by the coupled code ATHLET-BIPR-VVER applying a new enhanced model of the reactor pressure vessel

    International Nuclear Information System (INIS)

    Nikonov, S.; Pasichnyk, I.; Velkov, K.; Pautz, A.

    2011-01-01

    The paper describes the performed comparisons of measured and simulated local core data based on the OECD/NEA Benchmark on Kalinin-3 NPP: 'Switching off of one of the four operating main circulation pumps at nominal reactor power'. The local measurements of in core self-powered neutron detectors (SPND) in 64 fuel assemblies on 7 axial levels are used for the comparisons of the assemblies axial power distributions and the thermocouples readings at 93 fuel assembly heads are applied for the fuel assembly coolant temperature comparisons. The analyses are done on the base of benchmark transient calculations performed with the coupled system code ATHLET/BIPR-VVER. In order to describe more realistically the fluid mixing phenomena in a reactor pressure vessel a new enhanced nodalization scheme is being developed. It could take into account asymmetric flow behaviour in the reactor pressure vessel structures like downcomer, reactor core inlet and outlet, control rods' guided tubes, support grids etc. For this purpose details of the core geometry are modelled. About 58000 control volumes and junctions are applied. Cross connection are used to describe the interaction between the fluid objects. The performed comparisons are of great interest because they show some advantages by performing coupled code production pseudo-3D analysis of NPPs applying the parallel thermo-hydraulic channel methodology (or 1D thermo-hydraulic system code modeling). (Authors)

  2. Code Package to Analyze Parameters of the WWER Fuel Rod. TOPRA-2 Code - Verification Data

    International Nuclear Information System (INIS)

    Scheglov, A.; Proselkov, V.; Passage, G.; Stefanova, S.

    2009-01-01

    Presented are the data for computer codes to analyze WWER fuel rods, used in the WWER department of RRC 'Kurchatov Institute'. Presented is the description of TOPRA-2 code intended for the engineering analysis of thermophysical and strength parameters of the WWER fuel rod - temperature distributions along the fuel radius, gas pressures under the cladding, stresses in the cladding, etc. for the reactor operation in normal conditions. Presented are some results of the code verification against test problems and the data obtained in the experimental programs. Presented are comparison results of the calculations with TOPRA-2 and TRANSURANUS (V1M1J06) codes. Results obtained in the course of verification demonstrate possibility of application of the methodology and TOPRA-2 code for the engineering analysis of the WWER fuel rods

  3. Linear and nonlinear verification of gyrokinetic microstability codes

    Science.gov (United States)

    Bravenec, R. V.; Candy, J.; Barnes, M.; Holland, C.

    2011-12-01

    Verification of nonlinear microstability codes is a necessary step before comparisons or predictions of turbulent transport in toroidal devices can be justified. By verification we mean demonstrating that a code correctly solves the mathematical model upon which it is based. Some degree of verification can be accomplished indirectly from analytical instability threshold conditions, nonlinear saturation estimates, etc., for relatively simple plasmas. However, verification for experimentally relevant plasma conditions and physics is beyond the realm of analytical treatment and must rely on code-to-code comparisons, i.e., benchmarking. The premise is that the codes are verified for a given problem or set of parameters if they all agree within a specified tolerance. True verification requires comparisons for a number of plasma conditions, e.g., different devices, discharges, times, and radii. Running the codes and keeping track of linear and nonlinear inputs and results for all conditions could be prohibitive unless there was some degree of automation. We have written software to do just this and have formulated a metric for assessing agreement of nonlinear simulations. We present comparisons, both linear and nonlinear, between the gyrokinetic codes GYRO [J. Candy and R. E. Waltz, J. Comput. Phys. 186, 545 (2003)] and GS2 [W. Dorland, F. Jenko, M. Kotschenreuther, and B. N. Rogers, Phys. Rev. Lett. 85, 5579 (2000)]. We do so at the mid-radius for the same discharge as in earlier work [C. Holland, A. E. White, G. R. McKee, M. W. Shafer, J. Candy, R. E. Waltz, L. Schmitz, and G. R. Tynan, Phys. Plasmas 16, 052301 (2009)]. The comparisons include electromagnetic fluctuations, passing and trapped electrons, plasma shaping, one kinetic impurity, and finite Debye-length effects. Results neglecting and including electron collisions (Lorentz model) are presented. We find that the linear frequencies with or without collisions agree well between codes, as do the time averages of

  4. Development of 2-d cfd code

    International Nuclear Information System (INIS)

    Mirza, S.A.

    1999-01-01

    In the present study, a two-dimensional computer code has been developed in FORTRAN using CFD technique, which is basically a numerical scheme. This computer code solves the Navier Stokes equations and continuity equation to find out the velocity and pressure fields within a given domain. This analysis has been done for the developed within a square cavity driven by the upper wall which has become a bench mark for testing and comparing the newly developed numerical schemes. Before to handle this task, different one-dimensional cases have been studied by CFD technique and their FORTRAN programs written. The cases studied are Couette flow, Poiseuille flow with and without using symmetric boundary condition. Finally a comparison between CFD results and analytical results has also been made. For the cavity flow the results from the developed code have been obtained for different Reynolds numbers which are finally presented in the form of velocity vectors. The comparison of the developed code results have been made with the results obtained from the share ware version of a commercially available code for Reynolds number of 10.0. The disagreement in the results quantitatively and qualitatively at some grid points of the calculation domain have been discussed and future recommendations in this regard have also been made. (author)

  5. Comparison of the two thermal-hydraulic codes Thyc and Vipre-02 on Vatican experiment

    International Nuclear Information System (INIS)

    Montat, D.; Maurel, F.; Olive, J.; Srikantiah, G.

    1993-08-01

    EDF's THYC and EPRI's VIPRE-02 3D thermalhydraulics computer codes are based on strongly different approaches (mixing against two-fluid representation for the two-phase flow, and porous media against subchannel approach for the rod bundle geometry description). In order to assess their efficiencies, they were both used to compute the two-phase flow behavior in an EDF experimental set-up, VATICAN. This set-up consisted of a 2.10 m high vertically oriented rod bundle of 40 heated rods in a 10 by 4 matrix (9.5 mm diameter and 12.6 mm square pitch). Refrigerant - 114 was introduced through the bottom left and right sides of the bundle and exited at the top. The lower 1.6 m height of the bundle was separated into two symmetric halves by a vertical wall. Desequilibrium in flow and quality between the two halves could be set, so that strong lateral mixing occurred above the divider wall. Comparison of codes computing and experimental data showed to some extent the superiority of a 6-equations model, but also highlighted the dramatic need for good constitutive relations concerning for instance the turbulent mixing. Some systematic deviations from experimental data were detected, linked to poor accuracy of some chosen closure laws. (authors). 6 figs., 3 tabs., 5 refs

  6. Comparison of strongly heat-driven flow codes for unsaturated media

    International Nuclear Information System (INIS)

    Updegraff, C.D.

    1989-08-01

    Under the sponsorship of the US Nuclear Regulatory Commission, Sandia National Laboratories (SNL) is developing a performance assessment methodology for the analysis of long-term disposal of high-level radioactive waste (HLW) in unsaturated welded tuff. As part of this effort, SNL evaluated existing strongly heat-driven flow computer codes for simulating ground-water flow in unsaturated media. The three codes tested, NORIA, PETROS, and TOUGH, were compared against a suite of problems for which analytical and numerical solutions or experimental results exist. The problems were selected to test the abilities of the codes to simulate situations ranging from simple, uncoupled processes, such as two-phase flow or heat transfer, to fully coupled processes, such as vaporization caused by high temperatures. In general, all three codes were found to be difficult to use because of (1) built-in time stepping criteria, (2) the treatment of boundary conditions, and (3) handling of evaporation/condensation problems. A drawback of the study was that adequate problems related to expected repository conditions were not available in the literature. Nevertheless, the results of this study suggest the need for thorough investigations of the impact of heat on the flow field in the vicinity of an unsaturated HLW repository. Recommendations are to develop a new flow code combining the best features of these three codes and eliminating the worst ones. 19 refs., 49 figs

  7. Neutronic evolution of SENA reactor during the first and second cycles. Comparison between the experimental power distributions obtained from the in-core instrumentation evaluation code CIRCE and the theoretical power values computed with the two-dimensional diffusion-evolution code EVOE

    International Nuclear Information System (INIS)

    Andrieux, Chantal

    1976-03-01

    The neutronic evolution of the reacteur Sena during the first and second cycles is presented. The experimental power distributions, obtained from the in-core instrumentation evaluation code CIRCE are compared with the theoretical powers calculated with the two-dimensional diffusion-evolution code EVOE. The CIRCE code allows: the study of the evolution of the principal parameters of the core, the comparison of the results of measured and theoretical estimates. Therefore this study has a great interest for the knowledge of the neutronic evolution of the core, as well as the validation of the refinement of theoretical estimation methods. The core calculation methods and requisite data for the evaluation of the measurements are presented after a brief description of the SENA core and its inner instrumentation. The principle of the in-core instrumentation evaluation code CIRCE, and calculation of the experimental power distributions and nuclear core parameters are then exposed. The results of the evaluation are discussed, with a comparison of the theoretical and experimental results. Taking account of the approximations used, these results, as far as the first and second cycles at SENA are concerned, are satisfactory, the deviations between theoretical and experimental power distributions being lower than 3% at the middle of the reactor and 9% at the periphery [fr

  8. SUMMARY OF GENERAL WORKING GROUP A+B+D: CODES BENCHMARKING.

    Energy Technology Data Exchange (ETDEWEB)

    WEI, J.; SHAPOSHNIKOVA, E.; ZIMMERMANN, F.; HOFMANN, I.

    2006-05-29

    Computer simulation is an indispensable tool in assisting the design, construction, and operation of accelerators. In particular, computer simulation complements analytical theories and experimental observations in understanding beam dynamics in accelerators. The ultimate function of computer simulation is to study mechanisms that limit the performance of frontier accelerators. There are four goals for the benchmarking of computer simulation codes, namely debugging, validation, comparison and verification: (1) Debugging--codes should calculate what they are supposed to calculate; (2) Validation--results generated by the codes should agree with established analytical results for specific cases; (3) Comparison--results from two sets of codes should agree with each other if the models used are the same; and (4) Verification--results from the codes should agree with experimental measurements. This is the summary of the joint session among working groups A, B, and D of the HI32006 Workshop on computer codes benchmarking.

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

    International Nuclear Information System (INIS)

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

    2003-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-11-01

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

  11. 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

  12. Input data required for specific performance assessment codes

    International Nuclear Information System (INIS)

    Seitz, R.R.; Garcia, R.S.; Starmer, R.J.; Dicke, C.A.; Leonard, P.R.; Maheras, S.J.; Rood, A.S.; Smith, R.W.

    1992-02-01

    The Department of Energy's National Low-Level Waste Management Program at the Idaho National Engineering Laboratory generated this report on input data requirements for computer codes to assist States and compacts in their performance assessments. This report gives generators, developers, operators, and users some guidelines on what input data is required to satisfy 22 common performance assessment codes. Each of the codes is summarized and a matrix table is provided to allow comparison of the various input required by the codes. This report does not determine or recommend which codes are preferable

  13. The 1992 ENDF Pre-processing codes

    International Nuclear Information System (INIS)

    Cullen, D.E.

    1992-01-01

    This document summarizes the 1992 version of the ENDF pre-processing codes which are required for processing evaluated nuclear data coded in the format ENDF-4, ENDF-5, or ENDF-6. Included are the codes CONVERT, MERGER, LINEAR, RECENT, SIGMA1, LEGEND, FIXUP, GROUPIE, DICTION, MIXER, VIRGIN, COMPLOT, EVALPLOT, RELABEL. Some of the functions of these codes are: to calculate cross-sections from resonance parameters; to calculate angular distributions, group average, mixtures of cross-sections, etc; to produce graphical plottings and data comparisons. The codes are designed to operate on virtually any type of computer including PC's. They are available from the IAEA Nuclear Data Section, free of charge upon request, on magnetic tape or a set of HD diskettes. (author)

  14. Verification of 3-D generation code package for neutronic calculations of WWERs

    International Nuclear Information System (INIS)

    Sidorenko, V.D.; Aleshin, S.S.; Bolobov, P.A.; Bolshagin, S.N.; Lazarenko, A.P.; Markov, A.V.; Morozov, V.V.; Syslov, A.A.; Tsvetkov, V.M.

    2000-01-01

    Materials on verification of the 3 -d generation code package for WWERs neutronic calculations are presented. The package includes: - spectral code TVS-M; - 2-D fine mesh diffusion code PERMAK-A for 4- or 6-group calculation of WWER core burnup; - 3-D coarse mesh diffusion code BIPR-7A for 2-group calculations of quasi-stationary WWERs regimes. The materials include both TVS-M verification data and verification data on PERMAK-A and BIPR-7A codes using constant libraries generated with TVS-M. All materials are related to the fuel without Gd. TVS-M verification materials include results of comparison both with benchmark calculations obtained by other codes and with experiments carried out at ZR-6 critical facility. PERMAK-A verification materials contain results of comparison with TVS-M calculations and with ZR-6 experiments. BIPR-7A materials include comparison with operation data for Dukovany-2 and Loviisa-1 NPPs (WWER-440) and for Balakovo NPP Unit 4 (WWER-1000). The verification materials demonstrate rather good accuracy of calculations obtained with the use of code package of the 3 -d generation. (Authors)

  15. Timing comparison of two-dimensional discrete-ordinates codes for criticality calculations

    International Nuclear Information System (INIS)

    Miller, W.F. Jr.; Alcouffe, R.E.; Bosler, G.E.; Brinkley, F.W. Jr.; O'dell, R.D.

    1979-01-01

    The authors compare two-dimensional discrete-ordinates neutron transport computer codes to solve reactor criticality problems. The fundamental interest is in determining which code requires the minimum Central Processing Unit (CPU) time for a given numerical model of a reasonably realistic fast reactor core and peripherals. The computer codes considered are the most advanced available and, in three cases, are not officially released. The conclusion, based on the study of four fast reactor core models, is that for this class of problems the diffusion synthetic accelerated version of TWOTRAN, labeled TWOTRAN-DA, is superior to the other codes in terms of CPU requirements

  16. Spread-spectrum communication using binary spatiotemporal chaotic codes

    International Nuclear Information System (INIS)

    Wang Xingang; Zhan Meng; Gong Xiaofeng; Lai, C.H.; Lai, Y.-C.

    2005-01-01

    We propose a scheme to generate binary code for baseband spread-spectrum communication by using a chain of coupled chaotic maps. We compare the performances of this type of spatiotemporal chaotic code with those of a conventional code used frequently in digital communication, the Gold code, and demonstrate that our code is comparable or even superior to the Gold code in several key aspects: security, bit error rate, code generation speed, and the number of possible code sequences. As the field of communicating with chaos faces doubts in terms of performance comparison with conventional digital communication schemes, our work gives a clear message that communicating with chaos can be advantageous and it deserves further attention from the nonlinear science community

  17. CONSUL code package application for LMFR core calculations

    Energy Technology Data Exchange (ETDEWEB)

    Chibinyaev, A.V.; Teplov, P.S.; Frolova, M.V. [RNC ' Kurchatovskiy institute' , Kurchatov sq.1, Moscow (Russian Federation)

    2008-07-01

    CONSUL code package designed for the calculation of reactor core characteristics has been developed at the beginning of 90's. The calculation of nuclear reactor core characteristics is carried out on the basis of correlated neutron, isotope and temperature distributions. The code package has been generally used for LWR core characteristics calculations. At present CONSUL code package was adapted to calculate liquid metal fast reactors (LMFR). The comparisons with IAEA computational test 'Evaluation of benchmark calculations on a fast power reactor core with near zero sodium void effect' and BN-1800 testing calculations are presented in the paper. The IAEA benchmark core is based on the innovative core concept with sodium plenum above the core BN-800. BN-1800 core is the next development step which is foreseen for the Russian fast reactor concept. The comparison of the operational parameters has shown good agreement and confirms the possibility of CONSUL code package application for LMFR core calculation. (authors)

  18. Thermal-hydraulic codes validation for safety analysis of NPPs with RBMK

    International Nuclear Information System (INIS)

    Brus, N.A.; Ioussoupov, O.E.

    2000-01-01

    This work is devoted to validation of western thermal-hydraulic codes (RELAP5/MOD3 .2 and ATHLET 1.1 Cycle C) in application to Russian designed light water reactors. Such validation is needed due to features of RBMK reactor design and thermal-hydraulics in comparison with PWR and BWR reactors, for which these codes were developed and validated. These validation studies are concluded with a comparison of calculation results of modeling with the thermal-hydraulics codes with the experiments performed earlier using the thermal-hydraulics test facilities with the experimental data. (authors)

  19. Comparison of three gamma ray isotopic determination codes: FRAM, MGA, and TRIFID

    International Nuclear Information System (INIS)

    Cremers, T.L.; Malcom, J.E.; Bonner, C.A.

    1994-01-01

    The determination of the isotopic distribution of plutonium and the americium concentration is required for the assay of nuclear material by calorimetry or neutron coincidence counting. The isotopic information is used in calorimetric assay to compute the effective specific power from the measured isotopic fractions and the known specific power of each isotope. The effective specific power is combined with the heat measurement to obtain the mass of plutonium in the assayed nuclear material. The response of neutron coincidence counters is determined by the 240 Pu isotopic fraction with contributions from the other even plutonium isotopes. The effect of the 240 Pu isotopic fraction and the other neutron contributing isotopes are combined as 240 Pu effective. This is used to calculate the mass of nuclear material from the neutron counting data in a manner analogous to the effective specific power in calorimeter. Comparisons of the precision and accuracy of calorimetric assay and neutron coincidence counting often focus only on the precision and accuracy of the heat measurement (calorimetry) compared to the precision and accuracy of the neutron coincidence counting statistics. The major source of uncertainty for both calorimetric assay and neutron coincidence counting often lies in the determination of the plutonium isotopic distribution ad determined by gamma ray spectroscopy. Thus, the selection of the appropriate isotopic distribution code is of paramount importance to good calorimetric assay and neutron coincidence counting. Three gamma ray isotopic distribution codes, FRAM, MGA, and TRIFID have been compared at the Los Alamos Plutonium Facility under carefully controlled conditions of similar count rates, count times, and 240 Pu isotopic fraction

  20. Short-Block Protograph-Based LDPC Codes

    Science.gov (United States)

    Divsalar, Dariush; Dolinar, Samuel; Jones, Christopher

    2010-01-01

    Short-block low-density parity-check (LDPC) codes of a special type are intended to be especially well suited for potential applications that include transmission of command and control data, cellular telephony, data communications in wireless local area networks, and satellite data communications. [In general, LDPC codes belong to a class of error-correcting codes suitable for use in a variety of wireless data-communication systems that include noisy channels.] The codes of the present special type exhibit low error floors, low bit and frame error rates, and low latency (in comparison with related prior codes). These codes also achieve low maximum rate of undetected errors over all signal-to-noise ratios, without requiring the use of cyclic redundancy checks, which would significantly increase the overhead for short blocks. These codes have protograph representations; this is advantageous in that, for reasons that exceed the scope of this article, the applicability of protograph representations makes it possible to design highspeed iterative decoders that utilize belief- propagation algorithms.

  1. Design evaluation on sodium piping system and comparison of the design codes

    International Nuclear Information System (INIS)

    Lee, Dong Won; Jeong, Ji Young; Lee, Yong Bum; Lee, Hyeong Yeon

    2015-01-01

    A large-scale sodium test loop of STELLA-1 (Sodium integral effect test loop for safety simulation and assessment) with two main piping systems has been installed at KAERI. In this study, design evaluations on the main sodium piping systems in STELLA-1 have been conducted according to the DBR (design by rule) codes of the ASME B31.1 and RCC-MRx RB-3600. In addition, design evaluations according to the DBA (design by analysis) code of the ASME Section III Subsection NB-3200 have been conducted. The evaluation results for the present piping systems showed that results from the DBR codes were more conservative than those from the DBA code, and among the DBR codes, the non-nuclear code of the ASME B31.1 was more conservative than the French nuclear DBR code of the RCC-MRx RB-3600. The conservatism on the DBR codes of the ASME B31.1 and RCC-MRx RB-3600 was quantified based on the present sodium piping analyses.

  2. Design evaluation on sodium piping system and comparison of the design codes

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Jeong, Ji Young; Lee, Yong Bum; Lee, Hyeong Yeon [KAERI, Daejeon (Korea, Republic of)

    2015-03-15

    A large-scale sodium test loop of STELLA-1 (Sodium integral effect test loop for safety simulation and assessment) with two main piping systems has been installed at KAERI. In this study, design evaluations on the main sodium piping systems in STELLA-1 have been conducted according to the DBR (design by rule) codes of the ASME B31.1 and RCC-MRx RB-3600. In addition, design evaluations according to the DBA (design by analysis) code of the ASME Section III Subsection NB-3200 have been conducted. The evaluation results for the present piping systems showed that results from the DBR codes were more conservative than those from the DBA code, and among the DBR codes, the non-nuclear code of the ASME B31.1 was more conservative than the French nuclear DBR code of the RCC-MRx RB-3600. The conservatism on the DBR codes of the ASME B31.1 and RCC-MRx RB-3600 was quantified based on the present sodium piping analyses.

  3. 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.

  4. Comparison of european computer codes relative to the aerosol behavior in PWR containment buildings during severe core damage accidents. (Modelling of steam condensation on the particles)

    International Nuclear Information System (INIS)

    Bunz, H.; Dunbar, L.H.; Fermandjian, J.; Lhiaubet, G.

    1987-11-01

    An aerosol code comparison exercise was performed within the framework of the Commission of European Communities (Division of Safety of Nuclear Installations). This exercise, focused on the process of steam condensation onto the aerosols occurring in PWR containment buildings during severe core damage accidents, has allowed to understand the discrepancies between the results obtained. These discrepancies are due, in particular, to whether the curvature effect is modelled or not in the codes

  5. Comparison of two-dimensional and three-dimensional MHD equilibrium and stability codes

    International Nuclear Information System (INIS)

    Herrnegger, F.; Merkel, P.; Johnson, J.L.

    1986-02-01

    Stability results obtained with the fully three-dimensional magnetohydrodynamic code BETA, the helically invariant code HERA, and the asymptotic stellarator expansion code STEP agree well for a straight l = 2, M = 5 stellarator model. This good agreement between the BETA and STEP codes persists as toroidal curvature is introduced. This validation provides justification for confidence in work with these models. 20 refs., 11 figs

  6. Coding training for medical students: How good is diagnoses coding with ICD-10 by novices?

    Directory of Open Access Journals (Sweden)

    Stausberg, Jürgen

    2005-04-01

    Full Text Available Teaching of knowledge and competence in documentation and coding is an essential part of medical education. Therefore, coding training had been placed within the course of epidemiology, medical biometry, and medical informatics. From this, we can draw conclusions about the quality of coding by novices. One hundred and eighteen students coded diagnoses from 15 nephrological cases in homework. In addition to interrater reliability, validity was calculated by comparison with a reference coding. On the level of terminal codes, 59.3% of the students' results were correct. The completeness was calculated as 58.0%. The results on the chapter level increased up to 91.5% and 87.7% respectively. For the calculation of reliability a new, simple measure was developed that leads to values of 0.46 on the level of terminal codes and 0.87 on the chapter level for interrater reliability. The figures of concordance with the reference coding are quite similar. In contrary, routine data show considerably lower results with 0.34 and 0.63 respectively. Interrater reliability and validity of coding by novices is as good as coding by experts. The missing advantage of experts could be explained by the workload of documentation and a negative attitude to coding on the one hand. On the other hand, coding in a DRG-system is handicapped by a large number of detailed coding rules, which do not end in uniform results but rather lead to wrong and random codes. Anyway, students left the course well prepared for coding.

  7. The intercomparison of aerosol codes

    International Nuclear Information System (INIS)

    Dunbar, I.H.; Fermandjian, J.; Gauvain, J.

    1988-01-01

    The behavior of aerosols in a reactor containment vessel following a severe accident could be an important determinant of the accident source term to the environment. Various processes result in the deposition of the aerosol onto surfaces within the containment, from where they are much less likely to be released. Some of these processes are very sensitive to particle size, so it is important to model the aerosol growth processes: agglomeration and condensation. A number of computer codes have been written to model growth and deposition processes. They have been tested against each other in a series of code comparison exercises. These exercises have investigated sensitivities to physical and numerical assumptions and have also proved a useful means of quality control for the codes. Various exercises in which code predictions are compared with experimental results are now under way

  8. A Comparison of Athletic Movement Among Talent-Identified Juniors From Different Football Codes in Australia: Implications for Talent Development.

    Science.gov (United States)

    Woods, Carl T; Keller, Brad S; McKeown, Ian; Robertson, Sam

    2016-09-01

    Woods, CT, Keller, BS, McKeown, I, and Robertson, S. A comparison of athletic movement among talent-identified juniors from different football codes in Australia: implications for talent development. J Strength Cond Res 30(9): 2440-2445, 2016-This study aimed to compare the athletic movement skill of talent-identified (TID) junior Australian Rules football (ARF) and soccer players. The athletic movement skill of 17 TID junior ARF players (17.5-18.3 years) was compared against 17 TID junior soccer players (17.9-18.7 years). Players in both groups were members of an elite junior talent development program within their respective football codes. All players performed an athletic movement assessment that included an overhead squat, double lunge, single-leg Romanian deadlift (both movements performed on right and left legs), a push-up, and a chin-up. Each movement was scored across 3 essential assessment criteria using a 3-point scale. The total score for each movement (maximum of 9) and the overall total score (maximum of 63) were used as the criterion variables for analysis. A multivariate analysis of variance tested the main effect of football code (2 levels) on the criterion variables, whereas a 1-way analysis of variance identified where differences occurred. A significant effect was noted, with the TID junior ARF players outscoring their soccer counterparts when performing the overhead squat and push-up. No other criterions significantly differed according to the main effect. Practitioners should be aware that specific sporting requirements may incur slight differences in athletic movement skill among TID juniors from different football codes. However, given the low athletic movement skill noted in both football codes, developmental coaches should address the underlying movement skill capabilities of juniors when prescribing physical training in both codes.

  9. COCOA Code for Creating Mock Observations of Star Cluster Models

    OpenAIRE

    Askar, Abbas; Giersz, Mirek; Pych, Wojciech; Dalessandro, Emanuele

    2017-01-01

    We introduce and present results from the COCOA (Cluster simulatiOn Comparison with ObservAtions) code that has been developed to create idealized mock photometric observations using results from numerical simulations of star cluster evolution. COCOA is able to present the output of realistic numerical simulations of star clusters carried out using Monte Carlo or \\textit{N}-body codes in a way that is useful for direct comparison with photometric observations. In this paper, we describe the C...

  10. Implications of wall recycling and carbon source locations on core plasma fueling and impurity content in DIII-D

    International Nuclear Information System (INIS)

    Groth, M.; Porter, G.D.; Fenstermacher, M.E.; Lasnier, C.J.; Meyer, W.M.; Rensink, M.E.; Wolf, N.S.; Boedo, J.A.; Moyer, R.A.; Rudakov, D.L.; Brooks, N.H.; Groebner, R.J.; Petrie, T.W.; Owen, L.W.; Wang, G.; Zeng, L.; Watkins, J.G.

    2005-01-01

    Measurement and modeling of the 2-D poloidal D α intensity distribution in DIII-D low and medium density L-mode and ELMy H-mode plasmas indicate that hydrogen neutrals predominantly fuel the core from the divertor X-point region. The 2-D distribution of neutral deuterium and low-charge-state carbon were measured in the divertor and the high-field side midplane scrape-off layer (SOL) using tangentially viewing cameras. The emission in the high-field SOL at the equatorial plane was found to be three to four orders of magnitude lower than at the strike points in the divertor, suggesting a strong divertor particle source. Modeling using the UEDGE/DEGAS codes predicted the poloidal fueling distribution to be dependent on the direction of the ion Bx∇B drift. In plasmas with the Bx∇B drift into the divertor stronger fueling from the inner divertor than from the outer is predicted, due to a lower-temperature and higher-density plasma in the inner leg. UEDGE simulations with carbon produced by both physical and chemical sputtering at the divertor plates and walls only are in agreement with a large set of diagnostic data. The simulations indicate flow reversal in the inner divertor that augments the leakage of carbon ions from the divertor into the core. (author)

  11. Comparison of codes and standards for radiographic inspection

    International Nuclear Information System (INIS)

    Bingoeldag, M. M.; Aksu, M.; Akguen, A. F.

    1995-01-01

    This report compares the procedurel requirements and acceptance criteria for radiographic inspections specified in the relevant national and international codes and standards. In particular, detailed analysis of inspection conditions such as exposure arrangements, and contrast requirements are given

  12. Comparison of the THYC and FLICA-3M codes by the pseudo-cubic thin-plate method; Comparaison par la methode des plaques de predicteurs de flux critique obtenus a l`aide des codes THYC et FLICA-3

    Energy Technology Data Exchange (ETDEWEB)

    Banner, D; Crecy, F de

    1993-06-01

    The pseudo cubic Spline method (PCSM) is a statistical tool developed by the CEA. It is designed to analyse experimental points and in particular thermalhydraulic data. Predictors of the occurrence of critical heat flux are obtained by using Spline functions. In this paper, predictors have been computed from the same CHF databases by using two different flow analyses to derive local thermal-hydraulic variables at the CHF location. In fact, CEA`s FLICA-3M represents rod bundles by interconnected subchannels whereas EDF`s THYC code uses a porous 3D approach. In a first step, the PCSM is briefly presented as well as the two codes studied here. Then, the comparison methodology is explained in order to prove that advanced analysis of thermalhydraulic codes can be achieved with the PCSM. (authors). 6 figs., 2 tabs., 5 refs.

  13. Application of the three-dimensional Oak Ridge transport code

    International Nuclear Information System (INIS)

    Rhoades, W.A.; Childs, R.L.; Emmett, M.B.; Cramer, S.N.

    1984-01-01

    TORT, a 3-d extension of the DOT discrete ordinates transport code is now in production use for studies of radiation penetration into large concrete and masonry structures. This paper discusses certain features of the new code and shows representative results, including comparisons with Monte Carlo calculations

  14. Comparison of spectra for validation of Penelope code for the energy range used in mammography; Comparacao de espectros para validacao do codigo PENELOPE para faixa de energia usada em mamografia

    Energy Technology Data Exchange (ETDEWEB)

    Albuquerque, M.A.G.; Ferreira, N.M.P.D., E-mail: malbuqueque@hotmail.co [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil); Pires, E.; Ganizeu, M.D.; Almeida, C.E. de, E-mail: marianogd@uol.com.b [Universidade do Estado do Rio de Janeiro (UERJ), RJ (Brazil); Prizio, R.; Peixoto, J.G., E-mail: guilherm@ird.gov.b [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2009-07-01

    The spectra simulated by the Penelope code were compared with the spectra experimentally obtained through the silicon PIN photodiode detector, and with spectra calculated by the code of IPEN, and the comparison exhibited a concordance of 93.3 %, and make them an option for study of X-ray spectroscopy in the voltage range used in mammography

  15. Attention Filtering in the Design of Electronic Map Displays: A Comparison of Color-Coding, Intensity Coding, and Decluttering Techniques

    National Research Council Canada - National Science Library

    Yeh, Michelle; Wickens, Christopher D

    2000-01-01

    In a series of experiments, the use of color-coding, intensity coding, and decluttering were compared order to assess their potential benefits for accessing information from electronic map displays...

  16. Predictive Bias and Sensitivity in NRC Fuel Performance Codes

    Energy Technology Data Exchange (ETDEWEB)

    Geelhood, Kenneth J.; Luscher, Walter G.; Senor, David J.; Cunningham, Mitchel E.; Lanning, Donald D.; Adkins, Harold E.

    2009-10-01

    The latest versions of the fuel performance codes, FRAPCON-3 and FRAPTRAN were examined to determine if the codes are intrinsically conservative. Each individual model and type of code prediction was examined and compared to the data that was used to develop the model. In addition, a brief literature search was performed to determine if more recent data have become available since the original model development for model comparison.

  17. Comparison of SAND-II and FERRET

    International Nuclear Information System (INIS)

    Wootan, D.W.; Schmittroth, F.

    1981-01-01

    A comparison was made of the advantages and disadvantages of two codes, SAND-II and FERRET, for determining the neutron flux spectrum and uncertainty from experimental dosimeter measurements as anticipated in the FFTF Reactor Characterization Program. This comparison involved an examination of the methodology and the operational performance of each code. The merits of each code were identified with respect to theoretical basis, directness of method, solution uniqueness, subjective influences, and sensitivity to various input parameters

  18. Introduction of gadolinium in the library of Leopard code

    International Nuclear Information System (INIS)

    Claro, L.H.; Menezes, A.

    1989-12-01

    The materials Gd-154, Gd-155, Gd-156 and Gd-157 were included in the LEOPARD code library at the request of FURNAS Centrais Eletricas S.A. Results from comparison of LEOPARD and WIMSD/4 codes for a typical cell with 7 burnup steps, are presented. (author) [pt

  19. Code-To-Code Benchmarking Of The Porflow And GoldSim Contaminant Transport Models Using A Simple 1-D Domain - 11191

    International Nuclear Information System (INIS)

    Hiergesell, R.; Taylor, G.

    2010-01-01

    An investigation was conducted to compare and evaluate contaminant transport results of two model codes, GoldSim and Porflow, using a simple 1-D string of elements in each code. Model domains were constructed to be identical with respect to cell numbers and dimensions, matrix material, flow boundary and saturation conditions. One of the codes, GoldSim, does not simulate advective movement of water; therefore the water flux term was specified as a boundary condition. In the other code, Porflow, a steady-state flow field was computed and contaminant transport was simulated within that flow-field. The comparisons were made solely in terms of the ability of each code to perform contaminant transport. The purpose of the investigation was to establish a basis for, and to validate follow-on work that was conducted in which a 1-D GoldSim model developed by abstracting information from Porflow 2-D and 3-D unsaturated and saturated zone models and then benchmarked to produce equivalent contaminant transport results. A handful of contaminants were selected for the code-to-code comparison simulations, including a non-sorbing tracer and several long- and short-lived radionuclides exhibiting both non-sorbing to strongly-sorbing characteristics with respect to the matrix material, including several requiring the simulation of in-growth of daughter radionuclides. The same diffusion and partitioning coefficients associated with each contaminant and the half-lives associated with each radionuclide were incorporated into each model. A string of 10-elements, having identical spatial dimensions and properties, were constructed within each code. GoldSim's basic contaminant transport elements, Mixing cells, were utilized in this construction. Sand was established as the matrix material and was assigned identical properties (e.g. bulk density, porosity, saturated hydraulic conductivity) in both codes. Boundary conditions applied included an influx of water at the rate of 40 cm/yr at one

  20. Dual Coding and Bilingual Memory.

    Science.gov (United States)

    Paivio, Allan; Lambert, Wallace

    1981-01-01

    Describes study which tested a dual coding approach to bilingual memory using tasks that permit comparison of the effects of bilingual encoding with verbal-nonverbal dual encoding items. Results provide strong support for a version of the independent or separate stories view of bilingual memory. (Author/BK)

  1. RAID-6Plus: A Comprised Methodology for Extending RAID-6 Codes

    Directory of Open Access Journals (Sweden)

    Ming-Zhu Deng

    2017-01-01

    Full Text Available Existing RAID-6 code extensions assume that failures are independent and instantaneous, overlooking the underlying mechanism of multifailure occurrences. Also, the effect of reconstruction window is ignored. Additionally, these coding extensions have not been adapted to occurrence patterns of failure in real-world applications. As a result, the third parity drive is set to handle the triple-failure scenario; however, the lower level failure situations have been left unattended. Therefore, a new methodology of extending RAID-6 codes named RAID-6Plus with better compromise has been studied in this paper. RAID-6Plus (Deng et al., 2015 employs short combinations which can greatly reuse overlapped elements during reconstruction to remake the third parity drive. A sample extension code called RDP+ is given based on RDP. Moreover, we extended the study to present another extension example called X-code+ which has better update penalty and load balance. The analysis shows that RAID-6Plus is a balanced tradeoff of reliability, performance, and practicality. For instance, RDP+ could achieve speedups as high as 33.4% in comparison to the RTP with conventional rebuild, 11.9% in comparison to RTP with the optimal rebuild, 47.7% in comparison to STAR with conventional rebuild, and 26.2% for a single failure rebuild.

  2. A NEM diffusion code for fuel management and time average core calculation

    International Nuclear Information System (INIS)

    Mishra, Surendra; Ray, Sherly; Kumar, A.N.

    2005-01-01

    A computer code based on Nodal expansion method has been developed for solving two groups three dimensional diffusion equation. This code can be used for fuel management and time average core calculation. Explicit Xenon and fuel temperature estimation are also incorporated in this code. TAPP-4 phase-B physics experimental results were analyzed using this code and a code based on FD method. This paper gives the comparison of the observed data and the results obtained with this code and FD code. (author)

  3. AMZ, library of multigroup constants for EXPANDA computer codes, generated by NJOY computer code from ENDF/B-IV

    International Nuclear Information System (INIS)

    Chalhoub, E.S.; Moraes, M. de.

    1984-01-01

    A 70-group, 37-isotope library of multigroup constants for fast reactor nuclear design calculations is described. Nuclear cross sections, transfer matrices, and self-shielding factors were generated with NJOY code and an auxiliary program RGENDF using evaluated data from ENDF/B-IV. The output is being issued in a format suitable for EXPANDA code. Comparisons with JFS-2 library, as well as, test resuls for 14 CSEWG benchmark critical assemblies are presented. (Author) [pt

  4. Comparison between the Findings from the TROI Experiments and the Sensitivity Studies by Using the TEXAS-V Code

    International Nuclear Information System (INIS)

    Park, I. K.; Kim, J. H.; Hong, S. W.; Min, B. T.; Hong, S. H.; Song, J. H.; Kim, H. D.

    2006-01-01

    Since a steam explosion may breach the integrity of a reactor vessel and containment, it is one of the most important severe accident issues. So, a lot of experimental and analytical researches on steam explosions have been performed. Although many findings from the steam explosion researches have been obtained, there still exist unsolved issues such as the explosivity of the real core material(corium) and the conversion ratio from the thermal energy to the mechanical energy. TROI experiments were carried out to provide the experimental data for these issues. The TROI experiments were performed with a prototypic material such as ZrO 2 melt and a mixture of ZrO 2 and UO 2 melt (corium). Several steam explosion codes including TEXAS-V had been developed by considering the findings in the past steam explosion experiments. However, some unique findings on steam explosions have been obtained from a series of TROI experiments. These findings should be considered in the application to a reactor safety analysis by using a computational code. In this paper, several findings from TROI experiments are discussed and the sensitivity studies on the TROI experimental parameters were conducted by using TEXAS-V code and TROI-13 test. The comparison between the TROI experimental findings and the results of the sensitivity study might allow us to know which parameter is important and which model is uncertain for steam explosions

  5. Comparison of two numerical modelling codes for hydraulic and transport calculations in the near-field

    International Nuclear Information System (INIS)

    Kalin, J.; Petkovsek, B.; Montarnal, Ph.; Genty, A.; Deville, E.; Krivic, J.; Ratej, J.

    2011-01-01

    In the past years the Slovenian Performance Analysis/Safety Assessment team has performed many generic studies for the future Slovenian low and intermediate level waste repository, most recently a Special Safety Analysis for the Krsko site. The modelling approach was to split the problem into three parts: near-field (detailed model of the repository), far-field (i.e., geosphere) and biosphere. In the Special Safety Analysis the code used to perform the near-field calculations was Hydrus2D. Recently the team has begun a cooperation with the French Commisariat al'Energie Atomique/Saclay (CEA/Saclay) and, as a part of this cooperation, began investigations into using the Alliances numerical platform for near-field calculations in order to compare the overall approach and calculated results. The article presents the comparison between these two codes for a silo-type repository that was considered in the Special Safety Analysis. The physical layout and characteristics of the repository are presented and a hydraulic and transport model of the repository is developed and implemented in Alliances. Some analysis of sensitivity to mesh fineness and to simulation timestep has been preformed and is also presented. The compared quantity is the output flux of radionuclides on the boundary of the model. Finally the results from Hydrus2D and Alliances are compared and the differences and similarities are commented.

  6. Comparison of two numerical modelling codes for hydraulic and transport calculations in the near-field

    Energy Technology Data Exchange (ETDEWEB)

    Kalin, J., E-mail: jan.kalin@zag.s [Slovenian National Building and Civil Engineering Institute, Dimiceva 12, SI-1000 Ljubljana (Slovenia); Petkovsek, B., E-mail: borut.petkovsek@zag.s [Slovenian National Building and Civil Engineering Institute, Dimiceva 12, SI-1000 Ljubljana (Slovenia); Montarnal, Ph., E-mail: philippe.montarnal@cea.f [CEA/Saclay, DM2S/SFME/LSET, Gif-sur-Yvette, 91191 cedex (France); Genty, A., E-mail: alain.genty@cea.f [CEA/Saclay, DM2S/SFME/LSET, Gif-sur-Yvette, 91191 cedex (France); Deville, E., E-mail: estelle.deville@cea.f [CEA/Saclay, DM2S/SFME/LSET, Gif-sur-Yvette, 91191 cedex (France); Krivic, J., E-mail: jure.krivic@geo-zs.s [Geological Survey of Slovenia, Dimiceva 14, SI-1000 Ljubljana (Slovenia); Ratej, J., E-mail: joze.ratej@geo-zs.s [Geological Survey of Slovenia, Dimiceva 14, SI-1000 Ljubljana (Slovenia)

    2011-04-15

    In the past years the Slovenian Performance Analysis/Safety Assessment team has performed many generic studies for the future Slovenian low and intermediate level waste repository, most recently a Special Safety Analysis for the Krsko site. The modelling approach was to split the problem into three parts: near-field (detailed model of the repository), far-field (i.e., geosphere) and biosphere. In the Special Safety Analysis the code used to perform the near-field calculations was Hydrus2D. Recently the team has begun a cooperation with the French Commisariat al'Energie Atomique/Saclay (CEA/Saclay) and, as a part of this cooperation, began investigations into using the Alliances numerical platform for near-field calculations in order to compare the overall approach and calculated results. The article presents the comparison between these two codes for a silo-type repository that was considered in the Special Safety Analysis. The physical layout and characteristics of the repository are presented and a hydraulic and transport model of the repository is developed and implemented in Alliances. Some analysis of sensitivity to mesh fineness and to simulation timestep has been preformed and is also presented. The compared quantity is the output flux of radionuclides on the boundary of the model. Finally the results from Hydrus2D and Alliances are compared and the differences and similarities are commented.

  7. Towards Effective Intra-flow Network Coding in Software Defined Wireless Mesh Networks

    Directory of Open Access Journals (Sweden)

    Donghai Zhu

    2016-01-01

    Full Text Available Wireless Mesh Networks (WMNs have potential to provide convenient broadband wireless Internet access to mobile users.With the support of Software-Defined Networking (SDN paradigm that separates control plane and data plane, WMNs can be easily deployed and managed. In addition, by exploiting the broadcast nature of the wireless medium and the spatial diversity of multi-hop wireless networks, intra-flow network coding has shown a greater benefit in comparison with traditional routing paradigms in data transmission for WMNs. In this paper, we develop a novel OpenCoding protocol, which combines the SDN technique with intra-flow network coding for WMNs. Our developed protocol can simplify the deployment and management of the network and improve network performance. In OpenCoding, a controller that works on the control plane makes routing decisions for mesh routers and the hop-by-hop forwarding function is replaced by network coding functions in data plane. We analyze the overhead of OpenCoding. Through a simulation study, we show the effectiveness of the OpenCoding protocol in comparison with existing schemes. Our data shows that OpenCoding outperforms both traditional routing and intra-flow network coding schemes.

  8. The new lattice code Paragon and its qualification for PWR core applications

    International Nuclear Information System (INIS)

    Ouisloumen, M.; Huria, H.C.; Mayhue, L.T.; Smith, R.M.; Kichty, M.J.; Matsumoto, H.; Tahara, Y.

    2003-01-01

    Paragon is a new two-dimensional transport code based on collision probability with interface current method and written entirely in Fortran 90/95. The qualification of Paragon has been completed and the results are very good. This qualification included a number of critical experiments. Comparisons to the Monte Carlo code MCNP for a wide variety of PWR assembly lattice types were also performed. In addition, Paragon-based core simulator models have been compared against PWR plant startup and operational data for a large number of plants. Some results of these calculations and also comparisons against models developed with a licensed Westinghouse lattice code, Phoenix-P, are presented. The qualification described in this paper provided the basis for the qualification of Paragon both as a validated transport code and as the nuclear data source for core simulator codes

  9. Calculation of neutron spectra produced in neutron generator target: Code testing.

    Science.gov (United States)

    Gaganov, V V

    2018-03-01

    DT-neutron spectra calculated using the SRIANG code was benchmarked against the results obtained by widely used Monte Carlo codes: PROFIL, SHORIN, TARGET, ENEA-JSI, MCUNED, DDT and NEUSDESC. The comparison of the spectra obtained by different codes confirmed the correctness of SRIANG calculations. The cross-checking of the compared spectra revealed some systematic features and possible errors of analysed codes. Copyright © 2017 Elsevier Ltd. All rights reserved.

  10. CFD code comparison for 2D airfoil flows

    DEFF Research Database (Denmark)

    Sørensen, Niels N.; Méndez, B.; Muñoz, A.

    2016-01-01

    The current paper presents the effort, in the EU AVATAR project, to establish the necessary requirements to obtain consistent lift over drag ratios among seven CFD codes. The flow around a 2D airfoil case is studied, for both transitional and fully turbulent conditions at Reynolds numbers of 3...

  11. The numerical comparison of fire combustion model and water-mist suppression with experiments by FDS code

    International Nuclear Information System (INIS)

    Li Hsuennien; Ferng Yuhming; Shih Chunkuan; Hsu Wensheng

    2007-01-01

    FDS [1] code numerically solves a form of the Navier-Stokes equations appropriate for low-speed, thermally driven flow with an emphasis on smoke and heat transport from fires. FDS uses a mixture fraction combustion model. The mixture fraction is a conserved scalar quantity that is defined as the fraction of fuel gas at a given point in the flow field. The model assumes that combustion is mixing-controlled, and that the reaction of fuel and oxygen is infinitely fast. In FDS, Lagrangian particles are used to simulate smoke movement and sprinkling water-mist discharge. In order to evaluate the combustion model and water-mist suppression function of the code, FDS analyses are conducted to simulate two enclosure fire cases available in the literature. Comparisons with other combustion models are also made. For fires suppression by water-mist in FDS, parametric studies are performed to compare various water-mist injection characteristics for maximum suppression. Numerical results indicate that the flame suppression is closely related to characteristics of the water mist, such as droplet diameter, mist injection velocity, injection density. Our present investigations show that the combustion model and water-mist suppression in FDS can provide simulation results that are comparable with the experiments. (author)

  12. Comparison of the results of several heat transfer computer codes when applied to a hypothetical nuclear waste repository

    International Nuclear Information System (INIS)

    Claiborne, H.C.; Wagner, R.S.; Just, R.A.

    1979-12-01

    A direct comparison of transient thermal calculations was made with the heat transfer codes HEATING5, THAC-SIP-3D, ADINAT, SINDA, TRUMP, and TRANCO for a hypothetical nuclear waste repository. With the exception of TRUMP and SINDA (actually closer to the earlier CINDA3G version), the other codes agreed to within +-5% for the temperature rises as a function of time. The TRUMP results agreed within +-5% up to about 50 years, where the maximum temperature occurs, and then began an oscillary behavior with up to 25% deviations at longer times. This could have resulted from time steps that were too large or from some unknown system problems. The available version of the SINDA code was not compatible with the IBM compiler without using an alternative method for handling a variable thermal conductivity. The results were about 40% low, but a reasonable agreement was obtained by assuming a uniform thermal conductivity; however, a programming error was later discovered in the alternative method. Some work is required on the IBM version to make it compatible with the system and still use the recommended method of handling variable thermal conductivity. TRANCO can only be run as a 2-D model, and TRUMP and CINDA apparently required longer running times and did not agree in the 2-D case; therefore, only HEATING5, THAC-SIP-3D, and ADINAT were used for the 3-D model calculations. The codes agreed within +-5%; at distances of about 1 ft from the waste canister edge, temperature rises were also close to that predicted by the 3-D model

  13. Axisym finite element code: modifications for pellet-cladding mechanical interaction analysis

    International Nuclear Information System (INIS)

    Engelman, G.P.

    1978-10-01

    Local strain concentrations in nuclear fuel rods are known to be potential sites for failure initiation. Assessment of such strain concentrations requires a two-dimensional analysis of stress and strain in both the fuel and the cladding during pellet-cladding mechanical interaction. To provide such a capability in the FRAP (Fuel Rod Analysis Program) codes, the AXISYM code (a small finite element program developed at the Idaho National Engineering Laboratory) was modified to perform a detailed fuel rod deformation analysis. This report describes the modifications which were made to the AXISYM code to adapt it for fuel rod analysis and presents comparisons made between the two-dimensional AXISYM code and the FRACAS-II code. FRACAS-II is the one-dimensional (generalized plane strain) fuel rod mechanical deformation subcode used in the FRAP codes. Predictions of these two codes should be comparable away from the fuel pellet free ends if the state of deformation at the pellet midplane is near that of generalized plane strain. The excellent agreement obtained in these comparisons checks both the correctness of the AXISYM code modifications as well as the validity of the assumption of generalized plane strain upon which the FRACAS-II subcode is based

  14. Cavitation Modeling in Euler and Navier-Stokes Codes

    Science.gov (United States)

    Deshpande, Manish; Feng, Jinzhang; Merkle, Charles L.

    1993-01-01

    Many previous researchers have modeled sheet cavitation by means of a constant pressure solution in the cavity region coupled with a velocity potential formulation for the outer flow. The present paper discusses the issues involved in extending these cavitation models to Euler or Navier-Stokes codes. The approach taken is to start from a velocity potential model to ensure our results are compatible with those of previous researchers and available experimental data, and then to implement this model in both Euler and Navier-Stokes codes. The model is then augmented in the Navier-Stokes code by the inclusion of the energy equation which allows the effect of subcooling in the vicinity of the cavity interface to be modeled to take into account the experimentally observed reduction in cavity pressures that occurs in cryogenic fluids such as liquid hydrogen. Although our goal is to assess the practicality of implementing these cavitation models in existing three-dimensional, turbomachinery codes, the emphasis in the present paper will center on two-dimensional computations, most specifically isolated airfoils and cascades. Comparisons between velocity potential, Euler and Navier-Stokes implementations indicate they all produce consistent predictions. Comparisons with experimental results also indicate that the predictions are qualitatively correct and give a reasonable first estimate of sheet cavitation effects in both cryogenic and non-cryogenic fluids. The impact on CPU time and the code modifications required suggests that these models are appropriate for incorporation in current generation turbomachinery codes.

  15. 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.

  16. Fatigue and rupture codified rules comparison

    International Nuclear Information System (INIS)

    Faidy, C.

    2004-01-01

    The European Directive on Pressure Equipment requests risk studies and in particular to assure no risk of fatigue and rupture in operation. The answers to these questions are different in the different existing design codes (EN Standards, ASME III and VIII or RCC-M or CODAP-CODETI codes) and corresponding in operation codes (ASME or RSE-M). Design safety factors, material properties, fabrication, refinement in the analysis methods, monitoring in operation, hydro-proof test level... Around these Codes, different rules are under development. A16 in France, R6 in UK or FITNET at the EC level. This paper is concerned by a comparison between 2 different Codes to analyze the risk of fatigue or rupture of pressure equipments and mainly a comparison between RCC-M Code and EN 13445 standard for pressure vessel. Recommendations for future work will be proposed. (authors)

  17. Comparison of computer codes (CE-THERM, FRAP-T5, GT3-FLECHT, and TRUMP-FLECHT) with data from the NRU-LOCA thermal hydraulic tests

    International Nuclear Information System (INIS)

    Mohr, C.L.; Rausch, W.N.; Hesson, G.M.

    1981-07-01

    The LOCA Simulation Program in the NRU reactor is the first set of experiments to provide data on the behavior of full-length, nuclear-heated PWR fuel bundles during the heatup, reflood, and quench phases of a loss-of-coolant accident (LOCA). This paper compares the temperature time histories of 4 experimental test cases with 4 computer codes: CE-THERM, FRAP-T5, GT3-FLECHT, and TRUMP-FLECHT. The preliminary comparisons between prediction and experiment show that the state-of-the art fuel codes have large uncertainties and are not necessarily conservative in predicting peak temperatures, turn around times, and bundle quench times

  18. Towards a heavy-ion transport capability in the MARS15 Code

    International Nuclear Information System (INIS)

    Mokhov, N.V.; Gudima, K.K.; Mashnik, S.G.; Rakhno, I.L.; Striganov, S.

    2004-01-01

    In order to meet the challenges of new accelerator and space projects and further improve modelling of radiation effects in microscopic objects, heavy-ion interaction and transport physics have been recently incorporated into the MARS15 Monte Carlo code. A brief description of new modules is given in comparison with experimental data. The MARS Monte Carlo code is widely used in numerous accelerator, detector, shielding and cosmic ray applications. The needs of the Relativistic Heavy-Ion Collider, Large Hadron Collider, Rare Isotope Accelerator and NASA projects have recently induced adding heavy-ion interaction and transport physics to the MARS15 code. The key modules of the new implementation are described below along with their comparisons to experimental data.

  19. Model comparisons of the reactive burn model SURF in three ASC codes

    Energy Technology Data Exchange (ETDEWEB)

    Whitley, Von Howard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stalsberg, Krista Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reichelt, Benjamin Lee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Shipley, Sarah Jayne [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2018-01-12

    A study of the SURF reactive burn model was performed in FLAG, PAGOSA and XRAGE. In this study, three different shock-to-detonation transition experiments were modeled in each code. All three codes produced similar model results for all the experiments modeled and at all resolutions. Buildup-to-detonation time, particle velocities and resolution dependence of the models was notably similar between the codes. Given the current PBX 9502 equations of state and SURF calibrations, each code is equally capable of predicting the correct detonation time and distance when impacted by a 1D impactor at pressures ranging from 10-16 GPa, as long as the resolution of the mesh is not too coarse.

  20. The InterFrost benchmark of Thermo-Hydraulic codes for cold regions hydrology - first inter-comparison results

    Science.gov (United States)

    Grenier, Christophe; Roux, Nicolas; Anbergen, Hauke; Collier, Nathaniel; Costard, Francois; Ferrry, Michel; Frampton, Andrew; Frederick, Jennifer; Holmen, Johan; Jost, Anne; Kokh, Samuel; Kurylyk, Barret; McKenzie, Jeffrey; Molson, John; Orgogozo, Laurent; Rivière, Agnès; Rühaak, Wolfram; Selroos, Jan-Olof; Therrien, René; Vidstrand, Patrik

    2015-04-01

    The impacts of climate change in boreal regions has received considerable attention recently due to the warming trends that have been experienced in recent decades and are expected to intensify in the future. Large portions of these regions, corresponding to permafrost areas, are covered by water bodies (lakes, rivers) that interact with the surrounding permafrost. For example, the thermal state of the surrounding soil influences the energy and water budget of the surface water bodies. Also, these water bodies generate taliks (unfrozen zones below) that disturb the thermal regimes of permafrost and may play a key role in the context of climate change. Recent field studies and modeling exercises indicate that a fully coupled 2D or 3D Thermo-Hydraulic (TH) approach is required to understand and model the past and future evolution of landscapes, rivers, lakes and associated groundwater systems in a changing climate. However, there is presently a paucity of 3D numerical studies of permafrost thaw and associated hydrological changes, and the lack of study can be partly attributed to the difficulty in verifying multi-dimensional results produced by numerical models. Numerical approaches can only be validated against analytical solutions for a purely thermic 1D equation with phase change (e.g. Neumann, Lunardini). When it comes to the coupled TH system (coupling two highly non-linear equations), the only possible approach is to compare the results from different codes to provided test cases and/or to have controlled experiments for validation. Such inter-code comparisons can propel discussions to try to improve code performances. A benchmark exercise was initialized in 2014 with a kick-off meeting in Paris in November. Participants from USA, Canada, Germany, Sweden and France convened, representing altogether 13 simulation codes. The benchmark exercises consist of several test cases inspired by existing literature (e.g. McKenzie et al., 2007) as well as new ones. They

  1. Applications of the Los Alamos High Energy Transport code

    International Nuclear Information System (INIS)

    Waters, L.; Gavron, A.; Prael, R.E.

    1992-01-01

    Simulation codes reliable through a large range of energies are essential to analyze the environment of vehicles and habitats proposed for space exploration. The LAHET monte carlo code has recently been expanded to track high energy hadrons with FLUKA, while retaining the original Los Alamos version of HETC at lower energies. Electrons and photons are transported with EGS4, and an interface to the MCNP monte carlo code is provided to analyze neutrons with kinetic energies less than 20 MeV. These codes are benchmarked by comparison of LAHET/MCNP calculations to data from the Brookhaven experiment E814 participant calorimeter

  2. Toward a CFD nose-to-tail capability - Hypersonic unsteady Navier-Stokes code validation

    Science.gov (United States)

    Edwards, Thomas A.; Flores, Jolen

    1989-01-01

    Computational fluid dynamics (CFD) research for hypersonic flows presents new problems in code validation because of the added complexity of the physical models. This paper surveys code validation procedures applicable to hypersonic flow models that include real gas effects. The current status of hypersonic CFD flow analysis is assessed with the Compressible Navier-Stokes (CNS) code as a case study. The methods of code validation discussed to beyond comparison with experimental data to include comparisons with other codes and formulations, component analyses, and estimation of numerical errors. Current results indicate that predicting hypersonic flows of perfect gases and equilibrium air are well in hand. Pressure, shock location, and integrated quantities are relatively easy to predict accurately, while surface quantities such as heat transfer are more sensitive to the solution procedure. Modeling transition to turbulence needs refinement, though preliminary results are promising.

  3. ClinicalCodes: an online clinical codes repository to improve the validity and reproducibility of research using electronic medical records.

    Science.gov (United States)

    Springate, David A; Kontopantelis, Evangelos; Ashcroft, Darren M; Olier, Ivan; Parisi, Rosa; Chamapiwa, Edmore; Reeves, David

    2014-01-01

    Lists of clinical codes are the foundation for research undertaken using electronic medical records (EMRs). If clinical code lists are not available, reviewers are unable to determine the validity of research, full study replication is impossible, researchers are unable to make effective comparisons between studies, and the construction of new code lists is subject to much duplication of effort. Despite this, the publication of clinical codes is rarely if ever a requirement for obtaining grants, validating protocols, or publishing research. In a representative sample of 450 EMR primary research articles indexed on PubMed, we found that only 19 (5.1%) were accompanied by a full set of published clinical codes and 32 (8.6%) stated that code lists were available on request. To help address these problems, we have built an online repository where researchers using EMRs can upload and download lists of clinical codes. The repository will enable clinical researchers to better validate EMR studies, build on previous code lists and compare disease definitions across studies. It will also assist health informaticians in replicating database studies, tracking changes in disease definitions or clinical coding practice through time and sharing clinical code information across platforms and data sources as research objects.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-15

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  6. 10Gbps 2D MGC OCDMA Code over FSO Communication System

    Science.gov (United States)

    Professor Urmila Bhanja, Associate, Dr.; Khuntia, Arpita; Alamasety Swati, (Student

    2017-08-01

    Currently, wide bandwidth signal dissemination along with low latency is a leading requisite in various applications. Free space optical wireless communication has introduced as a realistic technology for bridging the gap in present high data transmission fiber connectivity and as a provisional backbone for rapidly deployable wireless communication infrastructure. The manuscript highlights on the implementation of 10Gbps SAC-OCDMA FSO communications using modified two dimensional Golomb code (2D MGC) that possesses better auto correlation, minimum cross correlation and high cardinality. A comparison based on pseudo orthogonal (PSO) matrix code and modified two dimensional Golomb code (2D MGC) is developed in the proposed SAC OCDMA-FSO communication module taking different parameters into account. The simulative outcome signifies that the communication radius is bounded by the multiple access interference (MAI). In this work, a comparison is made in terms of bit error rate (BER), and quality factor (Q) based on modified two dimensional Golomb code (2D MGC) and PSO matrix code. It is observed that the 2D MGC yields better results compared to the PSO matrix code. The simulation results are validated using optisystem version 14.

  7. Rapid installation of numerical models in multiple parent codes

    Energy Technology Data Exchange (ETDEWEB)

    Brannon, R.M.; Wong, M.K.

    1996-10-01

    A set of``model interface guidelines``, called MIG, is offered as a means to more rapidly install numerical models (such as stress-strain laws) into any parent code (hydrocode, finite element code, etc.) without having to modify the model subroutines. The model developer (who creates the model package in compliance with the guidelines) specifies the model`s input and storage requirements in a standardized way. For portability, database management (such as saving user inputs and field variables) is handled by the parent code. To date, NUG has proved viable in beta installations of several diverse models in vectorized and parallel codes written in different computer languages. A NUG-compliant model can be installed in different codes without modifying the model`s subroutines. By maintaining one model for many codes, MIG facilitates code-to-code comparisons and reduces duplication of effort potentially reducing the cost of installing and sharing models.

  8. Method for quantitative assessment of nuclear safety computer codes

    International Nuclear Information System (INIS)

    Dearien, J.A.; Davis, C.B.; Matthews, L.J.

    1979-01-01

    A procedure has been developed for the quantitative assessment of nuclear safety computer codes and tested by comparison of RELAP4/MOD6 predictions with results from two Semiscale tests. This paper describes the developed procedure, the application of the procedure to the Semiscale tests, and the results obtained from the comparison

  9. Accuracy assessment of a new Monte Carlo based burnup computer code

    International Nuclear Information System (INIS)

    El Bakkari, B.; ElBardouni, T.; Nacir, B.; ElYounoussi, C.; Boulaich, Y.; Meroun, O.; Zoubair, M.; Chakir, E.

    2012-01-01

    Highlights: ► A new burnup code called BUCAL1 was developed. ► BUCAL1 uses the MCNP tallies directly in the calculation of the isotopic inventories. ► Validation of BUCAL1 was done by code to code comparison using VVER-1000 LEU Benchmark Assembly. ► Differences from BM value were found to be ± 600 pcm for k ∞ and ±6% for the isotopic compositions. ► The effect on reactivity due to the burnup of Gd isotopes is well reproduced by BUCAL1. - Abstract: This study aims to test for the suitability and accuracy of a new home-made Monte Carlo burnup code, called BUCAL1, by investigating and predicting the neutronic behavior of a “VVER-1000 LEU Assembly Computational Benchmark”, at lattice level. BUCAL1 uses MCNP tally information directly in the computation; this approach allows performing straightforward and accurate calculation without having to use the calculated group fluxes to perform transmutation analysis in a separate code. ENDF/B-VII evaluated nuclear data library was used in these calculations. Processing of the data library is performed using recent updates of NJOY99 system. Code to code comparisons with the reported Nuclear OECD/NEA results are presented and analyzed.

  10. Analysis of ATLAS FLB-EC6 Experiment using SPACE Code

    International Nuclear Information System (INIS)

    Lee, Donghyuk; Kim, Yohan; Kim, Seyun

    2013-01-01

    The new code is named SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). As a part of code validation effort, simulation of ATLAS FLB(Feedwater Line Break) experiment using SPACE code has been performed. The FLB-EC6 experiment is economizer break of a main feedwater line. The calculated results using the SPACE code are compared with those from the experiment. The ATLAS FLB-EC6 experiment, which is economizer feedwater line break, was simulated using the SPACE code. The calculated results were compared with those from the experiment. The comparisons of break flow rate and steam generator water level show good agreement with the experiment. The SPACE code is capable of predicting physical phenomena occurring during ATLAS FLB-EC6 experiment

  11. Gamma streaming experiments for validation of Monte Carlo code

    International Nuclear Information System (INIS)

    Thilagam, L.; Mohapatra, D.K.; Subbaiah, K.V.; Iliyas Lone, M.; Balasubramaniyan, V.

    2012-01-01

    In-homogeneities in shield structures lead to considerable amount of leakage radiation (streaming) increasing the radiation levels in accessible areas. Development works on experimental as well as computational methods for quantifying this streaming radiation are still continuing. Monte Carlo based radiation transport code, MCNP is usually a tool for modeling and analyzing such problems involving complex geometries. In order to validate this computational method for streaming analysis, it is necessary to carry out some experimental measurements simulating these inhomogeneities like ducts and voids present in the bulk shields for typical cases. The data thus generated will be analysed by simulating the experimental set up employing MCNP code and optimized input parameters for the code in finding solutions for similar radiation streaming problems will be formulated. Comparison of experimental data obtained from radiation streaming experiments through ducts will give a set of thumb rules and analytical fits for total radiation dose rates within and outside the duct. The present study highlights the validation of MCNP code through the gamma streaming experiments carried out with the ducts of various shapes and dimensions. Over all, the present study throws light on suitability of MCNP code for the analysis of gamma radiation streaming problems for all duct configurations considered. In the present study, only dose rate comparisons have been made. Studies on spectral comparison of streaming radiation are in process. Also, it is planned to repeat the experiments with various shield materials. Since the penetrations and ducts through bulk shields are unavoidable in an operating nuclear facility the results on this kind of radiation streaming simulations and experiments will be very useful in the shield structure optimization without compromising the radiation safety

  12. Comparison report of the OECD/CSNI international standard problem 21 (Piper-one experiment PO-SB-7). Volume 1 comparison report. Volume 2 evaluation of code accuracy in the prediction of ISP 21

    International Nuclear Information System (INIS)

    1989-11-01

    The present report deals with the comparison of 6 blind predictions, submitted by 5 participants, and the experimental results measured during the test PO-SB-7 performed in PIPER-ONE facility. The PIPER-ONE apparatus is an experimental simulator of a General Electric BWR. The test PO-SB-7 simulates a SB-LOCA originated by a break in one recirculation line of the reference BWR-6 plant, without intervention of high pressure ECCS. The overall activity constitutes the CSNI ISP-21. The main parts of the report are: a) outline of the test facility and of the PO-SB-7 experiment; b) overview of input models used by participants; c) evaluation of participant predictions on the basis of one-by-one comparison with selected experimental trends; d) evaluation of present code capabilities and accuracy, on the basis of the overall comparison between measured data and participants double blind predictions. Finally, a judgement is given in relation to the overall value of the activity

  13. OECD International Standard Problem number 34. Falcon code comparison report

    International Nuclear Information System (INIS)

    Williams, D.A.

    1994-12-01

    ISP-34 is the first ISP to address fission product transport issues and has been strongly supported by a large number of different countries and organisations. The ISP is based on two experiments, FAL-ISP-1 and FAL-ISP-2, which were conducted in AEA's Falcon facility. Specific features of the experiments include quantification of chemical effects and aerosol behaviour. In particular, multi-component aerosol effects and vapour-aerosol interactions can all be investigated in the Falcon facility. Important parameters for participants to predict were the deposition profiles and composition, key chemical species and reactions, evolution of suspended material concentrations, and the effects of steam condensation onto aerosols and particle hygroscopicity. The results of the Falcon ISP support the belief that aerosol physics is generally well modelled in primary circuit codes, but the chemistry models in many of the codes need to be improved, since chemical speciation is one of the main factors which controls transport and deposition behaviour. The importance of chemical speciation, aerosol nucleation, and the role of multi-component aerosols in determining transport and deposition behaviour are evident. The role of re-vaporization in these Falcon experiments is not clear; it is not possible to compare those codes which predicted re-vaporization with quantitative data. The evidence from this ISP exercise indicates that the containment codes can predict thermal-hydraulics conditions satisfactorily. However, the differences in the predicted aerosol locations in the Falcon tests had shown that aerosol behaviour was very susceptible to parameters such as particle size distribution

  14. 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

  15. 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.

  16. 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

  17. 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

  18. 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.

  19. A computer code package for Monte Carlo photon-electron transport simulation Comparisons with experimental benchmarks

    International Nuclear Information System (INIS)

    Popescu, Lucretiu M.

    2000-01-01

    A computer code package (PTSIM) for particle transport Monte Carlo simulation was developed using object oriented techniques of design and programming. A flexible system for simulation of coupled photon, electron transport, facilitating development of efficient simulation applications, was obtained. For photons: Compton and photo-electric effects, pair production and Rayleigh interactions are simulated, while for electrons, a class II condensed history scheme was considered, in which catastrophic interactions (Moeller electron-electron interaction, bremsstrahlung, etc.) are treated in detail and all other interactions with reduced individual effect on electron history are grouped together using continuous slowing down approximation and energy straggling theories. Electron angular straggling is simulated using Moliere theory or a mixed model in which scatters at large angles are treated as distinct events. Comparisons with experimentally benchmarks for electron transmission and bremsstrahlung emissions energy and angular spectra, and for dose calculations are presented

  20. Comparing the coding of complications in Queensland and Victorian admitted patient data.

    Science.gov (United States)

    Michel, Jude L; Cheng, Diana; Jackson, Terri J

    2011-08-01

    To examine differences between Queensland and Victorian coding of hospital-acquired conditions and suggest ways to improve the usefulness of these data in the monitoring of patient safety events. Secondary analysis of admitted patient episode data collected in Queensland and Victoria. Comparison of depth of coding, and patterns in the coding of ten commonly coded complications of five elective procedures. Comparison of the mean complication codes assigned per episode revealed Victoria assigns more valid codes than Queensland for all procedures, with the difference between the states being significantly different in all cases. The proportion of the codes flagged as complications was consistently lower for Queensland when comparing 10 common complications for each of the five selected elective procedures. The estimated complication rates for the five procedures showed Victoria to have an apparently higher complication rate than Queensland for 35 of the 50 complications examined. Our findings demonstrate that the coding of complications is more comprehensive in Victoria than in Queensland. It is known that inconsistencies exist between states in routine hospital data quality. Comparative use of patient safety indicators should be viewed with caution until standards are improved across Australia. More exploration of data quality issues is needed to identify areas for improvement.

  1. Modification and application of TOUGH2 as a variable-density, saturated-flow code and comparison to SWIFT II results

    International Nuclear Information System (INIS)

    Christian-Frear, T.L.; Webb, S.W.

    1995-01-01

    Human intrusion scenarios at the Waste Isolation Pilot Plant (WIPP) involve penetration of the repository and an underlying brine reservoir by a future borehole. Brine and gas from the brine reservoir and the repository may flow up the borehole and into the overlying Culebra formation, which is saturated with water containing different amounts of dissolved 'solids resulting in a spatially varying density. Current modeling approaches involve perturbing a steady-state Culebra flow field by inflow of gas and/or brine from a breach borehole that has passed through the repository. Previous studies simulating steady-state flow in the Culebra have been done. One specific study by LaVenue et al. (1990) used the SWIFT 2 code, a single-phase flow and transport code, to develop the steady-state flow field. Because gas may also be present in the fluids from the intrusion borehole, a two-phase code such as TOUGH2 can be used to determine the effect that emitted fluids may have on the steady-state Culebra flow field. Thus a comparison between TOUGH2 and SWIFT2 was prompted. In order to compare the two codes and to evaluate the influence of gas on flow in the Culebra, modifications were made to TOUGH2. Modifications were performed by the authors to allow for element-specific values of permeability, porosity, and elevation. The analysis also used a new equation of state module for a water-brine-air mixture, EOS7 (Pruess, 1991), which was developed to simulate variable water densities by assuming a miscible mixture of water and brine phases and allows for element-specific brine concentration in the INCON file

  2. SWAAM code development, verification and application to steam generator design

    International Nuclear Information System (INIS)

    Shin, Y.W.; Valentin, R.A.

    1990-01-01

    This paper describes the family of SWAAM codes developed by Argonne National Laboratory to analyze the effects of sodium/water reactions on LMR steam generators. The SWAAM codes were developed as design tools for analyzing various phenomena related to steam generator leaks and to predict the resulting thermal and hydraulic effects on the steam generator and the intermediate heat transport system (IHTS). The theoretical foundations and numerical treatments on which the codes are based are discussed, followed by a description of code capabilities and limitations, verification of the codes by comparison with experiment, and applications to steam generator and IHTS design. (author). 25 refs, 14 figs

  3. Two dimension MDW OCDMA code cross-correlation for reduction of phase induced intensity noise

    Directory of Open Access Journals (Sweden)

    Sh. Ahmed Israa

    2017-01-01

    Full Text Available In this paper, we first review 2-D MDW code cross correlation equations and table to be improved significantly by using code correlation properties. These codes can be used in the synchronous optical CDMA systems for multi access interference cancellation and maximum suppress the phase induced intensity noise. Low Psr is due to the reduction of interference noise that is induced by the 2-D MDW code PIIN suppression. High data rate causes increases in BER, requires high effective power and severely deteriorates the system performance. The 2-D W/T MDW code has an excellent system performance where the value of PIIN is suppressed as low as possible at the optimum Psr with high data bit rate. The 2-D MDW code shows better tolerance to PIIN in comparison to others with enhanced system performance. We prove by numerical analysis that the PIIN maximally suppressed by MDW code through the minimizing property of cross correlation in comparison to 2-D PDC and 2-D MQC OCDMA code.scheme systems.

  4. Two dimension MDW OCDMA code cross-correlation for reduction of phase induced intensity noise

    Science.gov (United States)

    Ahmed, Israa Sh.; Aljunid, Syed A.; Nordin, Junita M.; Dulaimi, Layth A. Khalil Al; Matem, Rima

    2017-11-01

    In this paper, we first review 2-D MDW code cross correlation equations and table to be improved significantly by using code correlation properties. These codes can be used in the synchronous optical CDMA systems for multi access interference cancellation and maximum suppress the phase induced intensity noise. Low Psr is due to the reduction of interference noise that is induced by the 2-D MDW code PIIN suppression. High data rate causes increases in BER, requires high effective power and severely deteriorates the system performance. The 2-D W/T MDW code has an excellent system performance where the value of PIIN is suppressed as low as possible at the optimum Psr with high data bit rate. The 2-D MDW code shows better tolerance to PIIN in comparison to others with enhanced system performance. We prove by numerical analysis that the PIIN maximally suppressed by MDW code through the minimizing property of cross correlation in comparison to 2-D PDC and 2-D MQC OCDMA code.scheme systems.

  5. Improved Intra-coding Methods for H.264/AVC

    Directory of Open Access Journals (Sweden)

    Li Song

    2009-01-01

    Full Text Available The H.264/AVC design adopts a multidirectional spatial prediction model to reduce spatial redundancy, where neighboring pixels are used as a prediction for the samples in a data block to be encoded. In this paper, a recursive prediction scheme and an enhanced (block-matching algorithm BMA prediction scheme are designed and integrated into the state-of-the-art H.264/AVC framework to provide a new intra coding model. Extensive experiments demonstrate that the coding efficiency can be on average increased by 0.27 dB with comparison to the performance of the conventional H.264 coding model.

  6. Validation of thermohydraulic codes by comparison of experimental results with computer simulations

    International Nuclear Information System (INIS)

    Madeira, A.A.; Galetti, M.R.S.; Pontedeiro, A.C.

    1989-01-01

    The results obtained by simulation of three cases from CANON depressurization experience, using the TRAC-PF1 computer code, version 7.6, implanted in the VAX-11/750 computer of Brazilian CNEN, are presented. The CANON experience was chosen as first standard problem in thermo-hydraulic to be discussed at ENFIR for comparing results from different computer codes with results obtained experimentally. The ability of TRAC-PF1 code to prevent the depressurization phase of a loss of primary collant accident in pressurized water reactors is evaluated. (M.C.K.) [pt

  7. Description and application of the AERIN Code at LLNL

    International Nuclear Information System (INIS)

    King, W.C.

    1986-01-01

    The AERIN code was written at the Lawrence Livermore National Laboratory in 1976 to compute the organ burdens and absorbed dose resulting from a chronic or acute inhalation of transuranic isotopes. The code was revised in 1982 to reflect the concepts of ICRP-30. This paper will describe the AERIN code and how it has been used at LLNL to study more than 80 cases of internal deposition and obtain estimates of internal dose. A comparison with the computed values of the committed organ dose is made with ICRP-30 values. The benefits of using the code are described. 3 refs., 3 figs., 6 tabs

  8. Context based Coding of Quantized Alpha Planes for Video Objects

    DEFF Research Database (Denmark)

    Aghito, Shankar Manuel; Forchhammer, Søren

    2002-01-01

    In object based video, each frame is a composition of objects that are coded separately. The composition is performed through the alpha plane that represents the transparency of the object. We present an alternative to MPEG-4 for coding of alpha planes that considers their specific properties....... Comparisons in terms of rate and distortion are provided, showing that the proposed coding scheme for still alpha planes is better than the algorithms for I-frames used in MPEG-4....

  9. Comparison of TRAC calculations with experimental data

    International Nuclear Information System (INIS)

    Jackson, J.F.; Vigil, J.C.

    1980-01-01

    TRAC is an advanced best-estimate computer code for analyzing postulated accidents in light water reactors. This paper gives a brief description of the code followed by comparisons of TRAC calculations with data from a variety of separate-effects, system-effects, and integral experiments. Based on these comparisons, the capabilities and limitations of the early versions of TRAC are evaluated

  10. Preparation of the TRANSURANUS code for TEMELIN NPP

    International Nuclear Information System (INIS)

    Klouzal, J.

    2011-01-01

    Since 2010 Temelin NPP started using TVSA-T fuel supplied by JSC TVEL. The transition process included implementation of several new core reload design codes. TRANSURANUS code was selected for the evaluation of the fuel rod thermomechanical performance. The adaptation and validation of the code was performed by Nuclear Research Institute Rez. TRANSURANUS code contains wide selection of alternative models for most of phenomena important for the fuel behaviour. It was therefore necessary to select, based on a comparison with experimental data, those most suitable for the modeling of TVSA-T fuel rods. In some cases, new models were implemented. Software tools and methodology for the evaluation of the proposed core reload design using TRANSURANUS code were also developed in NRI. The software tools include the interface to core physics code ANDREA and a set of scripts for an automated execution and processing of the computational runs. Independent confirmation of some of the vendor specified core reload design criteria was performed using TRANSURANUS. (authors)

  11. Evaluation of the plastic characteristics of piping products in relation to ASME code criteria

    International Nuclear Information System (INIS)

    Rodabaugh, E.C.; Moore, S.E.

    1978-07-01

    Theories and test data relevant to the plastic characteristics of piping products are presented and compared with Code Equations in NB-3652 for Class 1 piping; in NC/ND-3652.2 for Class 2 and Class 3 piping. Comparisons are made for (a) straight pipe, (b) elbows, (c) branch connections, and (d) tees. The status of data (or lack of data) for other piping components is discussed. Comparisons are made between available data and the Code equations for two typical piping materials, SA106 Grade B and SA312 TP304, for Code Design Limits, and Service Limits A, B, C, and D. Conditions under which the Code Limits cannot be shown to be conservative from available data are pointed out. Based on the results of the study, recommendations for Code revisions are presented, along with recommendations for additional work

  12. International standard problem (ISP) no. 41 follow up exercise: Containment iodine computer code exercise: parametric studies

    Energy Technology Data Exchange (ETDEWEB)

    Ball, J.; Glowa, G.; Wren, J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Ewig, F. [GRS Koln (Germany); Dickenson, S. [AEAT, (United Kingdom); Billarand, Y.; Cantrel, L. [IPSN (France); Rydl, A. [NRIR (Czech Republic); Royen, J. [OECD/NEA (France)

    2001-11-01

    This report describes the results of the second phase of International Standard Problem (ISP) 41, an iodine behaviour code comparison exercise. The first phase of the study, which was based on a simple Radioiodine Test Facility (RTF) experiment, demonstrated that all of the iodine behaviour codes had the capability to reproduce iodine behaviour for a narrow range of conditions (single temperature, no organic impurities, controlled pH steps). The current phase, a parametric study, was designed to evaluate the sensitivity of iodine behaviour codes to boundary conditions such as pH, dose rate, temperature and initial I{sup -} concentration. The codes used in this exercise were IODE(IPSN), IODE(NRIR), IMPAIR(GRS), INSPECT(AEAT), IMOD(AECL) and LIRIC(AECL). The parametric study described in this report identified several areas of discrepancy between the various codes. In general, the codes agree regarding qualitative trends, but their predictions regarding the actual amount of volatile iodine varied considerably. The largest source of the discrepancies between code predictions appears to be their different approaches to modelling the formation and destruction of organic iodides. A recommendation arising from this exercise is that an additional code comparison exercise be performed on organic iodide formation, against data obtained front intermediate-scale studies (two RTF (AECL, Canada) and two CAIMAN facility, (IPSN, France) experiments have been chosen). This comparison will allow each of the code users to realistically evaluate and improve the organic iodide behaviour sub-models within their codes. (author)

  13. International standard problem (ISP) no. 41 follow up exercise: Containment iodine computer code exercise: parametric studies

    International Nuclear Information System (INIS)

    Ball, J.; Glowa, G.; Wren, J.; Ewig, F.; Dickenson, S.; Billarand, Y.; Cantrel, L.; Rydl, A.; Royen, J.

    2001-11-01

    This report describes the results of the second phase of International Standard Problem (ISP) 41, an iodine behaviour code comparison exercise. The first phase of the study, which was based on a simple Radioiodine Test Facility (RTF) experiment, demonstrated that all of the iodine behaviour codes had the capability to reproduce iodine behaviour for a narrow range of conditions (single temperature, no organic impurities, controlled pH steps). The current phase, a parametric study, was designed to evaluate the sensitivity of iodine behaviour codes to boundary conditions such as pH, dose rate, temperature and initial I - concentration. The codes used in this exercise were IODE(IPSN), IODE(NRIR), IMPAIR(GRS), INSPECT(AEAT), IMOD(AECL) and LIRIC(AECL). The parametric study described in this report identified several areas of discrepancy between the various codes. In general, the codes agree regarding qualitative trends, but their predictions regarding the actual amount of volatile iodine varied considerably. The largest source of the discrepancies between code predictions appears to be their different approaches to modelling the formation and destruction of organic iodides. A recommendation arising from this exercise is that an additional code comparison exercise be performed on organic iodide formation, against data obtained front intermediate-scale studies (two RTF (AECL, Canada) and two CAIMAN facility, (IPSN, France) experiments have been chosen). This comparison will allow each of the code users to realistically evaluate and improve the organic iodide behaviour sub-models within their codes. (author)

  14. Monte Carlo code criticality benchmark comparisons for waste packaging

    International Nuclear Information System (INIS)

    Alesso, H.P.; Annese, C.E.; Buck, R.M.; Pearson, J.S.; Lloyd, W.R.

    1992-07-01

    COG is a new point-wise Monte Carlo code being developed and tested at Lawrence Livermore National Laboratory (LLNL). It solves the Boltzmann equation for the transport of neutrons and photons. The objective of this paper is to report on COG results for criticality benchmark experiments both on a Cray mainframe and on a HP 9000 workstation. COG has been recently ported to workstations to improve its accessibility to a wider community of users. COG has some similarities to a number of other computer codes used in the shielding and criticality community. The recently introduced high performance reduced instruction set (RISC) UNIX workstations provide computational power that approach mainframes at a fraction of the cost. A version of COG is currently being developed for the Hewlett Packard 9000/730 computer with a UNIX operating system. Subsequent porting operations will move COG to SUN, DEC, and IBM workstations. In addition, a CAD system for preparation of the geometry input for COG is being developed. In July 1977, Babcock ampersand Wilcox Co. (B ampersand W) was awarded a contract to conduct a series of critical experiments that simulated close-packed storage of LWR-type fuel. These experiments provided data for benchmarking and validating calculational methods used in predicting K-effective of nuclear fuel storage in close-packed, neutron poisoned arrays. Low enriched UO2 fuel pins in water-moderated lattices in fuel storage represent a challenging criticality calculation for Monte Carlo codes particularly when the fuel pins extend out of the water. COG and KENO calculational results of these criticality benchmark experiments are presented

  15. 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)

  16. Library of neutron cross sections of the Thermos code

    International Nuclear Information System (INIS)

    Alonso V, G.; Hernandez L, H.

    1991-10-01

    The present work is the complement of the IT.SN/DFR-017 report in which the structure and the generation of the library of the Thermos code is described. In this report the comparison among the values of the cross sections that has the current library of the Thermos code and those generated by means of the ENDF-B/NJOY it is shown. (Author)

  17. The FOCON96 1.0 computer code

    International Nuclear Information System (INIS)

    Merle-Szeremeta, A.; Thomassin, A.

    1999-01-01

    The Institute of Protection and Nuclear Safety (I.P.S.N.) has developed a computer code, FOCON96 1.0 to calculate the dosimetric consequences of atmospheric radioactive releases from nuclear installations after several years of usual operation. This communication describes the principal characteristics of FOCON96 1.0 and its functionalities. The principal elements of a comparison between FOCON96 1.0 and PC-CREAM ( European computer code developed by the N.R.P.B. and answering the same criteria) are given here. (N.C.)

  18. 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

  19. Comparative calculations on selected two-phase flow phenomena using major PWR system codes

    International Nuclear Information System (INIS)

    1990-01-01

    In 1988 a comparative study on important features and models in six major best estimate thermal hydraulic codes for PWR systems was implemented (Comparison of thermal hydraulic safety codes for PWR Graham, Trotman, London, EUR 11522). It was a limitation of that study that the source codes themselves were not available but the comparison had to be based on the available documentation. In the present study, the source codes were available and the capability of four system codes to predict complex two-phase flow phenomena has been assessed. Two areas of investigation were selected: (a) pressurized spray phenomena; (b) boil-up phenomena in rod bundles. As regards the first area, experimental data obtained in 1972 on the Neptunus Facility (Delft University of Technology) were compared with the results of the calculations using Athlet, Cathare, Relap 5 and TRAC-PT1 and, concerning the second area, the results of two experimental facilities obtained in 1980 and 1985 on Thetis (UKEA) and Pericles (CEA-Grenoble) were considered

  20. Benchmarking NNWSI flow and transport codes: COVE 1 results

    International Nuclear Information System (INIS)

    Hayden, N.K.

    1985-06-01

    The code verification (COVE) activity of the Nevada Nuclear Waste Storage Investigations (NNWSI) Project is the first step in certification of flow and transport codes used for NNWSI performance assessments of a geologic repository for disposing of high-level radioactive wastes. The goals of the COVE activity are (1) to demonstrate and compare the numerical accuracy and sensitivity of certain codes, (2) to identify and resolve problems in running typical NNWSI performance assessment calculations, and (3) to evaluate computer requirements for running the codes. This report describes the work done for COVE 1, the first step in benchmarking some of the codes. Isothermal calculations for the COVE 1 benchmarking have been completed using the hydrologic flow codes SAGUARO, TRUST, and GWVIP; the radionuclide transport codes FEMTRAN and TRUMP; and the coupled flow and transport code TRACR3D. This report presents the results of three cases of the benchmarking problem solved for COVE 1, a comparison of the results, questions raised regarding sensitivities to modeling techniques, and conclusions drawn regarding the status and numerical sensitivities of the codes. 30 refs

  1. 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.

  2. Comparison of WDM/Pulse-Position-Modulation (WDM/PPM) with Code/Pulse-Position-Swapping (C/PPS) Based on Wavelength/Time Codes

    Energy Technology Data Exchange (ETDEWEB)

    Mendez, A J; Hernandez, V J; Gagliardi, R M; Bennett, C V

    2009-06-19

    Pulse position modulation (PPM) signaling is favored in intensity modulated/direct detection (IM/DD) systems that have average power limitations. Combining PPM with WDM over a fiber link (WDM/PPM) enables multiple accessing and increases the link's throughput. Electronic bandwidth and synchronization advantages are further gained by mapping the time slots of PPM onto a code space, or code/pulse-position-swapping (C/PPS). The property of multiple bits per symbol typical of PPM can be combined with multiple accessing by using wavelength/time [W/T] codes in C/PPS. This paper compares the performance of WDM/PPM and C/PPS for equal wavelengths and bandwidth.

  3. 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

  4. COCOA code for creating mock observations of star cluster models

    Science.gov (United States)

    Askar, Abbas; Giersz, Mirek; Pych, Wojciech; Dalessandro, Emanuele

    2018-04-01

    We introduce and present results from the COCOA (Cluster simulatiOn Comparison with ObservAtions) code that has been developed to create idealized mock photometric observations using results from numerical simulations of star cluster evolution. COCOA is able to present the output of realistic numerical simulations of star clusters carried out using Monte Carlo or N-body codes in a way that is useful for direct comparison with photometric observations. In this paper, we describe the COCOA code and demonstrate its different applications by utilizing globular cluster (GC) models simulated with the MOCCA (MOnte Carlo Cluster simulAtor) code. COCOA is used to synthetically observe these different GC models with optical telescopes, perform point spread function photometry, and subsequently produce observed colour-magnitude diagrams. We also use COCOA to compare the results from synthetic observations of a cluster model that has the same age and metallicity as the Galactic GC NGC 2808 with observations of the same cluster carried out with a 2.2 m optical telescope. We find that COCOA can effectively simulate realistic observations and recover photometric data. COCOA has numerous scientific applications that maybe be helpful for both theoreticians and observers that work on star clusters. Plans for further improving and developing the code are also discussed in this paper.

  5. Spectral/spatial optical CDMA code based on Diagonal Eigenvalue Unity

    Science.gov (United States)

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

    2017-11-01

    A new two dimensional Diagonal Eigenvalue Unity (2D-DEU) code is developed for the spectral⧹spatial optical code division multiple access (OCDMA) system. It has a lower cross correlation value compared to two dimensional diluted perfect difference (2D-DPD), two dimensional Extended Enhanced Double Weight (2D-Extended-EDW) codes. Also, for the same code length, the number of users can be generated by the 2D-DEU code is higher than that provided by the others codes. The Bit Error Rate (BER) numerical analysis is developed by considering the effects of shot noise, phase induced intensity noise (PIIN), and thermal noise. The main result shows that BER is strongly affected by PIIN for the higher source power. The 2D-DEU code performance is compared with 2D-DPD, 2D-Extended-EDW and two dimensional multi-diagonals (2D-MD) codes. This comparison proves that the proposed 2D-DEU system outperforms the related codes.

  6. Performance Comparison of Orthogonal and Quasi-orthogonal Codes in Quasi-Synchronous Cellular CDMA Communication

    Science.gov (United States)

    Jos, Sujit; Kumar, Preetam; Chakrabarti, Saswat

    Orthogonal and quasi-orthogonal codes are integral part of any DS-CDMA based cellular systems. Orthogonal codes are ideal for use in perfectly synchronous scenario like downlink cellular communication. Quasi-orthogonal codes are preferred over orthogonal codes in the uplink communication where perfect synchronization cannot be achieved. In this paper, we attempt to compare orthogonal and quasi-orthogonal codes in presence of timing synchronization error. This will give insight into the synchronization demands in DS-CDMA systems employing the two classes of sequences. The synchronization error considered is smaller than chip duration. Monte-Carlo simulations have been carried out to verify the analytical and numerical results.

  7. Comparison of GLIMPS and HFAST Stirling engine code predictions with experimental data

    Science.gov (United States)

    Geng, Steven M.; Tew, Roy C.

    1992-01-01

    Predictions from GLIMPS and HFAST design codes are compared with experimental data for the RE-1000 and SPRE free piston Stirling engines. Engine performance and available power loss predictions are compared. Differences exist between GLIMPS and HFAST loss predictions. Both codes require engine specific calibration to bring predictions and experimental data into agreement.

  8. Comparison of GLIMPS and HFAST Stirling engine code predictions with experimental data

    International Nuclear Information System (INIS)

    Geng, S.M.; Tew, R.C.

    1994-01-01

    Predictions from GLIMPS and HFAST design codes are compared with experimental data for the RE-1000 and SPRE free-piston Stirling engines. Engine performance and available power loss predictions are compared. Differences exist between GLIMPS and HFAST loss predictions. Both codes require engine-specific calibration to bring predictions and experimental data into agreement

  9. GAPCON-THERMAL-3 code description

    International Nuclear Information System (INIS)

    Lanning, D.D.; Mohr, C.L.; Panisko, F.E.; Stewart, K.B.

    1978-01-01

    GAPCON-3 is a computer program that predicts the thermal and mechanical behavior of an operating fuel rod during its normal lifetime. The code calculates temperatures, dimensions, stresses, and strains for the fuel and the cladding in both the radial and axial directions for each step of the user specified power history. The method of weighted residuals is for the steady state temperature calculation, and is combined with a finite difference approximation of the time derivative for transient conditions. The stress strain analysis employs an iterative axisymmetric finite element procedure that includes plasticity and creep for normal and pellet-clad mechanical interaction loads. GAPCON-3 can solve steady state and operational transient problems. Comparisons of GAPCON-3 predictions to both closed form analytical solutions and actual inpile instrumented fuel rod data have demonstrated the ability of the code to calculate fuel rod behavior. GAPCON-3 features a restart capability and an associated plot package unavailable in previous GAPCON series codes

  10. GAPCON-THERMAL-3 code description

    Energy Technology Data Exchange (ETDEWEB)

    Lanning, D.D.; Mohr, C.L.; Panisko, F.E.; Stewart, K.B.

    1978-01-01

    GAPCON-3 is a computer program that predicts the thermal and mechanical behavior of an operating fuel rod during its normal lifetime. The code calculates temperatures, dimensions, stresses, and strains for the fuel and the cladding in both the radial and axial directions for each step of the user specified power history. The method of weighted residuals is for the steady state temperature calculation, and is combined with a finite difference approximation of the time derivative for transient conditions. The stress strain analysis employs an iterative axisymmetric finite element procedure that includes plasticity and creep for normal and pellet-clad mechanical interaction loads. GAPCON-3 can solve steady state and operational transient problems. Comparisons of GAPCON-3 predictions to both closed form analytical solutions and actual inpile instrumented fuel rod data have demonstrated the ability of the code to calculate fuel rod behavior. GAPCON-3 features a restart capability and an associated plot package unavailable in previous GAPCON series codes.

  11. Computing Challenges in Coded Mask Imaging

    Science.gov (United States)

    Skinner, Gerald

    2009-01-01

    This slide presaentation reviews the complications and challenges in developing computer systems for Coded Mask Imaging telescopes. The coded mask technique is used when there is no other way to create the telescope, (i.e., when there are wide fields of view, high energies for focusing or low energies for the Compton/Tracker Techniques and very good angular resolution.) The coded mask telescope is described, and the mask is reviewed. The coded Masks for the INTErnational Gamma-Ray Astrophysics Laboratory (INTEGRAL) instruments are shown, and a chart showing the types of position sensitive detectors used for the coded mask telescopes is also reviewed. Slides describe the mechanism of recovering an image from the masked pattern. The correlation with the mask pattern is described. The Matrix approach is reviewed, and other approaches to image reconstruction are described. Included in the presentation is a review of the Energetic X-ray Imaging Survey Telescope (EXIST) / High Energy Telescope (HET), with information about the mission, the operation of the telescope, comparison of the EXIST/HET with the SWIFT/BAT and details of the design of the EXIST/HET.

  12. Optical code-division multiple-access networks

    Science.gov (United States)

    Andonovic, Ivan; Huang, Wei

    1999-04-01

    This review details the approaches adopted to implement classical code division multiple access (CDMA) principles directly in the optical domain, resulting in all optical derivatives of electronic systems. There are a number of ways of realizing all-optical CDMA systems, classified as incoherent and coherent based on spreading in the time and frequency dimensions. The review covers the basic principles of optical CDMA (OCDMA), the nature of the codes used in these approaches and the resultant limitations on system performance with respect to the number of stations (code cardinality), the number of simultaneous users (correlation characteristics of the families of codes), concluding with consideration of network implementation issues. The latest developments will be presented with respect to the integration of conventional time spread codes, used in the bulk of the demonstrations of these networks to date, with wavelength division concepts, commonplace in optical networking. Similarly, implementations based on coherent correlation with the aid of a local oscillator will be detailed and comparisons between approaches will be drawn. Conclusions regarding the viability of these approaches allowing the goal of a large, asynchronous high capacity optical network to be realized will be made.

  13. RBMK-LOCA-Analyses with the ATHLET-Code

    Energy Technology Data Exchange (ETDEWEB)

    Petry, A. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH Kurfuerstendamm, Berlin (Germany); Domoradov, A.; Finjakin, A. [Research and Development Institute of Power Engineering, Moscow (Russian Federation)

    1995-09-01

    The scientific technical cooperation between Germany and Russia includes the area of adaptation of several German codes for the Russian-designed RBMK-reactor. One point of this cooperation is the adaptation of the Thermal-Hydraulic code ATHLET (Analyses of the Thermal-Hydraulics of LEaks and Transients), for RBMK-specific safety problems. This paper contains a short description of a RBMK-1000 reactor circuit. Furthermore, the main features of the thermal-hydraulic code ATHLET are presented. The main assumptions for the ATHLET-RBMK model are discussed. As an example for the application, the results of test calculations concerning a guillotine type rupture of a distribution group header are presented and discussed, and the general analysis conditions are described. A comparison with corresponding RELAP-calculations is given. This paper gives an overview on some problems posed and experience by application of Western best-estimate codes for RBMK-calculations.

  14. Optimal codes as Tanner codes with cyclic component codes

    DEFF Research Database (Denmark)

    Høholdt, Tom; Pinero, Fernando; Zeng, Peng

    2014-01-01

    In this article we study a class of graph codes with cyclic code component codes as affine variety codes. Within this class of Tanner codes we find some optimal binary codes. We use a particular subgraph of the point-line incidence plane of A(2,q) as the Tanner graph, and we are able to describe ...

  15. Analysis of results of AZTRAN and AZKIND codes for a BWR

    International Nuclear Information System (INIS)

    Bastida O, G. E.; Vallejo Q, J. A.; Galicia A, J.; Francois L, J. L.; Xolocostli M, J. V.; Rodriguez H, A.; Gomez T, A. M.

    2016-09-01

    This paper presents an analysis of results obtained from simulations performed with the neutron transport code AZTRAN and the kinetic code of neutron diffusion AZKIND, based on comparisons with models corresponding to a typical BWR, in order to verify the behavior and reliability of the values obtained with said code for its current development. For this, simulations of different geometries were made using validated nuclear codes, such as CASMO, MCNP5 and Serpent. The results obtained are considered adequate since they are comparable with those obtained and reported with other codes, based mainly on the neutron multiplication factor and the power distribution of the same. (Author)

  16. 2D edge plasma modeling extended up to the main chamber

    Energy Technology Data Exchange (ETDEWEB)

    Dekeyser, W., E-mail: wouter.dekeyser@mech.kuleuven.be [Department of Mechanical Engineering, Katholieke Universiteit Leuven, Celestijnenlaan 300A, 3001 Leuven (Belgium); Baelmans, M. [Department of Mechanical Engineering, Katholieke Universiteit Leuven, Celestijnenlaan 300A, 3001 Leuven (Belgium); Reiter, D.; Boerner, P.; Kotov, V. [Institut fuer Plasmaphysik, Forschungszentrum Juelich GmbH, EURATOM-Association, Trilateral Euregio Cluster, D-52425 Juelich (Germany)

    2011-08-01

    Far SOL plasma flow, and hence main chamber recycling and plasma surface interaction, are today still only very poorly described by current 2D fluid edge codes, such as B2, UEDGE or EDGE2D, due to a common technical limitation. We have extended the B2 plasma fluid solver in the current ITER version of B2-EIRENE (SOLPS4.3) to allow plasma solutions to be obtained up to the 'real vessel wall', at least on the basis of ad hoc far SOL transport models. We apply here the kinetic Monte Carlo Code EIRENE on such plasma solutions to study effects of this model refinement on main chamber fluxes and sputtering, for an ITER configuration. We show that main chamber sputtering may be significantly modified both due to thermalization of CX neutrals in the far SOL and poloidally highly asymmetric plasma wall contact, as compared to hitherto applied teleportation of particle fluxes across this domain.

  17. Development of a radiative divertor for DIII-D

    International Nuclear Information System (INIS)

    Allen, S.L.; Brooks, N.H.; Campbell, R.B.; Fenstermacher, M.E.; Hill, D.N.; Hyatt, A.W.; Knoll, D.; Lasnier, C.J.; Lazarus, E.A.; Leonard, A.W.; Lippmann, S.I.; Mahdavi, M.A.; Maingi, R.; Meyer, W.; Moyer, R.A.; Petrie, T.W.; Porter, G.D.; Rensink, M.E.; Rognlien, T.D.; Schaffer, M.J.; Smith, J.P.; Staebler, G.M.; Stambaugh, R.D.; West, W.P.; Wood, R.D.

    1995-01-01

    We have used experiments and modeling to develop a new radiative divertor configuration for DIII-D. Gas puffing experiments with the existing open divertor have shown the creation of a localized ( similar 10 cm diameter) radiation zone which results in substantial reduction (3-10) in the divertor heat flux while τ E remains similar 2 times ITER-89P scaling. However, n e increases with D 2 puffing, and Z eff increases with neon puffing. Divertor structures are required to minimize the effects on the core plasma. The UEDGE fluid code, benchmarked with DIII-D data, and the DEGAS neutrals transport code are used to estimate the effectiveness of divertor configurations; slots reduce the core ionization more than baffles. The overall divertor shape is set by confinement studies which indicate that high triangularity (δ∼0.8) is important for high τ E VH-modes. Results from engineering feasibility studies, including diagnostic access, will be presented. ((orig.))

  18. Development of a radiative divertor for DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Allen, S.L. [Lawrence Livermore National Lab., CA (United States); Brooks, N.H. [General Atomics, San Diego, CA (United States); Campbell, R.B. [Sandia National Labs., Albuquerque, NM (United States); Fenstermacher, M.E. [Lawrence Livermore National Lab., CA (United States); Hill, D.N. [Lawrence Livermore National Lab., CA (United States); Hyatt, A.W. [General Atomics, San Diego, CA (United States); Knoll, D.; Lasnier, C.J. [Lawrence Livermore National Lab., CA (United States); Lazarus, E.A. [Oak Ridge National Lab., TN (United States); Leonard, A.W. [General Atomics, San Diego, CA (United States); Lippmann, S.I. [General Atomics, San Diego, CA (United States); Mahdavi, M.A. [General Atomics, San Diego, CA (United States); Maingi, R. [Oak Ridge National Lab., TN (United States); Meyer, W. [Lawrence Livermore National Lab., CA (United States); Moyer, R.A. [California Univ., Los Angeles, CA (United States); Petrie, T.W. [General Atomics, San Diego, CA (United States); Porter, G.D. [Lawrence Livermore National Lab., CA (United States); Rensink, M.E. [Lawrence Livermore National Lab., CA (United States); Rognlien, T.D. [Lawrence Livermore National Lab., CA (United States); Schaffer, M.J. [General Atomics, San Diego, CA (United States); Smith, J.P. [General Atomics, San Diego, CA (United States); Staebler, G.M. [General Atomics, San Diego, CA (United States); Stambaugh, R.D. [General Atomics, San Diego, CA (United States); West, W.P. [General Atomics, San Diego, CA (United States); Wood, R.D. [Lawrence Livermore National Lab., CA (United States)

    1995-04-01

    We have used experiments and modeling to develop a new radiative divertor configuration for DIII-D. Gas puffing experiments with the existing open divertor have shown the creation of a localized ( similar 10 cm diameter) radiation zone which results in substantial reduction (3-10) in the divertor heat flux while {tau}{sub E} remains similar 2 times ITER-89P scaling. However, n{sub e} increases with D{sub 2} puffing, and Z{sub eff} increases with neon puffing. Divertor structures are required to minimize the effects on the core plasma. The UEDGE fluid code, benchmarked with DIII-D data, and the DEGAS neutrals transport code are used to estimate the effectiveness of divertor configurations; slots reduce the core ionization more than baffles. The overall divertor shape is set by confinement studies which indicate that high triangularity ({delta}{approx}0.8) is important for high {tau}{sub E} VH-modes. Results from engineering feasibility studies, including diagnostic access, will be presented. ((orig.)).

  19. Hydrogen burn assessment with the CONTAIN code

    International Nuclear Information System (INIS)

    van Rij, H.M.

    1986-01-01

    The CONTAIN computer code was developed at Sandia National Laboratories, under contract to the US Nuclear Regulatory Commission (NRC). The code is intended for calculations of containment loads during severe accidents and for prediction of the radioactive source term in the event that the containment leaks or fails. A strong point of the CONTAIN code is the continuous interaction of the thermal-hydraulics phenomena, aerosol behavior and fission product behavior. The CONTAIN code can be used for Light Water Reactors as well as Liquid Metal Reactors. In order to evaluate the CONTAIN code on its merits, comparisons between the code and experiments must be made. In this paper, CONTAIN calculations for the hydrogen burn experiments, carried out at the Nevada Test Site (NTS), are presented and compared with the experimental data. In the Large-Scale Hydrogen Combustion Facility at the NTS, 21 tests have been carried out. These tests were sponsored by the NRC and the Electric Power Research Institute (EPRI). The tests, carried out by EG and G, were performed in a spherical vessel 16 m in diameter with a design pressure of 700 kPa, substantially higher than that of most commercial nuclear containment buildings

  20. Comparison of the MAAP4 code with the station blackout simulation in the IIST facility

    International Nuclear Information System (INIS)

    Robert E Henry; Christopher E Henry; Chan Y Paik; George M Hauser

    2005-01-01

    Full text of publication follows: The Modular Accident Analysis Program (MAAP) is an integral system model to assess challenges to the reactor core, Reactor Coolant System (RCS) and containment for accident conditions. MAAP4 is the current version used by the MAAP Users Group to assess the responses to a spectrum of accident conditions. Benchmarking of the MAAP code with integral system experiments has been a continuing effort by MAAP developers and users. Several of these have been configured into dynamic benchmarks and are included in Volume III (Benchmarking) of the MAAP4 Users Manual (EPRI, 2004). One such integral experiment is the INER integral system test (IIST) constructed at the Institute of Nuclear Energy Research in Taiwan. This experimental facility is a reduced height, reduced pressure representation of a 3-loop PWR and has been used to examine several different types of accident sequences. One of these is a station blackout simulation with loss of auxiliary feedwater at the time that the transient is initiated. This is an important integral experiment to be compared with the MAAP4 code models. A parameter file (those values representing the system design and boundary experimental conditions) has been developed for the IIST facility and an input deck has been configured to represent a station blackout sequence with instantaneous loss of auxiliary feedwater. Of importance in this benchmark is (a) the rate at which the secondary side inventory is depleted, (b) the depletion of water within the reactor pressure vessel and (c) the time at which the top of the reactor core is uncovered. Comparisons have been made with these three different intervals and there is good agreement between the timing of these events for the MAAP4 benchmark. This is important since this reference sequence represents a set of boundary conditions that is continually with subsequent analyses being perturbations on this type of accident sequence. The good agreement between MAAP4 and

  1. Comparison between KARBUS and APOLLO 1

    International Nuclear Information System (INIS)

    Payer, L.; Broeders, C.

    1995-01-01

    A comparison is made between benchmark calculations by the French APOLLO 1 code and the Karlsruhe KARBUS procedure. Independently these two codes had been developed for transport computations in infinite reactor configurations and for burnup calculations. (orig.)

  2. PFLOTRAN-RepoTREND Source Term Comparison Summary.

    Energy Technology Data Exchange (ETDEWEB)

    Frederick, Jennifer M

    2018-03-01

    Code inter-comparison studies are useful exercises to verify and benchmark independently developed software to ensure proper function, especially when the software is used to model high-consequence systems which cannot be physically tested in a fully representative environment. This summary describes the results of the first portion of the code inter-comparison between PFLOTRAN and RepoTREND, which compares the radionuclide source term used in a typical performance assessment.

  3. Comparison study on flexible pavement design using FAA (Federal Aviation Administration) and LCN (Load Classification Number) code in Ahmad Yani international airport’s runway

    Science.gov (United States)

    Santoso, S. E.; Sulistiono, D.; Mawardi, A. F.

    2017-11-01

    FAA code for airport design has been broadly used by Indonesian Ministry of Aviation since decades ago. However, there is not much comprehensive study about its relevance and efficiency towards current situation in Indonesia. Therefore, a further comparison study on flexible pavement design for airport runway using comparable method has become essential. The main focus of this study is to compare which method between FAA and LCN that offer the most efficient and effective way in runway pavement planning. The comparative methods in this study mainly use the variety of variable approach. FAA code for instance, will use the approach on the aircraft’s maximum take-off weight and annual departure. Whilst LCN code use the variable of equivalent single wheel load and tire pressure. Based on the variables mentioned above, a further classification and rated method will be used to determine which code is best implemented. According to the analysis, it is clear that FAA method is the most effective way to plan runway design in Indonesia with consecutively total pavement thickness of 127cm and LCN method total pavement thickness of 70cm. Although, FAA total pavement is thicker that LCN its relevance towards sustainable and pristine condition in the future has become an essential aspect to consider in design and planning.

  4. Evaluation of an electrocardiogram on QR code.

    Science.gov (United States)

    Nakayama, Masaharu; Shimokawa, Hiroaki

    2013-01-01

    An electrocardiogram (ECG) is an indispensable tool to diagnose cardiac diseases, such as ischemic heart disease, myocarditis, arrhythmia, and cardiomyopathy. Since ECG patterns vary depend on patient status, it is also used to monitor patients during treatment and comparison with ECGs with previous results is important for accurate diagnosis. However, the comparison requires connection to ECG data server in a hospital and the availability of data connection among hospitals is limited. To improve the portability and availability of ECG data regardless of server connection, we here introduce conversion of ECG data into 2D barcodes as text data and decode of the QR code for drawing ECG with Google Chart API. Fourteen cardiologists and six general physicians evaluated the system using iPhone and iPad. Overall, they were satisfied with the system in usability and accuracy of decoded ECG compared to the original ECG. This new coding system may be useful in utilizing ECG data irrespective of server connections.

  5. Optics code development at Los Alamos

    International Nuclear Information System (INIS)

    Mottershead, C.T.; Lysenko, W.P.

    1988-01-01

    This paper is an overview of part of the beam optics code development effort in the Accelerator Technology Division at Los Alamos National Laboratory. The aim of this effort is to improve our capability to design advanced beam optics systems. The work reported is being carried out by a collaboration of permanent staff members, visiting consultants, and student research assistants. The main components of the effort are: building a new framework of common supporting utilities and software tools to facilitate further development; research and development on basic computational techniques in classical mechanics and electrodynamics; and evaluation and comparison of existing beam optics codes, and support for their continuing development. 17 refs

  6. Optics code development at Los Alamos

    International Nuclear Information System (INIS)

    Mottershead, C.T.; Lysenko, W.P.

    1988-01-01

    This paper is an overview of part of the beam optics code development effort in the Accelerator Technology Division at Los Alamos National Laboratory. The aim of this effort is to improve our capability to design advanced beam optics systems. The work reported is being carried out by a collaboration of permanent staff members, visiting consultants, and student research assistants. The main components of the effort are building a new framework of common supporting utilities and software tools to facilitate further development. research and development on basic computational techniques in classical mechanics and electrodynamics, and evaluation and comparison of existing beam optics codes, and support for their continuing development

  7. The impact of three discharge coding methods on the accuracy of diagnostic coding and hospital reimbursement for inpatient medical care.

    Science.gov (United States)

    Tsopra, Rosy; Peckham, Daniel; Beirne, Paul; Rodger, Kirsty; Callister, Matthew; White, Helen; Jais, Jean-Philippe; Ghosh, Dipansu; Whitaker, Paul; Clifton, Ian J; Wyatt, Jeremy C

    2018-07-01

    Coding of diagnoses is important for patient care, hospital management and research. However coding accuracy is often poor and may reflect methods of coding. This study investigates the impact of three alternative coding methods on the inaccuracy of diagnosis codes and hospital reimbursement. Comparisons of coding inaccuracy were made between a list of coded diagnoses obtained by a coder using (i)the discharge summary alone, (ii)case notes and discharge summary, and (iii)discharge summary with the addition of medical input. For each method, inaccuracy was determined for the primary, secondary diagnoses, Healthcare Resource Group (HRG) and estimated hospital reimbursement. These data were then compared with a gold standard derived by a consultant and coder. 107 consecutive patient discharges were analysed. Inaccuracy of diagnosis codes was highest when a coder used the discharge summary alone, and decreased significantly when the coder used the case notes (70% vs 58% respectively, p coded from the discharge summary with medical support (70% vs 60% respectively, p coding with case notes, and 35% for coding with medical support. The three coding methods resulted in an annual estimated loss of hospital remuneration of between £1.8 M and £16.5 M. The accuracy of diagnosis codes and percentage of correct HRGs improved when coders used either case notes or medical support in addition to the discharge summary. Further emphasis needs to be placed on improving the standard of information recorded in discharge summaries. Copyright © 2018 Elsevier B.V. All rights reserved.

  8. Divertor target profiles and recycling studies in TCV single null lower standard discharges

    International Nuclear Information System (INIS)

    Pitts, R.A.; Nieswand, C.; Weisen, H.

    1996-05-01

    A 'standard', single null lower diverted discharge has been developed to enable continuous monitoring of the first wall conditions and to characterise the effectiveness and influence of wall conditioning in the TCV tokamak. Measurements over a period encompassing nearly 2000 ohmic discharges of varying configuration and input power show the global confinement time and main plasma impurity concentrations to be good general indicators of the first wall condition, whilst divertor target profiles demonstrate strikingly the short term beneficial effects of He glow. Good agreement, consistent with a reduction in recycling at the plates is found between the predictions of the fluid code UEDGE and the observed outer divertor profiles of T e and n e before and after He glow. (author) 5 figs., 7 refs

  9. Challenges in edge modeling

    International Nuclear Information System (INIS)

    Schneider, R.

    2007-01-01

    Fluid models like B2, UEDGE or EDGE2D are the working horses for scrape-off layer physics, both for design and experimental support. The concept of a numerical tokamak, aiming at a predictive code for ITER, triggers the need to re-assess the available tools and their necessary extensions. These additional physics issues will be summarized from a personal point-of-view. Depending on the specific problem, several complexity levels of scrape-off layer models will be needed. Therefore, a hierarchy of tools is necessary, which will be discussed. Furthermore, the experience existing in other scientific fields with multi-scale problems and modeling should be used. Here, the coupling of different length and time scales are in particular of interest for fusion problems. (author)

  10. Radiative divertor plasmas with convection in DIII-D

    International Nuclear Information System (INIS)

    Leornard, A.W.; Porter, G.D.; Wood, R.D.

    1998-01-01

    The radiation of divertor heat flux on DIII-D is shown to greatly exceed the limits imposed by assumptions of energy transport dominated by electron thermal conduction parallel to the magnetic field. Approximately 90% of the power flowing into the divertor is dissipated through low Z radiation and plasma recombination. The dissipation is made possible by an extended region of low electron temperature in the divertor. A one-dimensional analysis of the parallel heat flux finds that the electron temperature profile is incompatible with conduction dominated parallel transport. Plasma flow at up to the ion acoustic speed, produced by upstream ionization, can account for the parallel heat flux. Modeling with the two-dimensional fluid code UEDGE has reproduced many of the observed experimental features

  11. 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

  12. Hybrid Video Coding Based on Bidimensional Matching Pursuit

    Directory of Open Access Journals (Sweden)

    Lorenzo Granai

    2004-12-01

    Full Text Available Hybrid video coding combines together two stages: first, motion estimation and compensation predict each frame from the neighboring frames, then the prediction error is coded, reducing the correlation in the spatial domain. In this work, we focus on the latter stage, presenting a scheme that profits from some of the features introduced by the standard H.264/AVC for motion estimation and replaces the transform in the spatial domain. The prediction error is so coded using the matching pursuit algorithm which decomposes the signal over an appositely designed bidimensional, anisotropic, redundant dictionary. Comparisons are made among the proposed technique, H.264, and a DCT-based coding scheme. Moreover, we introduce fast techniques for atom selection, which exploit the spatial localization of the atoms. An adaptive coding scheme aimed at optimizing the resource allocation is also presented, together with a rate-distortion study for the matching pursuit algorithm. Results show that the proposed scheme outperforms the standard DCT, especially at very low bit rates.

  13. 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)

  14. A Unique Perspective on Data Coding and Decoding

    Directory of Open Access Journals (Sweden)

    Wen-Yan Wang

    2010-12-01

    Full Text Available The concept of a loss-less data compression coding method is proposed, and a detailed description of each of its steps follows. Using the Calgary Corpus and Wikipedia data as the experimental samples and compared with existing algorithms, like PAQ or PPMstr, the new coding method could not only compress the source data, but also further re-compress the data produced by the other compression algorithms. The final files are smaller, and by comparison with the original compression ratio, at least 1% redundancy could be eliminated. The new method is simple and easy to realize. Its theoretical foundation is currently under study. The corresponding Matlab source code is provided in  the Appendix.

  15. Comparison of two LES codes for wind turbine wake studies

    DEFF Research Database (Denmark)

    Chivaee, Hamid Sarlak; Pierella, F.; Mikkelsen, Robert Flemming

    2014-01-01

    of this paper is to investigate on two CFD solvers, the DTU's in-house code, EllipSys3D and the open-sourse toolbox, OpenFoam, for a set of actuator line based LES computations. Two types of simulations are performed: the wake behind a signle rotor and the wake behind a cluster of three inline rotors. Results...... are compared in terms of velocity deficit, turbulence kinetic energy and eddy viscosity. It is seen that both codes predict similar near-wake flow structures with the exception of OpenFoam's simulations without the subgrid-scale model. The differences begin to increase with increasing the distance from...... of the wake structures and more homogenuous flow structures. It is finally observed that OpenFoam computations are more sensitive to the SGS models....

  16. Extension of the code COCOSYS to a dispersion code for smoke and carbon monoxide

    International Nuclear Information System (INIS)

    Sdouz, Gert; Mayrhofer, Robert

    2009-01-01

    The code COCOSYS (Containment Code SYStem) was developed by GRS in Germany to simulate processes and nuclear plant states during severe accidents in the containments of light water reactors. It contains several physical models, especially a module for aerosol behaviour. The goal of this work was to extend COCOSYS for applications in more general geometries mainly for complex public buildings. For the application in public buildings models for air condition systems and different boundary conditions according to different environments were developed. The principal application of the extended code COCOSYS is in the area of emergency situations especially in the simulation for carbon monoxide and smoke dispersion. After developing and implementing the new models several test calculations were performed to evaluate the functionality of the extended code. The comparison of the results with those of the original COCOSYS code showed no discrepancies. For the first realistic application several fire emergency scenarios in the Vienna General Hospital (AKH) were selected in agreement with the fire department of the hospital. One of the scenarios addresses the danger of carbon monoxide (CO) and smoke leaking into a fire protection section through a damaged fire protection flap. As a result of the dispersion simulation the CO-concentration in all of the rooms is obtained. Together with additional results as deposition and smoke dispersion the outcome of the simulation can be used for training. Among the next steps are the validation of the new models and the selection of critical scenarios. (author)

  17. Product code optimization for determinate state LDPC decoding in robust image transmission.

    Science.gov (United States)

    Thomos, Nikolaos; Boulgouris, Nikolaos V; Strintzis, Michael G

    2006-08-01

    We propose a novel scheme for error-resilient image transmission. The proposed scheme employs a product coder consisting of low-density parity check (LDPC) codes and Reed-Solomon codes in order to deal effectively with bit errors. The efficiency of the proposed scheme is based on the exploitation of determinate symbols in Tanner graph decoding of LDPC codes and a novel product code optimization technique based on error estimation. Experimental evaluation demonstrates the superiority of the proposed system in comparison to recent state-of-the-art techniques for image transmission.

  18. 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)

  19. 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)

  20. FRAP-T1: a computer code for the transient analysis of oxide fuel rods

    International Nuclear Information System (INIS)

    Dearien, J.A.; Miller, R.L.; Hobbins, R.R.; Siefken, L.J.; Baston, V.F.; Coleman, D.R.

    1977-02-01

    FRAP-T is a FORTRAN IV computer code which can be used to solve for the transient response of a light water reactor (LWR) fuel rod during accident transients such as loss-of-coolant accident (LOCA) or a power-cooling-mismatch (PCM). The coupled effects of mechanical, thermal, internal gas, and material property response on the behavior of the fuel rod are considered. FRAP-T is a modular code with each major computational model isolated within the code and coupled to the main code by subroutine calls and data transfer through argument lists. FRAP-T is coupled to a materials properties subcode (MATPRO) which is used to provide gas, fuel, and cladding properties to the FRAP-T computational subcodes. No material properties need be supplied by the code user. The needed water properties are stored in tables built into the code. Critical heat flux (CHF) and heat transfer correlations for a wide range of coolant conditions are contained in modular subroutines. FRAP-T has been evaluated by making extensive comparisons between predictions of the code and experimental data. Comparison of predicted and experimental results are presented for a range of FRAP-T calculated parameters. The code is presently programmed and running on an IBM-360/75 and a CDC 7600 computer

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

    International Nuclear Information System (INIS)

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

    1997-09-01

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

  2. Comparison of energy conservation building codes of Iran, Turkey, Germany, China, ISO 9164 and EN 832

    International Nuclear Information System (INIS)

    Fayaz, Rima; Kari, Behrouz M.

    2009-01-01

    To improve the energy efficiency of buildings via compliance to regulation in Iran, Code No. 19 was devised in 1991. The code lacks high level aims and objectives, addressing the characteristics of Iranian buildings. As a consequence, the code has been revised and is not completely implemented in practice, and still remains inefficient. As with any energy coding system, this code has to identify the right balance between the different energy variables for the Iranian climate and way of life. In order to assist improvements to high level objectives of Code 19, this code is compared with ISO 9164, EN 832, German regulation, TS 825 of Turkey and China's GB 50189 to understand how these have adapted international standards to national features. In order to test the appropriateness of Code 19, five case study buildings in Iran are assessed against Code 19 as well as Turkish standard TS 825 and the results are compared. The results demonstrate that Code 19 is efficient in calculations of building envelope, but it needs improvements in the areas of ventilation, gains from internal and solar sources. The paper concludes by offering suggestions for improving the code.

  3. Verification of the MOTIF code version 3.0

    International Nuclear Information System (INIS)

    Chan, T.; Guvanasen, V.; Nakka, B.W.; Reid, J.A.K.; Scheier, N.W.; Stanchell, F.W.

    1996-12-01

    As part of the Canadian Nuclear Fuel Waste Management Program (CNFWMP), AECL has developed a three-dimensional finite-element code, MOTIF (Model Of Transport In Fractured/ porous media), for detailed modelling of groundwater flow, heat transport and solute transport in a fractured rock mass. The code solves the transient and steady-state equations of groundwater flow, solute (including one-species radionuclide) transport, and heat transport in variably saturated fractured/porous media. The initial development was completed in 1985 (Guvanasen 1985) and version 3.0 was completed in 1986. This version is documented in detail in Guvanasen and Chan (in preparation). This report describes a series of fourteen verification cases which has been used to test the numerical solution techniques and coding of MOTIF, as well as demonstrate some of the MOTIF analysis capabilities. For each case the MOTIF solution has been compared with a corresponding analytical or independently developed alternate numerical solution. Several of the verification cases were included in Level 1 of the International Hydrologic Code Intercomparison Project (HYDROCOIN). The MOTIF results for these cases were also described in the HYDROCOIN Secretariat's compilation and comparison of results submitted by the various project teams (Swedish Nuclear Power Inspectorate 1988). It is evident from the graphical comparisons presented that the MOTIF solutions for the fourteen verification cases are generally in excellent agreement with known analytical or numerical solutions obtained from independent sources. This series of verification studies has established the ability of the MOTIF finite-element code to accurately model the groundwater flow and solute and heat transport phenomena for which it is intended. (author). 20 refs., 14 tabs., 32 figs

  4. Child labor and multinational conduct : a comparison of international business and stakeholder codes

    NARCIS (Netherlands)

    Kolk, A.; van Tulder, R.

    2002-01-01

    Increasing attention to the issue of child labor has been reflected in codes of conduct that emerged in the past decade in particular. This paper examines the way in which multinationals, business associations, governmental and non-governmental organizations deal with child labor in their codes.

  5. Performance Analysis of New Binary User Codes for DS-CDMA Communication

    Science.gov (United States)

    Usha, Kamle; Jaya Sankar, Kottareddygari

    2016-03-01

    This paper analyzes new binary spreading codes through correlation properties and also presents their performance over additive white Gaussian noise (AWGN) channel. The proposed codes are constructed using gray and inverse gray codes. In this paper, a n-bit gray code appended by its n-bit inverse gray code to construct the 2n-length binary user codes are discussed. Like Walsh codes, these binary user codes are available in sizes of power of two and additionally code sets of length 6 and their even multiples are also available. The simple construction technique and generation of code sets of different sizes are the salient features of the proposed codes. Walsh codes and gold codes are considered for comparison in this paper as these are popularly used for synchronous and asynchronous multi user communications respectively. In the current work the auto and cross correlation properties of the proposed codes are compared with those of Walsh codes and gold codes. Performance of the proposed binary user codes for both synchronous and asynchronous direct sequence CDMA communication over AWGN channel is also discussed in this paper. The proposed binary user codes are found to be suitable for both synchronous and asynchronous DS-CDMA communication.

  6. Benchmarking of computer codes and approaches for modeling exposure scenarios

    International Nuclear Information System (INIS)

    Seitz, R.R.; Rittmann, P.D.; Wood, M.I.; Cook, J.R.

    1994-08-01

    The US Department of Energy Headquarters established a performance assessment task team (PATT) to integrate the activities of DOE sites that are preparing performance assessments for the disposal of newly generated low-level waste. The PATT chartered a subteam with the task of comparing computer codes and exposure scenarios used for dose calculations in performance assessments. This report documents the efforts of the subteam. Computer codes considered in the comparison include GENII, PATHRAE-EPA, MICROSHIELD, and ISOSHLD. Calculations were also conducted using spreadsheets to provide a comparison at the most fundamental level. Calculations and modeling approaches are compared for unit radionuclide concentrations in water and soil for the ingestion, inhalation, and external dose pathways. Over 30 tables comparing inputs and results are provided

  7. Intercomparison of liquid metal fast reactor seismic analysis codes. V. 3: Comparison of observed effects with computer simulated effects on reactor cores from seismic disturbances. Proceedings of a final research co-ordination meeting

    International Nuclear Information System (INIS)

    1996-05-01

    This publication contains the final papers summarizing the validation of the codes on the basis of comparison of observed effects with computer simulated effects on reactor cores from seismic disturbances. Refs, figs tabs

  8. Comparison of the nuclear code systems LINEAR-RECENT-NJOY and NJOY

    International Nuclear Information System (INIS)

    Seehusen, J.

    1983-07-01

    The reconstructed cross sections of the code systems LINEAR-RECENT-GROUPIE (Version 1982) and NJOY (Version 1982) have been compared for several materials. Some fuel cycle isotopes and structural materials of the ENDF/B-4 general purpose and ENDF/B-5 dosimetry files have been choosen. The reconstructed total, capture and fission cross sections calculated by LINEAR-RECENT and NJOY have been analized. The two sets of pointwise cross sections differ significantly. Another disagreement was found in the transformation of ENDF/B-4 and 5 files into data with a linear interpolation scheme. Unshielded multigroup constants at O 0 K (620 groups, SANDII) have been calculated by the three code systems LINEAR-RECENT-GROUPIE, NJOY and RESEND5-INTEND. The code system RESEND5-INTEND calculates wrong group constants and should not be used any more. The two sets of group constants obtained from ENDF/B-4 data using GROUPIE and NJOY differ for some group constants by more than 2%. Some disagreements at low energies (10 -3 -eV) of the total cross section of Na-23 and Al-27 are difficult to understand. For ENDF/B-5 dosimetry data the capture group constants differ significantly. (Author) [pt

  9. Accuracy comparison among different machine learning techniques for detecting malicious codes

    Science.gov (United States)

    Narang, Komal

    2016-03-01

    In this paper, a machine learning based model for malware detection is proposed. It can detect newly released malware i.e. zero day attack by analyzing operation codes on Android operating system. The accuracy of Naïve Bayes, Support Vector Machine (SVM) and Neural Network for detecting malicious code has been compared for the proposed model. In the experiment 400 benign files, 100 system files and 500 malicious files have been used to construct the model. The model yields the best accuracy 88.9% when neural network is used as classifier and achieved 95% and 82.8% accuracy for sensitivity and specificity respectively.

  10. Verification study of the FORE-2M nuclear/thermal-hydraulilc analysis computer code

    International Nuclear Information System (INIS)

    Coffield, R.D.; Tang, Y.S.; Markley, R.A.

    1982-01-01

    The verification of the LMFBR core transient performance code, FORE-2M, was performed in two steps. Different components of the computation (individual models) were verified by comparing with analytical solutions and with results obtained from other conventionally accepted computer codes (e.g., TRUMP, LIFE, etc.). For verification of the integral computation method of the code, experimental data in TREAT, SEFOR and natural circulation experiments in EBR-II were compared with the code calculations. Good agreement was obtained for both of these steps. Confirmation of the code verification for undercooling transients is provided by comparisons with the recent FFTF natural circulation experiments. (orig.)

  11. Evaluation of the General Atomic codes TAP and RECA for HTGR accident analyses

    International Nuclear Information System (INIS)

    Ball, S.J.; Cleveland, J.C.; Sanders, J.P.

    1978-01-01

    The General Atomic codes TAP (Transient Analysis Program) and RECA (Reactor Emergency Cooling Analysis) are evaluated with respect to their capability for predicting the dynamic behavior of high-temperature gas-cooled reactors (HTGRs) for postulated accident conditions. Several apparent modeling problems are noted, and the susceptibility of the codes to misuse and input errors is discussed. A critique of code verification plans is also included. The several cases where direct comparisons could be made between TAP/RECA calculations and those based on other independently developed codes indicated generally good agreement, thus contributing to the credibility of the codes

  12. Progress on China nuclear data processing code system

    Science.gov (United States)

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

    2017-09-01

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

  13. CONTAIN code analyses of direct containment heating experiments

    International Nuclear Information System (INIS)

    Williams, D.C.; Griffith, R.O.; Tadios, E.L.; Washington, K.E.

    1995-01-01

    In some nuclear reactor core-melt accidents, a potential exists for molten core-debris to be dispersed into the containment under high pressure. Resulting energy transfer to the containment atmosphere can pressurize the containment. This process, known as direct containment heating (DCH), has been the subject of extensive experimental and analytical programs sponsored by the U.S. Nuclear Regulatory Commission (NRC). The DCH modeling has been an important focus for the development of the CONTAIN code. Results of a detailed independent peer review of the CONTAIN code were published recently. This paper summarizes work performed in support of the peer review in which the CONTAIN code was applied to analyze DCH experiments. Goals of this work were comparison of calculated and experimental results, CONTAIN DCH model assessment, and development of guidance for code users, including development of a standardized input prescription for DCH analysis

  14. Codes of practice and related issues in biomedical waste management

    Energy Technology Data Exchange (ETDEWEB)

    Moy, D.; Watt, C. [Griffith Univ. (Australia)

    1996-12-31

    This paper outlines the development of a National Code of Practice for biomedical waste management in Australia. The 10 key areas addressed by the code are industry mission statement; uniform terms and definitions; community relations - public perceptions and right to know; generation, source separation, and handling; storage requirements; transportation; treatment and disposal; disposal of solid and liquid residues and air emissions; occupational health and safety; staff awareness and education. A comparison with other industry codes in Australia is made. A list of outstanding issues is also provided; these include the development of standard containers, treatment effectiveness, and reusable sharps containers.

  15. Coded aperture optimization using Monte Carlo simulations

    International Nuclear Information System (INIS)

    Martineau, A.; Rocchisani, J.M.; Moretti, J.L.

    2010-01-01

    Coded apertures using Uniformly Redundant Arrays (URA) have been unsuccessfully evaluated for two-dimensional and three-dimensional imaging in Nuclear Medicine. The images reconstructed from coded projections contain artifacts and suffer from poor spatial resolution in the longitudinal direction. We introduce a Maximum-Likelihood Expectation-Maximization (MLEM) algorithm for three-dimensional coded aperture imaging which uses a projection matrix calculated by Monte Carlo simulations. The aim of the algorithm is to reduce artifacts and improve the three-dimensional spatial resolution in the reconstructed images. Firstly, we present the validation of GATE (Geant4 Application for Emission Tomography) for Monte Carlo simulations of a coded mask installed on a clinical gamma camera. The coded mask modelling was validated by comparison between experimental and simulated data in terms of energy spectra, sensitivity and spatial resolution. In the second part of the study, we use the validated model to calculate the projection matrix with Monte Carlo simulations. A three-dimensional thyroid phantom study was performed to compare the performance of the three-dimensional MLEM reconstruction with conventional correlation method. The results indicate that the artifacts are reduced and three-dimensional spatial resolution is improved with the Monte Carlo-based MLEM reconstruction.

  16. Benchmarking of the PHOENIX-P/ANC [Advanced Nodal Code] advanced nuclear design system

    International Nuclear Information System (INIS)

    Nguyen, T.Q.; Liu, Y.S.; Durston, C.; Casadei, A.L.

    1988-01-01

    At Westinghouse, an advanced neutronic methods program was designed to improve the quality of the predictions, enhance flexibility in designing advanced fuel and related products, and improve design lead time. Extensive benchmarking data is presented to demonstrate the accuracy of the Advanced Nodal Code (ANC) and the PHOENIX-P advanced lattice code. Qualification data to demonstrate the accuracy of ANC include comparison of key physics parameters against a fine-mesh diffusion theory code, TORTISE. Benchmarking data to demonstrate the validity of the PHOENIX-P methodologies include comparison of physics predictions against critical experiments, isotopics measurements and measured power distributions from spatial criticals. The accuracy of the PHOENIX-P/ANC Advanced Design System is demonstrated by comparing predictions of hot zero power physics parameters and hot full power core follow against measured data from operating reactors. The excellent performance of this system for a broad range of comparisons establishes the basis for implementation of these tools for core design, licensing and operational follow of PWR [pressurized water reactor] cores at Westinghouse

  17. Application of the TRAC-PD2 code to the simulation of the CANON experiment

    International Nuclear Information System (INIS)

    Neves Conti, T. das; Freitas, R.L.

    1985-01-01

    A comparison between the TRAC -PD2 code calculations and results from the CANON experiment is presented. The CANON experiment simulates the loss of coolant accident through the depressurization of a horizontal tube containing water at different temperatures. The experiment consist of the instantaneous rupture at one end of the tubing and the corresponding pressure and void fraction measurements during the transient. The comparison shows that the TRAC-PD2 code predicts satisfactorily the pressure and void fraction evolution in the CANON experiment. (F.C.) [pt

  18. The PARET code and the analysis of the SPERT I transients

    Energy Technology Data Exchange (ETDEWEB)

    Woodruff, William L [Argonne National Laboratory, Argonne (United States)

    1983-09-01

    The PARET code has been adapted for the testing of methods and models and for subsequent use in the analysis of transient behavior in research reactors. Comparisons with the experimental results from the SPERT-I transients are provided. The code has also been applied to the analysis of the IAEA 10 MW benchmark cores for protected and unprotected transients. The PARET code was originally developed for the analysis of the SPERT-III experiments for temperatures and pressures typical of power reactors. This code has now been modified to include a selection of flow instability, departure from nucleate boiling (DNB), single and two-phase heat transfer correlations, and a properties library considered more applicable to the low pressures, temperatures, and flow rates encountered in research reactors. The PARET code provides a coupled thermal, hydraulic, and point kinetics capability with continuous reactivity feedback, and an optional voiding model which estimates the voiding produced by subcooled boiling. The present version of the PARET code provides a convenient means of assessing the various models and correlations proposed for use in the analysis of research reactor behavior. For comparison with experiments the SPERT-I cores B-24/32, B-12/64, and D-12/25 were chosen. The B-24/32 core is similar in design to many plate type research reactors in current operation, and the D-12/25 core is of interest because the test included both nondestructive and destructive transients.

  19. The PARET code and the analysis of the SPERT I transients

    International Nuclear Information System (INIS)

    Woodruff, William L.

    1983-01-01

    The PARET code has been adapted for the testing of methods and models and for subsequent use in the analysis of transient behavior in research reactors. Comparisons with the experimental results from the SPERT-I transients are provided. The code has also been applied to the analysis of the IAEA 10 MW benchmark cores for protected and unprotected transients. The PARET code was originally developed for the analysis of the SPERT-III experiments for temperatures and pressures typical of power reactors. This code has now been modified to include a selection of flow instability, departure from nucleate boiling (DNB), single and two-phase heat transfer correlations, and a properties library considered more applicable to the low pressures, temperatures, and flow rates encountered in research reactors. The PARET code provides a coupled thermal, hydraulic, and point kinetics capability with continuous reactivity feedback, and an optional voiding model which estimates the voiding produced by subcooled boiling. The present version of the PARET code provides a convenient means of assessing the various models and correlations proposed for use in the analysis of research reactor behavior. For comparison with experiments the SPERT-I cores B-24/32, B-12/64, and D-12/25 were chosen. The B-24/32 core is similar in design to many plate type research reactors in current operation, and the D-12/25 core is of interest because the test included both nondestructive and destructive transients

  20. Preliminary comparison of the conventional and quasi-snowflake divertor configurations with the 2D code EDGE2D/EIRENE in the FAST tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Viola, B.; Maddaluno, G.; Pericoli Ridolfini, V. [EURATOM-ENEA Association, C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Rome) (Italy); Corrigan, G.; Harting, D. [Culham Centre of Fusion Energy, EURATOM-Association, Abingdon (United Kingdom); Mattia, M. [Dipartimento di Informatica, Sistemi e Produzione, Universita di Roma, Tor Vergata, Via del Politecnico, 00133 Roma (Italy); Zagorski, R. [Institute of Plasma Physics and Laser Microfusion-EURATOM Association, 01-497 Warsaw (Poland)

    2014-06-15

    The new magnetic configurations for tokamak divertors, snowflake and super-X, proposed to mitigate the problem of the power exhaust in reactors have clearly evidenced the need for an accurate and reliable modeling of the physics governing the interaction with the plates. The initial effort undertaken jointly by ENEA and IPPLM has been focused to exploit a simple and versatile modeling tool, namely the 2D TECXY code, to obtain preliminary comparison between the conventional and snowflake configurations for the proposed new device FAST that should realize an edge plasma with properties quite close to those of a reactor. The very interesting features found for the snowflake, namely a power load mitigation much larger than expected directly from the change of the magnetic topology, has further pushed us to check these results with the more sophisticated computational tool EDGE2D coupled with the neutral code module EIRENE. After a preparatory work that has been carried out in order to adapt this code combination to deal with non-conventional, single null equilibria and in particular with second order nulls in the poloidal field generated in the snowflake configuration, in this paper we describe the first activity to compare these codes and discuss the first results obtained for FAST. The outcome of these EDGE2D runs is in qualitative agreement with those of TECXY, confirming the potential benefit obtainable from a snowflake configuration. (copyright 2014 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  1. [Coding Causes of Death with IRIS Software. Impact in Navarre Mortality Statistic].

    Science.gov (United States)

    Floristán Floristán, Yugo; Delfrade Osinaga, Josu; Carrillo Prieto, Jesus; Aguirre Perez, Jesus; Moreno-Iribas, Conchi

    2016-08-02

    There are few studies that analyze changes in mortality statistics derived from the use of IRIS software, an automatic system for coding multiple causes of death and for the selection of the underlying cause of death, compared to manual coding. This study evaluated the impact of the use of IRIS in the Navarre mortality statistic. We proceeded to double coding 5,060 death certificates corresponding to residents in Navarra in 2014. We calculated coincidence between the two encodings for ICD10 chapters and for the list of causes of the Spanish National Statistics Institute (INE-102) and we estimated the change on mortality rates. IRIS automatically coded 90% of death certificates. The coincidence to 4 characters and in the same chapter of the CIE10 was 79.1% and 92.0%, respectively. Furthermore, coincidence with the short INE-102 list was 88.3%. Higher matches were found in death certificate of people under 65 years. In comparison with manual coding there was an increase in deaths from endocrine diseases (31%), mental disorders (19%) and disease of nervous system (9%), while a decrease of genitourinary system diseases was observed (21%). The coincidence at level of ICD10 chapters coding by IRIS in comparison to manual coding was 9 out of 10 deaths, similar to what is observed in other studies. The implementation of IRIS has led to increased of endocrine diseases, especially diabetes and hyperlipidaemia, and mental disorders, especially dementias.

  2. An approach to verification and validation of MHD codes for fusion applications

    Energy Technology Data Exchange (ETDEWEB)

    Smolentsev, S., E-mail: sergey@fusion.ucla.edu [University of California, Los Angeles (United States); Badia, S. [Centre Internacional de Mètodes Numèrics en Enginyeria, Barcelona (Spain); Universitat Politècnica de Catalunya – Barcelona Tech (Spain); Bhattacharyay, R. [Institute for Plasma Research, Gandhinagar, Gujarat (India); Bühler, L. [Karlsruhe Institute of Technology (Germany); Chen, L. [University of Chinese Academy of Sciences, Beijing (China); Huang, Q. [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui (China); Jin, H.-G. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Krasnov, D. [Technische Universität Ilmenau (Germany); Lee, D.-W. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Mas de les Valls, E. [Centre Internacional de Mètodes Numèrics en Enginyeria, Barcelona (Spain); Universitat Politècnica de Catalunya – Barcelona Tech (Spain); Mistrangelo, C. [Karlsruhe Institute of Technology (Germany); Munipalli, R. [HyPerComp, Westlake Village (United States); Ni, M.-J. [University of Chinese Academy of Sciences, Beijing (China); Pashkevich, D. [St. Petersburg State Polytechnical University (Russian Federation); Patel, A. [Universitat Politècnica de Catalunya – Barcelona Tech (Spain); Pulugundla, G. [University of California, Los Angeles (United States); Satyamurthy, P. [Bhabha Atomic Research Center (India); Snegirev, A. [St. Petersburg State Polytechnical University (Russian Federation); Sviridov, V. [Moscow Power Engineering Institute (Russian Federation); Swain, P. [Bhabha Atomic Research Center (India); and others

    2015-11-15

    Highlights: • Review of status of MHD codes for fusion applications. • Selection of five benchmark problems. • Guidance for verification and validation of MHD codes for fusion applications. - Abstract: We propose a new activity on verification and validation (V&V) of MHD codes presently employed by the fusion community as a predictive capability tool for liquid metal cooling applications, such as liquid metal blankets. The important steps in the development of MHD codes starting from the 1970s are outlined first and then basic MHD codes, which are currently in use by designers of liquid breeder blankets, are reviewed. A benchmark database of five problems has been proposed to cover a wide range of MHD flows from laminar fully developed to turbulent flows, which are of interest for fusion applications: (A) 2D fully developed laminar steady MHD flow, (B) 3D laminar, steady developing MHD flow in a non-uniform magnetic field, (C) quasi-two-dimensional MHD turbulent flow, (D) 3D turbulent MHD flow, and (E) MHD flow with heat transfer (buoyant convection). Finally, we introduce important details of the proposed activities, such as basic V&V rules and schedule. The main goal of the present paper is to help in establishing an efficient V&V framework and to initiate benchmarking among interested parties. The comparison results computed by the codes against analytical solutions and trusted experimental and numerical data as well as code-to-code comparisons will be presented and analyzed in companion paper/papers.

  3. Validation of a new continuous Monte Carlo burnup code using a Mox fuel assembly

    International Nuclear Information System (INIS)

    El bakkari, B.; El Bardouni, T.; Merroun, O.; El Younoussi, C.; Boulaich, Y.; Boukhal, H.; Chakir, E.

    2009-01-01

    The reactivity of nuclear fuel decreases with irradiation (or burnup) due to the transformation of heavy nuclides and the formation of fission products. Burnup credit studies aim at accounting for fuel irradiation in criticality studies of the nuclear fuel cycle (transport, storage, etc...). The principal objective of this study is to evaluate the potential capabilities of a newly developed burnup code called 'BUCAL1'. BUCAL1 differs in comparison with other burnup codes as it does not use the calculated neutron flux as input to other computer codes to generate the nuclide inventory for the next time step. Instead, BUCAL1 directly uses the neutron reaction tally information generated by MCNP for each nuclide of interest to determine the new nuclides inventory. This allows the full capabilities of MCNP to be incorporated into the calculation and a more accurate and robust analysis to be performed. Validation of BUCAL1 was processed by code-to-code comparisons using predictions of several codes from the NEA/OCED. Infinite multiplication factors (k ∞ ) and important fission product and actinide concentrations were compared for a MOX core benchmark exercise. Results of calculations are analysed and discussed.

  4. Double-digit coding of examination math problems

    Directory of Open Access Journals (Sweden)

    Agnieszka Sułowska

    2013-09-01

    Full Text Available Various methods are used worldwide to evaluate student solutions to examination tasks. Usually the results simply provide information about student competency and after aggregation, are also used as a tool of making comparisons between schools. In particular, the standard evaluation methods do not allow conclusions to be drawn about possible improvements of teaching methods. There are however, task assessment methods which not only allow description of student achievement, but also possible causes of failure. One such method, which can be applied to extended response tasks, is double-digit coding which has been used in some international educational research. This paper presents the first Polish experiences of applying this method to examination tasks in mathematics, using a special coding key to carry out the evaluation. Lessons learned during the coding key construction and its application in the assessment process are described.

  5. QR code optical encryption using spatially incoherent illumination

    Science.gov (United States)

    Cheremkhin, P. A.; Krasnov, V. V.; Rodin, V. G.; Starikov, R. S.

    2017-02-01

    Optical encryption is an actively developing field of science. The majority of encryption techniques use coherent illumination and suffer from speckle noise, which severely limits their applicability. The spatially incoherent encryption technique does not have this drawback, but its effectiveness is dependent on the Fourier spectrum properties of the image to be encrypted. The application of a quick response (QR) code in the capacity of a data container solves this problem, and the embedded error correction code also enables errorless decryption. The optical encryption of digital information in the form of QR codes using spatially incoherent illumination was implemented experimentally. The encryption is based on the optical convolution of the image to be encrypted with the kinoform point spread function, which serves as an encryption key. Two liquid crystal spatial light modulators were used in the experimental setup for the QR code and the kinoform imaging, respectively. The quality of the encryption and decryption was analyzed in relation to the QR code size. Decryption was conducted digitally. The successful decryption of encrypted QR codes of up to 129  ×  129 pixels was demonstrated. A comparison with the coherent QR code encryption technique showed that the proposed technique has a signal-to-noise ratio that is at least two times higher.

  6. Comment on ‘egs_brachy: a versatile and fast Monte Carlo code for brachytherapy’

    Science.gov (United States)

    Yegin, Gultekin

    2018-02-01

    In a recent paper (Chamberland et al 2016 Phys. Med. Biol. 61 8214) develop a new Monte Carlo code called egs_brachy for brachytherapy treatments. It is based on EGSnrc, and written in the C++ programming language. In order to benchmark the egs_brachy code, the authors use it in various test case scenarios in which complex geometry conditions exist. Another EGSnrc based brachytherapy dose calculation engine, BrachyDose, is used for dose comparisons. The authors fail to prove that egs_brachy can produce reasonable dose values for brachytherapy sources in a given medium. The dose comparisons in the paper are erroneous and misleading. egs_brachy should not be used in any further research studies unless and until all the potential bugs are fixed in the code.

  7. RISKIND verification and benchmark comparisons

    Energy Technology Data Exchange (ETDEWEB)

    Biwer, B.M.; Arnish, J.J.; Chen, S.Y.; Kamboj, S.

    1997-08-01

    This report presents verification calculations and benchmark comparisons for RISKIND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the population from exposures associated with the transportation of spent nuclear fuel and other radioactive materials. Spreadsheet calculations were performed to verify the proper operation of the major options and calculational steps in RISKIND. The program is unique in that it combines a variety of well-established models into a comprehensive treatment for assessing risks from the transportation of radioactive materials. Benchmark comparisons with other validated codes that incorporate similar models were also performed. For instance, the external gamma and neutron dose rate curves for a shipping package estimated by RISKIND were compared with those estimated by using the RADTRAN 4 code and NUREG-0170 methodology. Atmospheric dispersion of released material and dose estimates from the GENII and CAP88-PC codes. Verification results have shown the program to be performing its intended function correctly. The benchmark results indicate that the predictions made by RISKIND are within acceptable limits when compared with predictions from similar existing models.

  8. RISKIND verification and benchmark comparisons

    International Nuclear Information System (INIS)

    Biwer, B.M.; Arnish, J.J.; Chen, S.Y.; Kamboj, S.

    1997-08-01

    This report presents verification calculations and benchmark comparisons for RISKIND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the population from exposures associated with the transportation of spent nuclear fuel and other radioactive materials. Spreadsheet calculations were performed to verify the proper operation of the major options and calculational steps in RISKIND. The program is unique in that it combines a variety of well-established models into a comprehensive treatment for assessing risks from the transportation of radioactive materials. Benchmark comparisons with other validated codes that incorporate similar models were also performed. For instance, the external gamma and neutron dose rate curves for a shipping package estimated by RISKIND were compared with those estimated by using the RADTRAN 4 code and NUREG-0170 methodology. Atmospheric dispersion of released material and dose estimates from the GENII and CAP88-PC codes. Verification results have shown the program to be performing its intended function correctly. The benchmark results indicate that the predictions made by RISKIND are within acceptable limits when compared with predictions from similar existing models

  9. Comparison of calculations of a reflected reactor with diffusion, SN and Monte Carlo codes

    International Nuclear Information System (INIS)

    McGregor, B.

    1975-01-01

    A diffusion theory code, POW, was compared with a Monte Carlo transport theory code, KENO, for the calculation of a small C/ 235 U cylindrical core with a graphite reflector. The calculated multiplication factors were in good agreement but differences were noted in region-averaged group fluxes. A one-dimensional spherical geometry was devised to approximate cylindrical geometry. Differences similar to those already observed were noted when the region-averaged fluxes from a diffusion theory (POW) calculation were compared with an SN transport theory (ANAUSN) calculation for the spherical model. Calculations made with SN and Monte Carlo transport codes were in good agreement. It was concluded that observed flux differences were attributable to the POW code, and were not inconsistent with inherent diffusion theory approximations. (author)

  10. The nonelastic reaction code BRIEFF and its intranuclear cascade BRIC

    International Nuclear Information System (INIS)

    Duarte, H.

    2008-01-01

    The intranuclear cascade (INC) code of Bruyeres-le-Chatel named BRIC is the first part of the nonelastic reaction code BRIEFF. Recent changes in our INC are presented. They were done for nucleon induced reaction to improve results below 100 MeV, and to calculate cross sections of compound nucleus formation. These cross sections are used in the evaporation component of our reaction code. BRIEFF is included in the Bruyeres-le-Chatel version of HETC (High Energy Transport Code) to perform thick targets calculations. BRIC has recently been incorporated into MCNPX 2.4.0 to verify thick target results of neutron yields and to do some comparisons with other nuclear models or libraries. Good agreement with data is obtained on average. (author)

  11. A point kernel shielding code, PKN-HP, for high energy proton incident

    Energy Technology Data Exchange (ETDEWEB)

    Kotegawa, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1996-06-01

    A point kernel integral technique code PKN-HP, and the related thick target neutron yield data have been developed to calculate neutron and secondary gamma-ray dose equivalents in ordinary concrete and iron shields for fully stopping length C, Cu and U-238 target neutrons produced by 100 MeV-10 GeV proton incident in a 3-dimensional geometry. The comparisons among calculation results of the present code and other calculation techniques, and measured values showed the usefulness of the code. (author)

  12. Dynamic benchmarking of simulation codes

    International Nuclear Information System (INIS)

    Henry, R.E.; Paik, C.Y.; Hauser, G.M.

    1996-01-01

    output includes a plot of the MAAP calculation and the plant data. For the large integral experiments, a major part, but not all of the MAAP code is needed. These use an experiment specific benchmark routine that includes all of the information and boundary conditions for performing the calculation, as well as the information of which parts of MAAP are unnecessary and can be 'bypassed'. Lastly, the separate effects tests only require a few MAAP routines. These are exercised through their own specific benchmark routine that includes the experiment specific information and boundary conditions. This benchmark routine calls the appropriate MAAP routines from the source code, performs the calculations, including integration where necessary and provide the comparison between the MAAP calculation and the experimental observations. (author)

  13. Comparison study of inelastic analysis codes for high temperature structure

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Bum; Lee, H. Y.; Park, C. K.; Geon, G. P.; Lee, J. H

    2004-02-01

    LMR high temperature structures subjected to operating and transient loadings may exhibit very complex deformation behaviors due to the use of ductile material such as 316SS and the systematic analysis technology of high temperature structure for reliable safety assessment is essential. In this project, comparative study with developed inelastic analysis program NONSTA and the existing analysis codes was performed applying various types of loading including non-proportional loading. The performance of NONSTA was confirmed and the effect of inelastic constants on the analysis result was analyzed. Also, the applicability of the inelastic analysis was enlarged as a result of applying both the developed program and the existing codes to the analyses of the enhanced creep behavior and the elastic follow-up behavior of high temperature structures and the necessary items for improvements were deduced. Further studies on the improvement of NONSTA program and the decision of the proper values of inelastic constants are necessary.

  14. Comparison of thick-target (alpha,n yield calculation codes

    Directory of Open Access Journals (Sweden)

    Fernandes Ana C.

    2017-01-01

    Full Text Available Neutron production yields and energy distributions from (α,n reactions in light elements were calculated using three different codes (SOURCES, NEDIS and USD and compared with the existing experimental data in the 3.5-10 MeV alpha energy range. SOURCES and NEDIS display an agreement between calculated and measured yields in the decay series of 235U, 238U and 232Th within ±10% for most materials. The discrepancy increases with alpha energy but still an agreement of ±20% applies to all materials with reliable elemental production yields (the few exceptions are identified. The calculated neutron energy distributions describe the experimental data, with NEDIS retrieving very well the detailed features. USD generally underestimates the measured yields, in particular for compounds with heavy elements and/or at high alpha energies. The energy distributions exhibit sharp peaks that do not match the observations. These findings may be caused by a poor accounting of the alpha particle energy loss by the code. A big variability was found among the calculated neutron production yields for alphas from Sm decay; the lack of yield measurements for low (~2 MeV alphas does not allow to conclude on the codes’ accuracy in this energy region.

  15. Dosskin code for radiological evaluation of skin radioactive contaminations

    International Nuclear Information System (INIS)

    Cornejo D, N.

    1996-01-01

    The conceptual procedure and computational features of the DOSSKIN code are shown. This code calculates, in a very interactive way, skin equivalent doses and radiological risk related to skin radioactive contaminations. The evaluation takes into account the contributions of contaminant daughter nuclides and backscattering of beta particles in any skin cover. DOSSKIN also allows to estimate the maximum time needed to decontaminate the affected zone, using, as input quantity, the limit value of skin equivalent dose considered by users. The comparison of the results obtained by the DOSSKIN code with those reported by different authors are showed. The differences of results are less than 30%. (authors). 4 refs., 3 fig., 1 tab

  16. Radiant Energy Measurements from a Scaled Jet Engine Axisymmetric Exhaust Nozzle for a Baseline Code Validation Case

    Science.gov (United States)

    Baumeister, Joseph F.

    1994-01-01

    A non-flowing, electrically heated test rig was developed to verify computer codes that calculate radiant energy propagation from nozzle geometries that represent aircraft propulsion nozzle systems. Since there are a variety of analysis tools used to evaluate thermal radiation propagation from partially enclosed nozzle surfaces, an experimental benchmark test case was developed for code comparison. This paper briefly describes the nozzle test rig and the developed analytical nozzle geometry used to compare the experimental and predicted thermal radiation results. A major objective of this effort was to make available the experimental results and the analytical model in a format to facilitate conversion to existing computer code formats. For code validation purposes this nozzle geometry represents one validation case for one set of analysis conditions. Since each computer code has advantages and disadvantages based on scope, requirements, and desired accuracy, the usefulness of this single nozzle baseline validation case can be limited for some code comparisons.

  17. A post-processor for the PEST code

    International Nuclear Information System (INIS)

    Priesche, S.; Manickam, J.; Johnson, J.L.

    1992-01-01

    A new post-processor has been developed for use with output from the PEST tokamak stability code. It allows us to use quantities calculated by PEST and take better advantage of the physical picture of the plasma instability which they can provide. This will improve comparison with experimentally measured quantities as well as facilitate understanding of theoretical studies

  18. Development of LWR fuel performance code FEMAXI-6

    International Nuclear Information System (INIS)

    Suzuki, Motoe

    2006-01-01

    LWR fuel performance code: FEMAXI-6 (Finite Element Method in AXIs-symmetric system) is a representative fuel analysis code in Japan. Development history, background, design idea, features of model, and future are stated. Characteristic performance of LWR fuel and analysis code, what is model, development history of FEMAXI, use of FEMAXI code, fuel model, and a special feature of FEMAXI model is described. As examples of analysis, PCMI (Pellet-Clad Mechanical Interaction), fission gas release, gap bonding, and fission gas bubble swelling are reported. Thermal analysis and dynamic analysis system of FEMAXI-6, function block at one time step of FEMAXI-6, analytical example of PCMI in the output increase test by FEMAXI-III, analysis of fission gas release in Halden reactor by FEMAXI-V, comparison of the center temperature of fuel in Halden reactor, and analysis of change of diameter of fuel rod in high burn up BWR fuel are shown. (S.Y.)

  19. A DRAGON-MCNP comparison of void reactivity calculations

    Energy Technology Data Exchange (ETDEWEB)

    Marleau, G [Ecole Polytechnique, Montreal, PQ (Canada). Inst. de Genie Nucleaire; Milgram, M S [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

    1996-12-31

    The determination of the reactivity coefficients associated with coolant voiding in a CANDU reactor is a subject which has attracted a large amount of interest in the last few years both from the theoretical and experimental point of view. One expects that deterministic codes such as DRAGON and WIMS-AECL or the MCNP4 Monte Carlo code should be able to adequately simulate the cell behaviour upon coolant voiding. However, the absence of an experimental database at equilibrium and discharge burnups has not permitted the full validation of any of these lattice codes, although a partial validation through comparison of two different computer codes has been considered. Here we present a comparison between DRAGON and MCNP4 of the void reactivity evaluation for fresh fuel. (author). 16 refs., 5 tabs.

  20. A DRAGON-MCNP comparison of void reactivity calculations

    International Nuclear Information System (INIS)

    Marleau, G.

    1995-01-01

    The determination of the reactivity coefficients associated with coolant voiding in a CANDU reactor is a subject which has attracted a large amount of interest in the last few years both from the theoretical and experimental point of view. One expects that deterministic codes such as DRAGON and WIMS-AECL or the MCNP4 Monte Carlo code should be able to adequately simulate the cell behaviour upon coolant voiding. However, the absence of an experimental database at equilibrium and discharge burnups has not permitted the full validation of any of these lattice codes, although a partial validation through comparison of two different computer codes has been considered. Here we present a comparison between DRAGON and MCNP4 of the void reactivity evaluation for fresh fuel. (author). 16 refs., 5 tabs

  1. Benchmarking the Multidimensional Stellar Implicit Code MUSIC

    Science.gov (United States)

    Goffrey, T.; Pratt, J.; Viallet, M.; Baraffe, I.; Popov, M. V.; Walder, R.; Folini, D.; Geroux, C.; Constantino, T.

    2017-04-01

    We present the results of a numerical benchmark study for the MUltidimensional Stellar Implicit Code (MUSIC) based on widely applicable two- and three-dimensional compressible hydrodynamics problems relevant to stellar interiors. MUSIC is an implicit large eddy simulation code that uses implicit time integration, implemented as a Jacobian-free Newton Krylov method. A physics based preconditioning technique which can be adjusted to target varying physics is used to improve the performance of the solver. The problems used for this benchmark study include the Rayleigh-Taylor and Kelvin-Helmholtz instabilities, and the decay of the Taylor-Green vortex. Additionally we show a test of hydrostatic equilibrium, in a stellar environment which is dominated by radiative effects. In this setting the flexibility of the preconditioning technique is demonstrated. This work aims to bridge the gap between the hydrodynamic test problems typically used during development of numerical methods and the complex flows of stellar interiors. A series of multidimensional tests were performed and analysed. Each of these test cases was analysed with a simple, scalar diagnostic, with the aim of enabling direct code comparisons. As the tests performed do not have analytic solutions, we verify MUSIC by comparing it to established codes including ATHENA and the PENCIL code. MUSIC is able to both reproduce behaviour from established and widely-used codes as well as results expected from theoretical predictions. This benchmarking study concludes a series of papers describing the development of the MUSIC code and provides confidence in future applications.

  2. A computer code for Cohort Analysis of Increased Risks of Death (CAIRD). Technical report

    International Nuclear Information System (INIS)

    Cook, J.R.; Bunger, B.M.; Barrick, M.K.

    1978-06-01

    The most serious health risk confronting individuals exposed to radiation is death from an induced cancer. Since cancers usually do no develop until many years after exposure, other causes of death may intervene and take the lives of those destined to die from cancer. This computer code has been developed to aid risk analysis by calculating the number of premature deaths and loss of years of life produced by a hypothetical population after exposure to a given risk situation. The code generates modified life tables and estimates the impact of increased risk through several numerical comparisons with the appropriate reference life tables. One of the code's frequent applications is in estimating the number of radiation induced deaths that would result from exposing an initial population of 100,000 individuals to an annual radiation dose. For each risk situation analyzed, the computer code generates a summary table which documents the input, data and contains the results of the comparisons with reference life tables

  3. Development of thermal hydraulic evaluation code for CANDU reactors

    International Nuclear Information System (INIS)

    Kim, Man Woong; Yu, Seon Oh; Choi, Yong Seog; Shin, Chull; Hwang, Soo Hyun

    2004-02-01

    To enhance the safety of operating CANDU reactors, the establishment of the safety analysis codes system for CANDU reactors is in progress. As for the development of thermal-hydraulic analysis code for CANDU system, the studies for improvement of evaluation model inside RELAP/CANDU code and the development of safety assessment methodology for GSI (Generic Safety Issues) are in progress as a part of establishment of CANDU safety assessment system. To develop the 3-D thermal-hydraulic analysis code for moderator system, the CFD models for analyzing the CANDU-6 moderator circulation are developed. One model uses a structured grid system with the porous media approach for the 380 Calandria tubes in the core region. The other uses a unstructured grid system on the real geometry of 380 Calandria tubes, so that the detailed fluid flow between the Calandria tubes can be observed. As to the development of thermal-hydraulic analysis code for containment, the study on the applicability of CONTAIN 2.0 code to a CANDU containment was conducted and a simulation of the thermal-hydraulic phenomena during the accident was performed. Besides, the model comparison of ESFs (Engineered Safety Features) inside CONTAIN 2.0 code and PRESCON code has also conducted

  4. OC3—Benchmark Exercise of Aero-elastic Offshore Wind Turbine Codes

    Science.gov (United States)

    Passon, P.; Kühn, M.; Butterfield, S.; Jonkman, J.; Camp, T.; Larsen, T. J.

    2007-07-01

    This paper introduces the work content and status of the first international investigation and verification of aero-elastic codes for offshore wind turbines as performed by the "Offshore Code Comparison Collaboration"(OC3) within the "IEA Wind Annex XXIII - Subtask 2". An overview is given on the state-of-the-art of the concerned offshore wind turbine simulation codes. Exemplary results of benchmark simulations from the first phase of the project are presented and discussed while subsequent phases are introduced. Furthermore, the paper discusses areas where differences between the codes have been identified and the sources of those differences, such as the differing theories implemented into the individual codes. Finally, further research and code development needs are presented based on the latest findings from the current state of the project.

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

    Science.gov (United States)

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

    2006-10-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  7. Development of three-dimensional transport code by the double finite element method

    International Nuclear Information System (INIS)

    Fujimura, Toichiro

    1985-01-01

    Development of a three-dimensional neutron transport code by the double finite element method is described. Both of the Galerkin and variational methods are adopted to solve the problem, and then the characteristics of them are compared. Computational results of the collocation method, developed as a technique for the vaviational one, are illustrated in comparison with those of an Ssub(n) code. (author)

  8. SPIDERMAN: Fast code to simulate secondary transits and phase curves

    Science.gov (United States)

    Louden, Tom; Kreidberg, Laura

    2017-11-01

    SPIDERMAN calculates exoplanet phase curves and secondary eclipses with arbitrary surface brightness distributions in two dimensions. The code uses a geometrical algorithm to solve exactly the area of sections of the disc of the planet that are occulted by the star. Approximately 1000 models can be generated per second in typical use, which makes making Markov Chain Monte Carlo analyses practicable. The code is modular and allows comparison of the effect of multiple different brightness distributions for a dataset.

  9. Application of containment codes to LMFBRs in the United States

    International Nuclear Information System (INIS)

    Chang, Y.W.

    1977-01-01

    This paper describes the application of containment codes to predict the response of the fast reactor containment and the primary piping loops to HCDAs. Five sample problems are given to illustrate their applications. The first problem deals with the response of the primary containment to an HCDA. The second problem deals with the coolant flow in the reactor lower plenum. The third problem concerns sodium spillage and slug impact. The fourth problem deals with the response of a piping loop. The fifth problem analyzes the response of a reactor head closure. Application of codes in parametric studies and comparison of code predictions with experiments are also discussed. (Auth.)

  10. Real-time minimal-bit-error probability decoding of convolutional codes

    Science.gov (United States)

    Lee, L.-N.

    1974-01-01

    A recursive procedure is derived for decoding of rate R = 1/n binary convolutional codes which minimizes the probability of the individual decoding decisions for each information bit, subject to the constraint that the decoding delay be limited to Delta branches. This new decoding algorithm is similar to, but somewhat more complex than, the Viterbi decoding algorithm. A real-time, i.e., fixed decoding delay, version of the Viterbi algorithm is also developed and used for comparison to the new algorithm on simulated channels. It is shown that the new algorithm offers advantages over Viterbi decoding in soft-decision applications, such as in the inner coding system for concatenated coding.

  11. Real-time minimal bit error probability decoding of convolutional codes

    Science.gov (United States)

    Lee, L. N.

    1973-01-01

    A recursive procedure is derived for decoding of rate R=1/n binary convolutional codes which minimizes the probability of the individual decoding decisions for each information bit subject to the constraint that the decoding delay be limited to Delta branches. This new decoding algorithm is similar to, but somewhat more complex than, the Viterbi decoding algorithm. A real-time, i.e. fixed decoding delay, version of the Viterbi algorithm is also developed and used for comparison to the new algorithm on simulated channels. It is shown that the new algorithm offers advantages over Viterbi decoding in soft-decision applications such as in the inner coding system for concatenated coding.

  12. Central Decoding for Multiple Description Codes based on Domain Partitioning

    Directory of Open Access Journals (Sweden)

    M. Spiertz

    2006-01-01

    Full Text Available Multiple Description Codes (MDC can be used to trade redundancy against packet loss resistance for transmitting data over lossy diversity networks. In this work we focus on MD transform coding based on domain partitioning. Compared to Vaishampayan’s quantizer based MDC, domain based MD coding is a simple approach for generating different descriptions, by using different quantizers for each description. Commonly, only the highest rate quantizer is used for reconstruction. In this paper we investigate the benefit of using the lower rate quantizers to enhance the reconstruction quality at decoder side. The comparison is done on artificial source data and on image data. 

  13. Application of containment codes to LMFBRs in the United States

    International Nuclear Information System (INIS)

    Chang, Y.W.

    1977-01-01

    The application of containment codes to predict the response of the fast reactor containment and the primary piping loops to HCDAs is described. Five sample problems are given to illustrate their applications. The first problem deals with the response of the primary containment to an HCDA. The second problem deals with the coolant flow in the reactor lower plenum. The third proem concerns sodium spillage and slug impact. The fourth problem deals with the response of a piping loop. The fifth problem analyzes the response of a reactor head closure. Application of codes in parametric studies and comparison of code predictions with experiments are also discussed

  14. Whist code calculations of ignition margin in an ignition tokamak

    International Nuclear Information System (INIS)

    Sheffield, J.

    1985-01-01

    A simple global model was developed to determine the ignition margin of tokamaks including electron and ion conduction losses. A comparison of this model with results from a 1 1/2 dimensional Whist code is made

  15. An RNA-Seq strategy to detect the complete coding and non-coding transcriptome including full-length imprinted macro ncRNAs.

    Directory of Open Access Journals (Sweden)

    Ru Huang

    Full Text Available Imprinted macro non-protein-coding (nc RNAs are cis-repressor transcripts that silence multiple genes in at least three imprinted gene clusters in the mouse genome. Similar macro or long ncRNAs are abundant in the mammalian genome. Here we present the full coding and non-coding transcriptome of two mouse tissues: differentiated ES cells and fetal head using an optimized RNA-Seq strategy. The data produced is highly reproducible in different sequencing locations and is able to detect the full length of imprinted macro ncRNAs such as Airn and Kcnq1ot1, whose length ranges between 80-118 kb. Transcripts show a more uniform read coverage when RNA is fragmented with RNA hydrolysis compared with cDNA fragmentation by shearing. Irrespective of the fragmentation method, all coding and non-coding transcripts longer than 8 kb show a gradual loss of sequencing tags towards the 3' end. Comparisons to published RNA-Seq datasets show that the strategy presented here is more efficient in detecting known functional imprinted macro ncRNAs and also indicate that standardization of RNA preparation protocols would increase the comparability of the transcriptome between different RNA-Seq datasets.

  16. AECL international standard problem ISP-41 FU/1 follow-up exercise (Phase 1): Containment Iodine Computer Code Exercise: Parametric Studies

    International Nuclear Information System (INIS)

    Ball, J.; Glowa, G.; Wren, J.; Ewig, F.; Dickenson, S.; Billarand, Y.; Cantrel, L.; Rydl, A.; Royen, J.

    2001-06-01

    This report describes the results of the second phase of International Standard Problem (ISP) 41, an iodine behaviour code comparison exercise. The first phase of the study, which was based on a simple Radioiodine Test Facility (RTF) experiment, demonstrated that all of the iodine behaviour codes had the capability to reproduce iodine behaviour for a narrow range of conditions (single temperature, no organic impurities, controlled pH steps). The current phase, a parametric study, was designed to evaluate the sensitivity of iodine behaviour codes to boundary conditions such as pH, dose rate, temperature and initial I- concentration. The codes used in this exercise were IODE (IPSN), IODE (NRIR), IMPAIR (GRS), INSPECT (AEAT), IMOD (AECL) and LIRIC (AECL). The parametric study described in this report identified several areas of discrepancy between the various codes. In general, the codes agree regarding qualitative trends, but their predictions regarding the actual amount of volatile iodine varied considerably. The largest source of the discrepancies between code predictions appears to be their different approaches to modelling the formation and destruction of organic iodides. A recommendation arising from this exercise is that an additional code comparison exercise be performed on organic iodide formation, against data obtained from intermediate-scale studies (two RTF (AECL, Canada) and two CAIMAN facility (IPSN, France) experiments have been chosen). This comparison will allow each of the code users to realistically evaluate and improve the organic iodide behaviour sub-models within their codes. (authors)

  17. From concatenated codes to graph codes

    DEFF Research Database (Denmark)

    Justesen, Jørn; Høholdt, Tom

    2004-01-01

    We consider codes based on simple bipartite expander graphs. These codes may be seen as the first step leading from product type concatenated codes to more complex graph codes. We emphasize constructions of specific codes of realistic lengths, and study the details of decoding by message passing...

  18. Development of input data to energy code for analysis of reactor fuel bundles

    International Nuclear Information System (INIS)

    Carre, F.O.; Todreas, N.E.

    1975-05-01

    The ENERGY 1 code is a semi-empirical method for predicting temperature distributions in wire wrapped rod bundles of a LMFBR. A comparison of ENERGY 1 and MISTRAL 2 is presented. The predictions of ENERGY 1 for special sets of data taken under geometric conditions at the limits of the code are analyzed. 14 references

  19. Results of the recent investigations of the accuracy of C-PORCA code

    International Nuclear Information System (INIS)

    Pos, I.

    2001-01-01

    During the last year it was an essential demand to establish the C-PORCA code accuracy concerning the newly developed BNFL fuel. As it is well known the BNFL fuel assembly developed for the NPP Paks contains fuel pins with gadolinium. These pins are situated near the central tube so their unusual effects to be investigated. At the present investigation the C-PORCA code and the APA-H Westinghouse code system results were compared. During this comparison the differences between node and pin wise power distributions have been calculated as well. The fulfilled test have reinforced the applicability and capability of the C-PORCA code for WWER-440 core analysis (Authors)

  20. Line-Shape Code Comparison through Modeling and Fitting of Experimental Spectra of the C ii 723-nm Line Emitted by the Ablation Cloud of a Carbon Pellet

    Directory of Open Access Journals (Sweden)

    Mohammed Koubiti

    2014-07-01

    Full Text Available Various codes of line-shape modeling are compared to each other through the profile of the C ii 723-nm line for typical plasma conditions encountered in the ablation clouds of carbon pellets, injected in magnetic fusion devices. Calculations were performed for a single electron density of 1017 cm−3 and two plasma temperatures (T = 2 and 4 eV. Ion and electron temperatures were assumed to be equal (Te = Ti = T. The magnetic field, B, was set equal to either to zero or 4 T. Comparisons between the line-shape modeling codes and two experimental spectra of the C ii 723-nm line, measured perpendicularly to the B-field in the Large Helical Device (LHD using linear polarizers, are also discussed.

  1. Representing high-dimensional data to intelligent prostheses and other wearable assistive robots: A first comparison of tile coding and selective Kanerva coding.

    Science.gov (United States)

    Travnik, Jaden B; Pilarski, Patrick M

    2017-07-01

    Prosthetic devices have advanced in their capabilities and in the number and type of sensors included in their design. As the space of sensorimotor data available to a conventional or machine learning prosthetic control system increases in dimensionality and complexity, it becomes increasingly important that this data be represented in a useful and computationally efficient way. Well structured sensory data allows prosthetic control systems to make informed, appropriate control decisions. In this study, we explore the impact that increased sensorimotor information has on current machine learning prosthetic control approaches. Specifically, we examine the effect that high-dimensional sensory data has on the computation time and prediction performance of a true-online temporal-difference learning prediction method as embedded within a resource-limited upper-limb prosthesis control system. We present results comparing tile coding, the dominant linear representation for real-time prosthetic machine learning, with a newly proposed modification to Kanerva coding that we call selective Kanerva coding. In addition to showing promising results for selective Kanerva coding, our results confirm potential limitations to tile coding as the number of sensory input dimensions increases. To our knowledge, this study is the first to explicitly examine representations for realtime machine learning prosthetic devices in general terms. This work therefore provides an important step towards forming an efficient prosthesis-eye view of the world, wherein prompt and accurate representations of high-dimensional data may be provided to machine learning control systems within artificial limbs and other assistive rehabilitation technologies.

  2. High fidelity analysis of BWR fuel assembly with COBRA-TF/PARCS and trace codes

    International Nuclear Information System (INIS)

    Abarca, A.; Miro, R.; Barrachina, T.; Verdu, G.; Soler, A.

    2013-01-01

    The growing importance of detailed reactor core and fuel assembly description for light water reactors (LWRs) as well as the sub-channel safety analysis requires high fidelity models and coupled neutronic/thermalhydraulic codes. Hand in hand with advances in the computer technology, the nuclear safety analysis is beginning to use a more detailed thermal hydraulics and neutronics. Previously, a PWR core and a 16 by 16 fuel assembly models were developed to test and validate our COBRA-TF/PARCS v2.7 (CTF/PARCS) coupled code. In this work, a comparison of the modeling and simulation advantages and disadvantages of modern 10 by 10 BWR fuel assembly with CTF/PARCS and TRACE codes has been done. The objective of the comparison is making known the main advantages of using the sub-channel codes to perform high resolution nuclear safety analysis. The sub-channel codes, like CTF, permits obtain accurate predictions, in two flow regime, of the thermalhydraulic parameters important to safety with high local resolution. The modeled BWR fuel assembly has 91 fuel rods (81 full length and 10 partial length fuel rods) and a big square central water rod. This assembly has been modeled with high level of detail with CTF code and using the BWR modeling parameters provided by TRACE. The same neutronic PARCS's model has been used for the simulation with both codes. To compare the codes a coupled steady state has be performed. (author)

  3. Bioethical and deontological approaches of the new occupational therapy code of ethics in Brazil

    Directory of Open Access Journals (Sweden)

    Leandro Correa Figueiredo

    2017-03-01

    Full Text Available Introduction: Currently, conflicts found in the health field bring new discussions on ethical and bioethical issues also in the Occupational Therapy domain. As noted in previous studies, the codes of professional ethics are not sufficient to face the current challenges daily experienced during professional practice. Objective: The present study aimed to find documental evidence of deontological and bioethical approaches in the new Brazilian code for Occupational Therapists through content analysis compared with the same analysis conducted in the preceding code. Method: Content analysis methods were applied to written documents to reveal deontological and bioethical approaches among textual fragments obtained from the new code of ethics. Results: The bioethical approaches found in the totality of the new code were increased in content and number (53.6% proportionally compared with those found in the former code. It seems that this increase was a result of the number of fragments classified in the justice-related category (22.6% - one of the most evident differences observed. Considering the ratio between the total number of fragments classified as professional autonomy and client autonomy in the new code - although the number of professional-related fragments have remained higher in comparison with client-related fragments - a significant decrease in the percentages of this ratio was detected. Conclusion: In conclusion, comparison between the codes revealed a bioethical embedding accompanied by a more client-centered practice, which reflects the way professionals have always conducted Occupational Therapy practice.

  4. Direct calculation of current drive efficiency in FISIC code

    International Nuclear Information System (INIS)

    Wright, J.C.; Phillips, C.K.; Bonoli, P.T.

    1996-01-01

    Two-dimensional RF modeling codes use a parameterization (1) of current drive efficiencies to calculate fast wave driven currents. This parameterization assumes a uniform quasi-linear diffusion coefficient and requires a priori knowledge of the wave polarizations. These difficulties may be avoided by a direct calculation of the quasilinear diffusion coefficient from the Kennel-Englemann form with the field polarizations calculated by the full wave code, FISIC (2). Current profiles are calculated using the adjoint formulation (3). Comparisons between the two formulations are presented. copyright 1996 American Institute of Physics

  5. Comparison of a Ring On-Chip Network and a Code-Division Multiple-Access On-Chip Network

    Directory of Open Access Journals (Sweden)

    Xin Wang

    2007-01-01

    Full Text Available Two network-on-chip (NoC designs are examined and compared in this paper. One design applies a bidirectional ring connection scheme, while the other design applies a code-division multiple-access (CDMA connection scheme. Both of the designs apply globally asynchronous locally synchronous (GALS scheme in order to deal with the issue of transferring data in a multiple-clock-domain environment of an on-chip system. The two NoC designs are compared with each other by their network structures, data transfer principles, network node structures, and their asynchronous designs. Both the synchronous and the asynchronous designs of the two on-chip networks are realized using a hardware-description language (HDL in order to make the entire designs suit the commonly used synchronous design tools and flow. The performance estimation and comparison of the two NoC designs which are based on the HDL realizations are addressed. By comparing the two NoC designs, the advantages and disadvantages of applying direct connection and CDMA connection schemes in an on-chip communication network are discussed.

  6. The Fourth Workshop on Lineshape Code Comparison: Line Merging

    Directory of Open Access Journals (Sweden)

    Spiros Alexiou

    2018-03-01

    Full Text Available For a given set of plasma parameters, along a single series (Lyman, Balmer, etc. the lines with higher principal quantum number (n lines get progressively wider, closer to each other, and start merging for a certain critical n. In the present work, four different codes (with further options are used to calculate the entire Balmer series for moderate and high electron densities. Particular attention is paid to the relevant physics, such as the cutoff criteria, strong and penetrating electron collisions.

  7. Performance of code 'FAIR' in IAEA CRP on FUMEX

    International Nuclear Information System (INIS)

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

    1996-01-01

    A modern fuel performance analysis code FAIR has been developed for analysing high burnup fuel pins of water/heavy water cooled reactors. The code employs finite element method for modelling thermo mechanical behaviour of fuel pins and mechanistic models for modelling various physical and chemical phenomena affecting the behaviour of nuclear reactor fuel pins. High burnup affects such as pellet thermal conductivity degradation, enhanced fission gas release and radial flux redistribution are incorporated in the code FAIR. The code FAIR is capable of performing statistical analysis of fuel pins using Monte Carlo technique. The code is implemented on BARC parallel processing system ANUPAM. The code has recently participated in an International Atomic Energy Agency (IAEA) coordinated research program (CRP) on fuel modelling at extended burnups (FUMEX). Nineteen agencies from different countries participated in this exercise. In this CRP, spread over a period of three years, a number of high burnup fuel pins irradiated at Halden reactor are analysed. The first phase of the CRP is a blind code comparison exercise, where the computed results are compared with experimental results. The second phase consists of modifications to the code based on the experimental results of first phase and statistical analysis of fuel pins. The performance of the code FAIR in this CRP has been very good. The present report highlights the main features of code FAIR and its performance in the IAEA CRP on FUMEX. 14 refs., 5 tabs., ills

  8. Verification of the DEFENS Code through the CANDU Problems with Rectangular Geometry

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Eun Hyun; Song, Yong Mann [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Because a finite element method (FEM) based code can explicitly describe the core geometry, it has an advantage in a core analysis such as the CANDU core. For the reactor physics calculation in the CANDU core, the RFSP-IST code is used for the core analysis, and the RFSP-IST code is based on the finite difference method (FDM). Thus, the convergence with the mesh size and the geometry shape is not consistent. In this research, the convergence with the mesh size of the RFSP code is investigated, a method comparison between the FEM and FDM is done for the usefulness of the FEM based code with the same rectangular geometry. The target problems are the imaginary core and initial core with the uniform parameter, which is produced by the WIMS-IST code based on the parameters of Wolsong unit 1. The reference solution is generated by running the multi-group calculation of the McCARD code. In this research, the convergence of the RFSP code is investigated and the DEFENS code is compared with the RFSP code for the imaginary and initial cores. The accuracy of the DEFENS code and the disadvantage of the RFSP code are verified.

  9. Comparison of PWR-IMF and FR fuel cycles

    International Nuclear Information System (INIS)

    Darilek, Petr; Zajac, Radoslav; Breza, Juraj; Necas, Vladimir

    2007-01-01

    The paper gives a comparison of PWR (Russia origin VVER-440) cycle with improved micro-heterogeneous inert matrix fuel assemblies and FR cycle. Micro-heterogeneous combined assembly contains transmutation pins with Pu and MAs from burned uranium reprocessing and standard uranium pins. Cycle analyses were performed by HELIOS spectral code and SCALE code system. Comparison is based on fuel cycle indicators, used in the project RED-IMPACT - part of EU FP6. Advantages of both closed cycles are pointed out. (authors)

  10. Development of the point-depletion code DEPTH

    International Nuclear Information System (INIS)

    She, Ding; Wang, Kan; Yu, Ganglin

    2013-01-01

    Highlights: ► The DEPTH code has been developed for the large-scale depletion system. ► DEPTH uses the data library which is convenient to couple with MC codes. ► TTA and matrix exponential methods are implemented and compared. ► DEPTH is able to calculate integral quantities based on the matrix inverse. ► Code-to-code comparisons prove the accuracy and efficiency of DEPTH. -- Abstract: The burnup analysis is an important aspect in reactor physics, which is generally done by coupling of transport calculations and point-depletion calculations. DEPTH is a newly-developed point-depletion code of handling large burnup depletion systems and detailed depletion chains. For better coupling with Monte Carlo transport codes, DEPTH uses data libraries based on the combination of ORIGEN-2 and ORIGEN-S and allows users to assign problem-dependent libraries for each depletion step. DEPTH implements various algorithms of treating the stiff depletion systems, including the Transmutation trajectory analysis (TTA), the Chebyshev Rational Approximation Method (CRAM), the Quadrature-based Rational Approximation Method (QRAM) and the Laguerre Polynomial Approximation Method (LPAM). Three different modes are supported by DEPTH to execute the decay, constant flux and constant power calculations. In addition to obtaining the instantaneous quantities of the radioactivity, decay heats and reaction rates, DEPTH is able to calculate the integral quantities by a time-integrated solver. Through calculations compared with ORIGEN-2, the validity of DEPTH in point-depletion calculations is proved. The accuracy and efficiency of depletion algorithms are also discussed. In addition, an actual pin-cell burnup case is calculated to illustrate the DEPTH code performance in coupling with the RMC Monte Carlo code

  11. Toric Varieties and Codes, Error-correcting Codes, Quantum Codes, Secret Sharing and Decoding

    DEFF Research Database (Denmark)

    Hansen, Johan Peder

    We present toric varieties and associated toric codes and their decoding. Toric codes are applied to construct Linear Secret Sharing Schemes (LSSS) with strong multiplication by the Massey construction. Asymmetric Quantum Codes are obtained from toric codes by the A.R. Calderbank P.W. Shor and A.......M. Steane construction of stabilizer codes (CSS) from linear codes containing their dual codes....

  12. Code compression for VLIW embedded processors

    Science.gov (United States)

    Piccinelli, Emiliano; Sannino, Roberto

    2004-04-01

    The implementation of processors for embedded systems implies various issues: main constraints are cost, power dissipation and die area. On the other side, new terminals perform functions that require more computational flexibility and effort. Long code streams must be loaded into memories, which are expensive and power consuming, to run on DSPs or CPUs. To overcome this issue, the "SlimCode" proprietary algorithm presented in this paper (patent pending technology) can reduce the dimensions of the program memory. It can run offline and work directly on the binary code the compiler generates, by compressing it and creating a new binary file, about 40% smaller than the original one, to be loaded into the program memory of the processor. The decompression unit will be a small ASIC, placed between the Memory Controller and the System bus of the processor, keeping unchanged the internal CPU architecture: this implies that the methodology is completely transparent to the core. We present comparisons versus the state-of-the-art IBM Codepack algorithm, along with its architectural implementation into the ST200 VLIW family core.

  13. Study of experimental validation for combustion analysis of GOTHIC code

    International Nuclear Information System (INIS)

    Lee, J. Y.; Yang, S. Y.; Park, K. C.; Jeong, S. H.

    2001-01-01

    In this study, present lumped and subdivided GOTHIC6 code analyses of the premixed hydrogen combustion experiment at the Seoul National University and comparison with the experiment results. The experimental facility has 16367 cc free volume and rectangular shape. And the test was performed with unit equivalence ratio of the hydrogen and air, and with various location of igniter position. Using the lumped and mechanistic combustion model in GOTHIC6 code, the experiments were simulated with the same conditions. In the comparison between experiment and calculated results, the GOTHIC6 prediction of the combustion response does not compare well with the experiment results. In the point of combustion time, the lumped combustion model of GOTHIC6 code does not simulate the physical phenomena of combustion appropriately. In the case of mechanistic combustion model, the combustion time is predicted well, but the induction time of calculation data is longer than the experiment data remarkably. Also, the laminar combustion model of GOTHIC6 has deficiency to simulate combustion phenomena unless control the user defined value appropriately. And the pressure is not a proper variable that characterize the three dimensional effect of combustion

  14. DYNSUB: A high fidelity coupled code system for the evaluation of local safety parameters – Part II: Comparison of different temporal schemes

    International Nuclear Information System (INIS)

    Gomez-Torres, Armando Miguel; Sanchez-Espinoza, Victor Hugo; Ivanov, Kostadin; Macian-Juan, Rafael

    2012-01-01

    Highlights: ► A fixed point iteration (FPI) is implemented in DYNSUB. ► Comparisons between the explicit scheme and the FPI are done. ► The FPI scheme allows moving from one time step to the other with converged solution. ► FPI allows the use of larger time steps without compromising the accuracy of results. ► FPI results are promising and represent an option in order to optimize calculations. -- Abstract: DYNSUB is a novel two-way pin-based coupling of the simplified transport (SP 3 ) version of DYN3D with the subchannel code SUBCHANFLOW. The new coupled code system allows for a more realistic description of the core behaviour under steady state and transients conditions, and has been widely described in Part I of this paper. Additionally to the explicit coupling developed and described in Part I, a nested loop iteration or fixed point iteration (FPI) is implemented in DYNSUB. A FPI is not an implicit scheme but approximates it by adding an iteration loop to the current explicit scheme. The advantage of the method is that it allows the use of larger time steps; however the nested loop iteration could take much more time in getting a converged solution that could be less efficient than the explicit scheme with small time steps. A comparison of the two temporal schemes is performed. The results using FPI are very promising and represent a very good option in order to optimize computational times without losing accuracy. However it is also shown that a FPI scheme can produce inaccurate results if the time step is not chosen in agreement with the analyzed transient.

  15. Results from the Metis code participation to the Hydrocoin exercise

    International Nuclear Information System (INIS)

    Raimbault, P.

    1987-04-01

    The METIS code, developed at the ENSMP is a 2D finite element radionuclide transport and groundwater flow model based on the hypothesis of an equivalent porous medium with an explicit description of the main fractures. It is integrated in the global risk assessment code MELODIE for nuclear waste repositories in geological formations. The participation of the METIS code to the HYDROCOIN exercise is of prime importance for its development and its incorporation in the performance assessment procedure in France. Results from HYDROCOIN cases show that the code can handle correctly fractured media, high permeability contrast formations and buoyancy effects. A 3D version of the code has been developed for carrying comparisons of field experiments and groundwater flow models in HYDROCOIN level 2. In order to carry out the exercise, several pre and post-processing programs were developed and integrated in a conversational module. They include: contour plots, velocity field representations, interpolations, particule tracking routines and uncertainty and sensitivity analysis modules

  16. Description and validation of ANTEO, an optimised PC code the thermalhydraulic analysis of fuel bundles

    International Nuclear Information System (INIS)

    Cevolani, S.

    1995-01-01

    The paper deals with the description of a Personal Computer oriented subchannel code, devoted to the steady state thermal hydraulic analysis of nuclear reactor fuel bundles. The development of such a code was made possible by two facts: firstly, the increase, in the computing power of the desk machines; secondly, the fact that several years of experience into operate subchannels codes have shown how to simplify many of the physical models without a sensible loss of accuracy. For sake of validation, the developed code was compared with a traditional subchannel code, the COBRA one. The results of the comparison show a very good agreement between the two codes. (author)

  17. Improved lossless intra coding for H.264/MPEG-4 AVC.

    Science.gov (United States)

    Lee, Yung-Lyul; Han, Ki-Hun; Sullivan, Gary J

    2006-09-01

    A new lossless intra coding method based on sample-by-sample differential pulse code modulation (DPCM) is presented as an enhancement of the H.264/MPEG-4 AVC standard. The H.264/AVC design includes a multidirectional spatial prediction method to reduce spatial redundancy by using neighboring samples as a prediction for the samples in a block of data to be encoded. In the new lossless intra coding method, the spatial prediction is performed based on samplewise DPCM instead of in the block-based manner used in the current H.264/AVC standard, while the block structure is retained for the residual difference entropy coding process. We show that the new method, based on samplewise DPCM, does not have a major complexity penalty, despite its apparent pipeline dependencies. Experiments show that the new lossless intra coding method reduces the bit rate by approximately 12% in comparison with the lossless intra coding method previously included in the H.264/AVC standard. As a result, the new method is currently being adopted into the H.264/AVC standard in a new enhancement project.

  18. Automatic coding method of the ACR Code

    International Nuclear Information System (INIS)

    Park, Kwi Ae; Ihm, Jong Sool; Ahn, Woo Hyun; Baik, Seung Kook; Choi, Han Yong; Kim, Bong Gi

    1993-01-01

    The authors developed a computer program for automatic coding of ACR(American College of Radiology) code. The automatic coding of the ACR code is essential for computerization of the data in the department of radiology. This program was written in foxbase language and has been used for automatic coding of diagnosis in the Department of Radiology, Wallace Memorial Baptist since May 1992. The ACR dictionary files consisted of 11 files, one for the organ code and the others for the pathology code. The organ code was obtained by typing organ name or code number itself among the upper and lower level codes of the selected one that were simultaneous displayed on the screen. According to the first number of the selected organ code, the corresponding pathology code file was chosen automatically. By the similar fashion of organ code selection, the proper pathologic dode was obtained. An example of obtained ACR code is '131.3661'. This procedure was reproducible regardless of the number of fields of data. Because this program was written in 'User's Defined Function' from, decoding of the stored ACR code was achieved by this same program and incorporation of this program into program in to another data processing was possible. This program had merits of simple operation, accurate and detail coding, and easy adjustment for another program. Therefore, this program can be used for automation of routine work in the department of radiology

  19. Computer code selection criteria for flow and transport code(s) to be used in undisturbed vadose zone calculations for TWRS environmental analyses

    International Nuclear Information System (INIS)

    Mann, F.M.

    1998-01-01

    The Tank Waste Remediation System (TWRS) is responsible for the safe storage, retrieval, and disposal of waste currently being held in 177 underground tanks at the Hanford Site. In order to successfully carry out its mission, TWRS must perform environmental analyses describing the consequences of tank contents leaking from tanks and associated facilities during the storage, retrieval, or closure periods and immobilized low-activity tank waste contaminants leaving disposal facilities. Because of the large size of the facilities and the great depth of the dry zone (known as the vadose zone) underneath the facilities, sophisticated computer codes are needed to model the transport of the tank contents or contaminants. This document presents the code selection criteria for those vadose zone analyses (a subset of the above analyses) where the hydraulic properties of the vadose zone are constant in time the geochemical behavior of the contaminant-soil interaction can be described by simple models, and the geologic or engineered structures are complicated enough to require a two-or three dimensional model. Thus, simple analyses would not need to use the fairly sophisticated codes which would meet the selection criteria in this document. Similarly, those analyses which involve complex chemical modeling (such as those analyses involving large tank leaks or those analyses involving the modeling of contaminant release from glass waste forms) are excluded. The analyses covered here are those where the movement of contaminants can be relatively simply calculated from the moisture flow. These code selection criteria are based on the information from the low-level waste programs of the US Department of Energy (DOE) and of the US Nuclear Regulatory Commission as well as experience gained in the DOE Complex in applying these criteria. Appendix table A-1 provides a comparison between the criteria in these documents and those used here. This document does not define the models (that

  20. Description and assessment of structural and temperature models in the FRAP-T6 code

    International Nuclear Information System (INIS)

    Siefken, L.J.

    1983-01-01

    The FRAP-T6 code was developed at the Idaho National Engineering Laboratory (INEL) for the purpose of calculating the transient performance of light water reactor fuel rods during reactor transients ranging from mild operational transients to severe hypothetical loss-of-coolant accidents. An important application of the FRAP-T6 code is to calculate the structural performance of fuel rod cladding. The capabilities of the FRAP-T6 code are assessed by comparisons of code calculations with the measurements of several hundred in-pile experiments on fuel rods. The results of the assessments show that the code accurately and efficiently models the structural and thermal response of fuel rods

  1. GAM-HEAT -- a computer code to compute heat transfer in complex enclosures

    International Nuclear Information System (INIS)

    Cooper, R.E.; Taylor, J.R.; Kielpinski, A.L.; Steimke, J.L.

    1991-02-01

    The GAM-HEAT code was developed for heat transfer analyses associated with postulated Double Ended Guillotine Break Loss Of Coolant Accidents (DEGB LOCA) resulting in a drained reactor vessel. In these analyses the gamma radiation resulting from fission product decay constitutes the primary source of energy as a function of time. This energy is deposited into the various reactor components and is re- radiated as thermal energy. The code accounts for all radiant heat exchanges within and leaving the reactor enclosure. The SRS reactors constitute complex radiant exchange enclosures since there are many assemblies of various types within the primary enclosure and most of the assemblies themselves constitute enclosures. GAM-HEAT accounts for this complexity by processing externally generated view factors and connectivity matrices, and also accounts for convective, conductive, and advective heat exchanges. The code is applicable for many situations involving heat exchange between surfaces within a radiatively passive medium. The GAM-HEAT code has been exercised extensively for computing transient temperatures in SRS reactors with specific charges and control components. Results from these computations have been used to establish the need for and to evaluate hardware modifications designed to mitigate results of postulated accident scenarios, and to assist in the specification of safe reactor operating power limits. The code utilizes temperature dependence on material properties. The efficiency of the code has been enhanced by the use of an iterative equation solver. Verification of the code to date consists of comparisons with parallel efforts at Los Alamos National Laboratory and with similar efforts at Westinghouse Science and Technology Center in Pittsburgh, PA, and benchmarked using problems with known analytical or iterated solutions. All comparisons and tests yield results that indicate the GAM-HEAT code performs as intended

  2. The SABRE code for fuel rod cluster thermohydraulics

    International Nuclear Information System (INIS)

    Macdougall, J.D.; Lillington, J.N.

    1984-01-01

    This paper describes the capabilities of the SABRE code for the calculation of single phase and two phase fluid flow and temperature in fuel pin bundles, discusses the methods used in the modelling and solution of the problem, and presents some results including comparison with experiments. The SABRE code permits calculation of steady-state or transient, single or two phase flows and the geometrical options include general representation of grids, wire wraps, multiple blockages, bowed pins, etc. The derivation and solution of the difference equations is discussed. Emphasis is given to the derivation of the spatial differences in triangular subchannel geometry, and the use of central, upward or vector upwind schemes. The method of solution of the difference equations is described for both steady state and transient problems. Together with these topics we consider the problems involved in turbulence modelling and how it is implemented in SABRE. This includes supporting work with a fine scale curvilinear coordinate programme to provide turbulence source data. The problem of modelling boiling flows is discussed, with particular reference to the numerical problems caused by the rapid density change on boiling. The final part of the paper presents applications of the code to the analysis of blockage situations, the study of flow and power transients and analysis of natural circulation within clusters to demonstrate the scope of the code and compare with available experimental results. The comparisons include the calculation of a flow pressure drop characteristic of a boiling channel showing the Ledinegg instability, examples of overpower and flow rundown transients which lead to coolant boiling, and calculation of natural circulation within a rod cluster. (orig./GL)

  3. International standard problem (ISP) No. 43 Rapid boron-dilution transient tests for code verification. Comparison report

    International Nuclear Information System (INIS)

    2001-03-01

    established to be the average temperature of the 24 channel-centered thermocouples at the downcomer outlet. The frequency of data acquisition for code predictions was set to 2 Hz, which corresponds to the acquisition frequency of the two experimental setups. The figures of merits were expanded at the second workshop (February 2000) after experimental and computational data was collected. The revised list of figures of merit included the level average temperatures before the downcomer expansion and just above the injection cold leg. No turbulence measurements were provided. Quantitative measures of azimuthal distribution were also added to the figures of merit list: standard deviations, minima and maxima at the aforementioned levels. Furthermore, since discrepancies in flow pattern are noticeable among the tests of series A, the flow patterns obtained experimentally and from computations at 4 distinct times during the transients, were also extracted. Data were reduced into a standard format that facilitates comparisons. Ten organizations from eight countries participated in the exercise. All used computational fluid dynamics (CFD) codes to obtain predictions. Most participants used commercial software (Fluent, CFX) while some used independently developed codes (TRIO-U, PLASHY). The number of nodes used to discretize the simulation volume ranged over two orders of magnitude. The turbulence models ranged from no turbulence to renormalization group (RNG) k-epsilon. The spatial and temporal discretizations also data for first order figures of merit. Contrary to previous observations, flow patterns showed high sensitivity to buoyancy effects and the level to which temperature dependent properties and buoyancy-related turbulence effects were included in code models affected the predictions of higher order figures of merit. One more important factor needs to be considered in this discussion, namely the local refinement in regions where the flow is subject to large gradients or where

  4. The PLTEMP V2.1 code

    International Nuclear Information System (INIS)

    Olson, A.P.

    2003-01-01

    Recent improvements to the computer code PLTEMP/ANL V2.1 are described. A new iterative, error-minimization solution technique is used to obtain the thermal distribution both within each fuel plate, and along the axial length of each coolant channel. A new, radial geometry solution is available for tube-type fuel assemblies. Software comparisons of these and other new models are described. Applications to Russian-designed IRT-type research reactors are described. (author)

  5. Feasibility Study of Core Design with a Monte Carlo Code for APR1400 Initial core

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jinsun; Chang, Do Ik; Seong, Kibong [KEPCO NF, Daejeon (Korea, Republic of)

    2014-10-15

    The Monte Carlo calculation becomes more popular and useful nowadays due to the rapid progress in computing power and parallel calculation techniques. There have been many attempts to analyze a commercial core by Monte Carlo transport code using the enhanced computer capability, recently. In this paper, Monte Carlo calculation of APR1400 initial core has been performed and the results are compared with the calculation results of conventional deterministic code to find out the feasibility of core design using Monte Carlo code. SERPENT, a 3D continuous-energy Monte Carlo reactor physics burnup calculation code is used for this purpose and the KARMA-ASTRA code system, which is used for a deterministic code of comparison. The preliminary investigation for the feasibility of commercial core design with Monte Carlo code was performed in this study. Simplified core geometry modeling was performed for the reactor core surroundings and reactor coolant model is based on two region model. The reactivity difference at HZP ARO condition between Monte Carlo code and the deterministic code is consistent with each other and the reactivity difference during the depletion could be reduced by adopting the realistic moderator temperature. The reactivity difference calculated at HFP, BOC, ARO equilibrium condition was 180 ±9 pcm, with axial moderator temperature of a deterministic code. The computing time will be a significant burden at this time for the application of Monte Carlo code to the commercial core design even with the application of parallel computing because numerous core simulations are required for actual loading pattern search. One of the remedy will be a combination of Monte Carlo code and the deterministic code to generate the physics data. The comparison of physics parameters with sophisticated moderator temperature modeling and depletion will be performed for a further study.

  6. 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

  7. Ex-vessel break in ITER divertor cooling loop analysis with the ECART code

    CERN Document Server

    Cambi, G; Parozzi, F; Porfiri, MT

    2003-01-01

    A hypothetical double-ended pipe rupture in the ex-vessel section of the International Thermonuclear Experimental Reactor (ITER) divertor primary heat transfer system during pulse operation has been assessed using the nuclear source term ECART code. That code was originally designed and validated for traditional nuclear power plant safety analyses, and has been internationally recognized as a relevant nuclear source term codes for nuclear fission plants. It permits the simulation of chemical reactions and transport of radioactive gases and aerosols under two-phase flow transients in generic flow systems, using a built-in thermal-hydraulic model. A comparison with the results given in ITER Generic Site Safety Report, obtained using a thermal-hydraulic system code (ATHENA), a containment code (INTRA) and an aerosol transportation code (NAUA), in a sequential way, is also presented and discussed.

  8. Orthopedics coding and funding.

    Science.gov (United States)

    Baron, S; Duclos, C; Thoreux, P

    2014-02-01

    The French tarification à l'activité (T2A) prospective payment system is a financial system in which a health-care institution's resources are based on performed activity. Activity is described via the PMSI medical information system (programme de médicalisation du système d'information). The PMSI classifies hospital cases by clinical and economic categories known as diagnosis-related groups (DRG), each with an associated price tag. Coding a hospital case involves giving as realistic a description as possible so as to categorize it in the right DRG and thus ensure appropriate payment. For this, it is essential to understand what determines the pricing of inpatient stay: namely, the code for the surgical procedure, the patient's principal diagnosis (reason for admission), codes for comorbidities (everything that adds to management burden), and the management of the length of inpatient stay. The PMSI is used to analyze the institution's activity and dynamism: change on previous year, relation to target, and comparison with competing institutions based on indicators such as the mean length of stay performance indicator (MLS PI). The T2A system improves overall care efficiency. Quality of care, however, is not presently taken account of in the payment made to the institution, as there are no indicators for this; work needs to be done on this topic. Copyright © 2014. Published by Elsevier Masson SAS.

  9. Validation of the containment code Sirius: interpretation of an explosion experiment on a scale model

    International Nuclear Information System (INIS)

    Blanchet, Y.; Obry, P.; Louvet, J.; Deshayes, M.; Phalip, C.

    1979-01-01

    The explicit 2-D axisymmetric Langrangian code SIRIUS, developed at the CEA/DRNR, Cadarache, deals with transient compressive flows in deformable primary tanks with more or less complex internal component geometries. This code has been subjected to a two-year intensive validation program on scale model experiments and a number of improvements have been incorporated. This paper presents a recent calculation of one of these experiments using the SIRIUS code, and the comparison with experimental results shows the encouraging possibilities of this Lagrangian code

  10. State of art in FE-based fuel performance codes

    International Nuclear Information System (INIS)

    Kim, Hyo Chan; Yang, Yong Sik; Kim, Dae Ho; Bang, Je Geon; Kim, Sun Ki; Koo, Yang Hyun

    2013-01-01

    Fuel performance codes approximate this complex behavior using an axisymmetric, axially-stacked, one-dimensional radial representation to save computation cost. However, the need for improved modeling of PCMI and, particularly, the importance of multidimensional capability for accurate fuel performance simulation has been identified as safety margin decreases. Finite element (FE) method that is reliable and proven solution in mechanical field has been introduced into fuel performance codes for multidimensional analysis. The present state of the art in numerical simulation of FE-based fuel performance predominantly involves 2-D axisymmetric model and 3-D volumetric model. The FRAPCON and FRAPTRAN own 1.5-D and 2-D FE model to simulate PCMI and cladding ballooning. In 2-D simulation, the FALCON code, developed by EPRI, is a 2-D (R-Z and R-θ) fully thermal-mechanically coupled steady-state and transient FE-based fuel behavior code. The French codes TOUTATIS and ALCYONE which are 3-D, and typically used to investigate localized behavior. In 2008, the Idaho National Laboratory (INL) has been developing multidimensional (2-D and 3-D) nuclear fuel performance code called BISON. In this paper, the current state of FE-based fuel performance code and their models are presented. Based on investigation into the codes, requirements and direction of development for new FE-based fuel performance code can be discussed. Based on comparison of models in FE-based fuel performance code, status of art in the codes can be discussed. A new FE-based fuel performance code should include typical pellet and cladding models which all codes own. In particular, specified pellet and cladding model such as gaseous swelling and high burnup structure (HBS) model should be developed to improve accuracy of code as well as consider AC condition. To reduce computation cost, the approximated gap and the optimized contact model should be also developed

  11. A seismic data compression system using subband coding

    Science.gov (United States)

    Kiely, A. B.; Pollara, F.

    1995-01-01

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

  12. Code Cactus; Code Cactus

    Energy Technology Data Exchange (ETDEWEB)

    Fajeau, M; Nguyen, L T; Saunier, J [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

    1966-09-01

    This code handles the following problems: -1) Analysis of thermal experiments on a water loop at high or low pressure; steady state or transient behavior; -2) Analysis of thermal and hydrodynamic behavior of water-cooled and moderated reactors, at either high or low pressure, with boiling permitted; fuel elements are assumed to be flat plates: - Flowrate in parallel channels coupled or not by conduction across plates, with conditions of pressure drops or flowrate, variable or not with respect to time is given; the power can be coupled to reactor kinetics calculation or supplied by the code user. The code, containing a schematic representation of safety rod behavior, is a one dimensional, multi-channel code, and has as its complement (FLID), a one-channel, two-dimensional code. (authors) [French] Ce code permet de traiter les problemes ci-dessous: 1. Depouillement d'essais thermiques sur boucle a eau, haute ou basse pression, en regime permanent ou transitoire; 2. Etudes thermiques et hydrauliques de reacteurs a eau, a plaques, a haute ou basse pression, ebullition permise: - repartition entre canaux paralleles, couples on non par conduction a travers plaques, pour des conditions de debit ou de pertes de charge imposees, variables ou non dans le temps; - la puissance peut etre couplee a la neutronique et une representation schematique des actions de securite est prevue. Ce code (Cactus) a une dimension d'espace et plusieurs canaux, a pour complement Flid qui traite l'etude d'un seul canal a deux dimensions. (auteurs)

  13. Los Alamos and Lawrence Livermore National Laboratories Code-to-Code Comparison of Inter Lab Test Problem 1 for Asteroid Impact Hazard Mitigation

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, Robert P. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Miller, Paul [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Howley, Kirsten [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ferguson, Jim Michael [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Gisler, Galen Ross [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Plesko, Catherine Suzanne [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Managan, Rob [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Owen, Mike [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wasem, Joseph [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Bruck-Syal, Megan [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-01-15

    The NNSA Laboratories have entered into an interagency collaboration with the National Aeronautics and Space Administration (NASA) to explore strategies for prevention of Earth impacts by asteroids. Assessment of such strategies relies upon use of sophisticated multi-physics simulation codes. This document describes the task of verifying and cross-validating, between Lawrence Livermore National Laboratory (LLNL) and Los Alamos National Laboratory (LANL), modeling capabilities and methods to be employed as part of the NNSA-NASA collaboration. The approach has been to develop a set of test problems and then to compare and contrast results obtained by use of a suite of codes, including MCNP, RAGE, Mercury, Ares, and Spheral. This document provides a short description of the codes, an overview of the idealized test problems, and discussion of the results for deflection by kinetic impactors and stand-off nuclear explosions.

  14. The use of the codes from MCU family for calculations of WWER type reactors

    International Nuclear Information System (INIS)

    Abagijan, L.P.; Alexeyev, N.I.; Bryzgalov, V.I.; Gomin, E.A.; Glushkov, A.E.; Gorodkov, S.S.; Gurevich, M.I.; Kalugin, M.A.; Marin, S.V.; Shkarovsky, D.A.; Yudkevich, M.S.

    2000-01-01

    The MCU-RFFI/A and MCU-REA codes developed within the framework of the long term MCU project are widely used for calculations of neutron physic characteristics of WWER type reactors. Complete descriptions of the codes are available in both Russian and English. The codes are verified and validated by means of the comparison of calculated results with experimental data and mathematical benchmarks. The codes are licensed by Russian Nuclear and Criticality Safety Regulatory Body (Gosatomnadzor RF) (Code Passports: N 61 of 17.10.1966 and N 115 of 02.03.2000 accordingly)). The report gives examples of WWER reactor physic tasks important for practice solved using the codes from the MCU family. Some calculational results are given too. (Authors)

  15. Source Coding for Wireless Distributed Microphones in Reverberant Environments

    DEFF Research Database (Denmark)

    Zahedi, Adel

    2016-01-01

    . However, it comes with the price of several challenges, including the limited power and bandwidth resources for wireless transmission of audio recordings. In such a setup, we study the problem of source coding for the compression of the audio recordings before the transmission in order to reduce the power...... consumption and/or transmission bandwidth by reduction in the transmission rates. Source coding for wireless microphones in reverberant environments has several special characteristics which make it more challenging in comparison with regular audio coding. The signals which are acquired by the microphones......Modern multimedia systems are more and more shifting toward distributed and networked structures. This includes audio systems, where networks of wireless distributed microphones are replacing the traditional microphone arrays. This allows for flexibility of placement and high spatial diversity...

  16. Comparison of the sand liquefaction estimated based on codes and practical earthquake damage phenomena

    Science.gov (United States)

    Fang, Yi; Huang, Yahong

    2017-12-01

    Conducting sand liquefaction estimated based on codes is the important content of the geotechnical design. However, the result, sometimes, fails to conform to the practical earthquake damages. Based on the damage of Tangshan earthquake and engineering geological conditions, three typical sites are chosen. Moreover, the sand liquefaction probability was evaluated on the three sites by using the method in the Code for Seismic Design of Buildings and the results were compared with the sand liquefaction phenomenon in the earthquake. The result shows that the difference between sand liquefaction estimated based on codes and the practical earthquake damage is mainly attributed to the following two aspects: The primary reasons include disparity between seismic fortification intensity and practical seismic oscillation, changes of groundwater level, thickness of overlying non-liquefied soil layer, local site effect and personal error. Meanwhile, although the judgment methods in the codes exhibit certain universality, they are another reason causing the above difference due to the limitation of basic data and the qualitative anomaly of the judgment formulas.

  17. Radiative and SOL experiments in open and baffled divertors on DIII-D

    International Nuclear Information System (INIS)

    Allen, S.L.; Brooks, N.H.; Bastasz, R.

    1998-11-01

    The authors present recent progress towards an understanding of the physical processes in the divertor and scrape-off-layer (SOL) plasmas in DIII-D. This has been made possible by a combination of new diagnostics, improved computational models, and changes in divertor geometry. They have focused primarily on ELMing H-mode discharges. The physics of Partially Detached Divertor (PDD) plasmas, with divertor heat flux reduction by divertor radiation enhancement using D 2 puffing, has been studied in 2-D, and a model of the heat and particle transport has been developed that includes conduction, convection, ionization, recombination, and flows. Plasma and impurity particle flows have been measured with Mach probes and spectroscopy and these flows have been compared with the UEDGE model. The model now includes self-consistent calculations of carbon impurities. Impurity radiation has been increased in the divertor and SOL with puff and pump techniques using SOL D 2 puffing, divertor cryopumping, and argon puffing. The important physical processes in plasma-wall interactions have been examined with a DiMES probe, plasma characterization near the divertor plate, and the REDEP code. Experiments comparing single-null (SN) plasma operation in baffled and open divertors have demonstrated a change in the edge plasma profiles. These results are consistent with a reduction in the core ionization source calculated with UEDGE. Divertor particle control in ELMing H-mode with pumping and baffling has resulted in reduction in H-mode core densities to n e /n gw ∼ 0.25. Divertor particle exhaust and heat flux has been studied as the plasma shape was varied from a lower SN, to a balanced double null (DN), and finally to an upper SN

  18. Development of CAP code for nuclear power plant containment: Lumped model

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Soon Joon, E-mail: sjhong90@fnctech.com [FNC Tech. Co. Ltd., Heungdeok 1 ro 13, Giheung-gu, Yongin-si, Gyeonggi-do 446-908 (Korea, Republic of); Choo, Yeon Joon; Hwang, Su Hyun; Lee, Byung Chul [FNC Tech. Co. Ltd., Heungdeok 1 ro 13, Giheung-gu, Yongin-si, Gyeonggi-do 446-908 (Korea, Republic of); Ha, Sang Jun [Central Research Institute, Korea Hydro & Nuclear Power Company, Ltd., 70, 1312-gil, Yuseong-daero, Yuseong-gu, Daejeon 305-343 (Korea, Republic of)

    2015-09-15

    Highlights: • State-of-art containment analysis code, CAP, has been developed. • CAP uses 3-field equations, water level oriented upwind scheme, local head model. • CAP has a function of linked calculation with reactor coolant system code. • CAP code assessments showed appropriate prediction capabilities. - Abstract: CAP (nuclear Containment Analysis Package) code has been developed in Korean nuclear society for the analysis of nuclear containment thermal hydraulic behaviors including pressure and temperature trends and hydrogen concentration. Lumped model of CAP code uses 2-phase, 3-field equations for fluid behaviors, and has appropriate constitutive equations, 1-dimensional heat conductor model, component models, trip and control models, and special process models. CAP can run in a standalone mode or a linked mode with a reactor coolant system analysis code. The linked mode enables the more realistic calculation of a containment response and is expected to be applicable to a more complicated advanced plant design calculation. CAP code assessments were carried out by gradual approaches: conceptual problems, fundamental phenomena, component and principal phenomena, experimental validation, and finally comparison with other code calculations on the base of important phenomena identifications. The assessments showed appropriate prediction capabilities of CAP.

  19. Sensibility analysis of fuel depletion using different nuclear fuel depletion codes

    Energy Technology Data Exchange (ETDEWEB)

    Martins, F.; Velasquez, C.E.; Castro, V.F.; Pereira, C.; Silva, C. A. Mello da, E-mail: felipmartins94@gmail.com, E-mail: carlosvelcab@hotmail.com, E-mail: victorfariascastro@gmail.com, E-mail: claubia@nuclear.ufmg.br, E-mail: clarysson@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-07-01

    Nowadays, the utilization of different nuclear codes to perform the depletion and criticality calculations has been used to simulated nuclear reactors problems. Therefore, the goal is to analyze the sensibility of the fuel depletion of a PWR assembly using three different nuclear fuel depletion codes. The burnup calculations are performed using the codes MCNP5/ORIGEN2.1 (MONTEBURNS), KENO-VI/ORIGEN-S (TRITONSCALE6.0) and MCNPX (MCNPX/CINDER90). Each nuclear code performs the burnup using different depletion codes. Each depletion code works with collapsed energies from a master library in 1, 3 and 63 groups, respectively. Besides, each code uses different ways to obtain neutron flux that influences the depletions calculation. The results present a comparison of the neutronic parameters and isotopes composition such as criticality and nuclides build-up, the deviation in results are going to be assigned to features of the depletion code in use, such as the different radioactive decay internal libraries and the numerical method involved in solving the coupled differential depletion equations. It is also seen that the longer the period is and the more time steps are chosen, the larger the deviation become. (author)

  20. Development of CAP code for nuclear power plant containment: Lumped model

    International Nuclear Information System (INIS)

    Hong, Soon Joon; Choo, Yeon Joon; Hwang, Su Hyun; Lee, Byung Chul; Ha, Sang Jun

    2015-01-01

    Highlights: • State-of-art containment analysis code, CAP, has been developed. • CAP uses 3-field equations, water level oriented upwind scheme, local head model. • CAP has a function of linked calculation with reactor coolant system code. • CAP code assessments showed appropriate prediction capabilities. - Abstract: CAP (nuclear Containment Analysis Package) code has been developed in Korean nuclear society for the analysis of nuclear containment thermal hydraulic behaviors including pressure and temperature trends and hydrogen concentration. Lumped model of CAP code uses 2-phase, 3-field equations for fluid behaviors, and has appropriate constitutive equations, 1-dimensional heat conductor model, component models, trip and control models, and special process models. CAP can run in a standalone mode or a linked mode with a reactor coolant system analysis code. The linked mode enables the more realistic calculation of a containment response and is expected to be applicable to a more complicated advanced plant design calculation. CAP code assessments were carried out by gradual approaches: conceptual problems, fundamental phenomena, component and principal phenomena, experimental validation, and finally comparison with other code calculations on the base of important phenomena identifications. The assessments showed appropriate prediction capabilities of CAP

  1. Sensibility analysis of fuel depletion using different nuclear fuel depletion codes

    International Nuclear Information System (INIS)

    Martins, F.; Velasquez, C.E.; Castro, V.F.; Pereira, C.; Silva, C. A. Mello da

    2017-01-01

    Nowadays, the utilization of different nuclear codes to perform the depletion and criticality calculations has been used to simulated nuclear reactors problems. Therefore, the goal is to analyze the sensibility of the fuel depletion of a PWR assembly using three different nuclear fuel depletion codes. The burnup calculations are performed using the codes MCNP5/ORIGEN2.1 (MONTEBURNS), KENO-VI/ORIGEN-S (TRITONSCALE6.0) and MCNPX (MCNPX/CINDER90). Each nuclear code performs the burnup using different depletion codes. Each depletion code works with collapsed energies from a master library in 1, 3 and 63 groups, respectively. Besides, each code uses different ways to obtain neutron flux that influences the depletions calculation. The results present a comparison of the neutronic parameters and isotopes composition such as criticality and nuclides build-up, the deviation in results are going to be assigned to features of the depletion code in use, such as the different radioactive decay internal libraries and the numerical method involved in solving the coupled differential depletion equations. It is also seen that the longer the period is and the more time steps are chosen, the larger the deviation become. (author)

  2. Application of an atmospheric CFD code to wind resource assessment in complex terrain

    International Nuclear Information System (INIS)

    Laporte, Laurent

    2008-01-01

    This thesis is organized in two parts. The first part presents the use of the atmospheric CFD code Mercure Saturne to estimate the wind resource in complex terrain. A measurement campaign was led by EDF to obtain data for validation. A methodology was developed using meso-scale profiles as boundary conditions. Clustering of meteorological situations was used to reduce the number of simulations needed to calculate the wind resource. The validation of the code on the Askervein hill, the methodology and comparisons with measurements from the complex site are presented. The second part presents the modeling of wakes with the Mercure Saturne code. Forces, generated by the blades on the wind, are modeled by source terms, calculated by the BEM method. Two comparisons are proposed to validate the method: the first compares the numerical model with wind tunnel measurements from a small wind turbine, the second with measurements made on porous disks in an atmospheric boundary layer wind tunnel (author) [fr

  3. Comparison Results on the Basic Phenomena between CAP and CONTEMPT-LT

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Yeon Joon; Hong, Soon Joon; Hwang, Su Hyun; Kim, Min Ki; Lee, Byung Chul [FNC Tech., SNU, Seoul (Korea, Republic of); Ha, Sang Jun; Choi, Hoon [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    A development project for the domestic design code was launched to be used for the safety and performance analysis of pressurized light water reactors. As a part of this project, CAP (Containment Analysis Package) code has been developing for the containment safety and performance analysis side by side with SPACE. CAP Beta version has been released lately and validation processes are under way currently. Code by code comparison activity is scheduled in the validation processes and the first comparable code is CONTEMPT-LT. CONTEMPT-LT was developed to predict the longterm behavior of water-cooled nuclear reactor containment systems subjected to postulated loss-ofcoolant accident (LOCA) conditions. CONTEMPT-LT calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and energy exchange with adjacent compartments, leakage on containment response. Models are provided for fan cooler and cooling spray as engineered safety systems. Any compartment may have both a liquid pool region and an air-vapor atmosphere region above the pool. Each region is assumed to have a uniform temperature, but the temperatures of the two regions may be different. As mentioned above, CONTEMP-LT has the similar code features and it, therefore, is expected to show the similar analysis performance with CAP. In this study, the code performances were compared for the same phenomena between CAP and CONTEMPT-LT. Code comparison is carried out through two stages; separate and integral effect comparison

  4. Comparison Results on the Basic Phenomena between CAP and CONTEMPT-LT

    International Nuclear Information System (INIS)

    Choo, Yeon Joon; Hong, Soon Joon; Hwang, Su Hyun; Kim, Min Ki; Lee, Byung Chul; Ha, Sang Jun; Choi, Hoon

    2011-01-01

    A development project for the domestic design code was launched to be used for the safety and performance analysis of pressurized light water reactors. As a part of this project, CAP (Containment Analysis Package) code has been developing for the containment safety and performance analysis side by side with SPACE. CAP Beta version has been released lately and validation processes are under way currently. Code by code comparison activity is scheduled in the validation processes and the first comparable code is CONTEMPT-LT. CONTEMPT-LT was developed to predict the longterm behavior of water-cooled nuclear reactor containment systems subjected to postulated loss-ofcoolant accident (LOCA) conditions. CONTEMPT-LT calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and energy exchange with adjacent compartments, leakage on containment response. Models are provided for fan cooler and cooling spray as engineered safety systems. Any compartment may have both a liquid pool region and an air-vapor atmosphere region above the pool. Each region is assumed to have a uniform temperature, but the temperatures of the two regions may be different. As mentioned above, CONTEMP-LT has the similar code features and it, therefore, is expected to show the similar analysis performance with CAP. In this study, the code performances were compared for the same phenomena between CAP and CONTEMPT-LT. Code comparison is carried out through two stages; separate and integral effect comparison

  5. Comparison of the General Electric BWR/6 standard plant design to the IAEA NUSS codes and guides

    International Nuclear Information System (INIS)

    D'Ardenne, W.H.; Sherwood, G.G.

    1985-01-01

    The General Electric BWR/6 Mark III standard plant design meets or exceeds current requirements of published International Atomic Energy Agency (IAEA) Nuclear Safety Standards (NUSS) codes and guides. This conclusion is based on a review of the NUSS codes and guides by General Electric and by the co-ordinated US review of the NUSS codes and guides during their development. General Electric compared the published IAEA NUSS codes and guides with the General Electric design. The applicability of each code and guide to the BWR/6 Mark III standard plant design was determined. Each code or guide was reviewed by a General Electric engineer knowledgeable about the structures, systems and components addressed and the technical area covered by that code or guide. The results of this review show that the BWR/6 Mark III standard plant design meets or exceeds the applicable requirements of the published IAEA NUSS codes and guides. The co-ordinated US review of the IAEA NUSS codes and guides corroborates the General Electric review. In the co-ordinated US review, the USNRC and US industry organizations (including General Electric) review the NUSS codes and guides during their development. This review ensures that the NUSS codes and guides are consistent with the current US government regulations, guidance and regulatory practices, US voluntary industry codes and standards, and accepted US industry design, construction and operational practices. If any inconsistencies are identified, comments are submitted to the IAEA by the USNRC. All US concerns submitted to the IAEA have been resolved. General Electric design reviews and the Final Design Approval (FDA) issued by the USNRC have verified that the General Electric BWR/6 Mark III standard plant design meets or exceeds the current US requirements, guidance and practices. Since these requirements, guidance and practices meet or exceed those of the NUSS codes and guides, so does the General Electric design. (author)

  6. Field-based tests of geochemical modeling codes: New Zealand hydrothermal systems

    International Nuclear Information System (INIS)

    Bruton, C.J.; Glassley, W.E.; Bourcier, W.L.

    1993-12-01

    Hydrothermal systems in the Taupo Volcanic Zone, North Island, New Zealand are being used as field-based modeling exercises for the EQ3/6 geochemical modeling code package. Comparisons of the observed state and evolution of the hydrothermal systems with predictions of fluid-solid equilibria made using geochemical modeling codes will determine how the codes can be used to predict the chemical and mineralogical response of the environment to nuclear waste emplacement. Field-based exercises allow us to test the models on time scales unattainable in the laboratory. Preliminary predictions of mineral assemblages in equilibrium with fluids sampled from wells in the Wairakei and Kawerau geothermal field suggest that affinity-temperature diagrams must be used in conjunction with EQ6 to minimize the effect of uncertainties in thermodynamic and kinetic data on code predictions

  7. WWER-440 fuel rod performance analysis with PIN-Micro and TRANSURANUS codes

    International Nuclear Information System (INIS)

    Vitkova, M.; Manolova, M.; Stefanova, S.; Simeonova, V.; Passage, G.; Lassmann, K.

    1994-01-01

    PIN-micro and TRANSURANUS codes were used to analyse the WWER-440 fuel rod behaviour at normal operation conditions. Two highest loaded fuel rods of the fuel assemblies irradiated in WWER-440 with different power histories were selected. A set of the most probable average values of all geometrical and technological parameters were used. A comparison between PIN-micro and TRANSURANUS codes was performed using identical input data. The results for inner gas pressure, gap size, local linear heat rate, fuel central temperature and fission gas release as a function of time calculated for the selected fuel rods are presented. The following conclusions were drawn: 1) The PIN-micro code predicts adequately the thermal and mechanical behaviour of the two fuel rods; 2) The comparison of the results obtained by PIN-micro and TRANSURANUS shows a reasonable agreement and the discrepancies could be explained by the lack of thoroughly WWER oriented verification of TRANSURANUS; 3) The advanced TRANSURANUS code could be successfully applied for WWER fuel rod thermal and mechanical analysis after incorporation of all necessary WWER specific material properties and models for the Zr+1%Nb cladding, for the fuel rod as a whole and after validation against WWER experimental and operational data. 1 tab., 10 figs., 10 refs

  8. Assessment of the system code DRUFAN/ATHLET using results of LOBI tests

    International Nuclear Information System (INIS)

    Burwell, J.M.; Kirmse, R.E.; Kyncl, M.; Malhotra, P.K.

    1989-09-01

    Four post-test analyses have been performed by GRS within the Shared Cost Action Programme (SCAP) sponsored by the Commission of the European Communities (contract 3015-86-07 EL ISP D) and by the Bundesminister fuer Forschung und Technologie of the Federal Republic of Germany (Research project RS 739). The four tests were mutually selected by the contractors (CEA, GRS, IKE, Univ. Pisa) of activity No. 3 and by the project organizer. Some of the tests were selected to be analyzed by more than one participant in order to allow comparison between analytical results obtained with different codes or obtained by different code-users. DRUFAN/ATHLET verification analyses were performed by IKE too. The four tests selected for the GRS activity are: - A2-77A (Natural Circulation Test), Analysis with ATHLET - A1-76 (Steam Generator Performance Test), Analysis with DRUFAN - BL-01 (Intermediate Leak), Analysis with ATHLET - A2-81 (Small Leak), Analysis with ATHLET. This final report contains the results of the four post test analysis including the comparison between measured and calculated quantities and the description of the applied codes, the selected model of the LOBI facility and the conclusions drawn for the improvement of the codes models

  9. WWER-440 fuel rod performance analysis with PIN-Micro and TRANSURANUS codes

    Energy Technology Data Exchange (ETDEWEB)

    Vitkova, M; Manolova, M; Stefanova, S; Simeonova, V; Passage, G [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika; Kharalampieva, Ts [Kombinat Atomna Energetika, Kozloduj (Bulgaria); Lassmann, K [European Atomic Energy Community, Karlsruhe (Germany). European Inst. for Transuranium Elements

    1994-12-31

    PIN-micro and TRANSURANUS codes were used to analyse the WWER-440 fuel rod behaviour at normal operation conditions. Two highest loaded fuel rods of the fuel assemblies irradiated in WWER-440 with different power histories were selected. A set of the most probable average values of all geometrical and technological parameters were used. A comparison between PIN-micro and TRANSURANUS codes was performed using identical input data. The results for inner gas pressure, gap size, local linear heat rate, fuel central temperature and fission gas release as a function of time calculated for the selected fuel rods are presented. The following conclusions were drawn: (1) The PIN-micro code predicts adequately the thermal and mechanical behaviour of the two fuel rods; (2) The comparison of the results obtained by PIN-micro and TRANSURANUS shows a reasonable agreement and the discrepancies could be explained by the lack of thoroughly WWER oriented verification of TRANSURANUS; (3) The advanced TRANSURANUS code could be successfully applied for WWER fuel rod thermal and mechanical analysis after incorporation of all necessary WWER specific material properties and models for the Zr+1%Nb cladding, for the fuel rod as a whole and after validation against WWER experimental and operational data. 1 tab., 10 figs., 10 refs.

  10. Analysis of the SPERT III E-core experiment using the EUREKA-2 code

    International Nuclear Information System (INIS)

    Harami, Taikan; Uemura, Mutsumi; Ohnishi, Nobuaki

    1986-09-01

    EUREKA-2, a coupled nuclear thermal hydrodynamic kinetic code, was adapted for the testing of models and methods. Code evaluations were made with the reactivity addition experiments of the SPERT III E-Core, a slightly enriched oxide core. The code was tested for non damaging power excursions including a wide range of initial operating conditions, such as cold-startup, hot-startup, hot-standby and operating-power initial conditions. Comparisons resulted in a good agreement within the experimental errors between calculated and experimental power, energy, reactivity and clad surface temperature. (author)

  11. Evaluation Codes from an Affine Veriety Code Perspective

    DEFF Research Database (Denmark)

    Geil, Hans Olav

    2008-01-01

    Evaluation codes (also called order domain codes) are traditionally introduced as generalized one-point geometric Goppa codes. In the present paper we will give a new point of view on evaluation codes by introducing them instead as particular nice examples of affine variety codes. Our study...... includes a reformulation of the usual methods to estimate the minimum distances of evaluation codes into the setting of affine variety codes. Finally we describe the connection to the theory of one-pointgeometric Goppa codes. Contents 4.1 Introduction...... . . . . . . . . . . . . . . . . . . . . . . . 171 4.9 Codes form order domains . . . . . . . . . . . . . . . . . . . . . . . . . . . . 173 4.10 One-point geometric Goppa codes . . . . . . . . . . . . . . . . . . . . . . . . 176 4.11 Bibliographical Notes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 178 References...

  12. Hanford Dose Overview Program. Comparison of AIRDOS-EPA and Hanford site dose codes

    International Nuclear Information System (INIS)

    Aaberg, R.L.; Napier, B.A.

    1985-11-01

    Radiation dose commitments for persons in the Hanford environs calculated using AIRDOS-EPA were compared with those calculated using a suite of Hanford codes: FOOD, PABLM, DACRIN, and KRONIC. Dose commitments to the population and to the maximally exposed individual (MI) based on annual releases of eight radionuclides from the N-Reactor, were calculated by these codes. Dose commitments from each pathway to the total body, lung, thyroid, and lower large intestine (LLI) are given for the population and MI, respectively. 11 refs., 25 tabs

  13. Dynamic Load Balancing for PIC code using Eulerian/Lagrangian partitioning

    OpenAIRE

    Sauget, Marc; Latu, Guillaume

    2017-01-01

    This document presents an analysis of different load balance strategies for a Plasma physics code that models high energy particle beams with PIC method. A comparison of different load balancing algorithms is given: static or dynamic ones. Lagrangian and Eulerian partitioning techniques have been investigated.

  14. 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.

  15. 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

  16. CAPRICORN subchannel code for sodium boiling in LMFBR fuel bundles

    International Nuclear Information System (INIS)

    Padilla, A. Jr.; Smith, D.E.; O'Dell, L.D.

    1983-01-01

    The CAPRICORN computer code analyzes steady-state and transient, single-phase and boiling problems in LMFBR fuel bundles. CAPRICORN uses the same type of subchannel geometry as the COBRA family of codes and solves a similar system of conservation equations for mass, momentum, and energy. However, CAPRICORN uses a different numerical solution method which allows it to handle the full liquid-to-vapor density change for sodium boiling. Results of the initial comparison with data (the W-1 SLSF pipe rupture experiment) are very promising and provide an optimistic basis for proceeding with further development

  17. ANTEO: An optimised PC computer code for the steady state thermal hydraulic analysis of rod bundles

    International Nuclear Information System (INIS)

    Cevolani, S.

    1996-07-01

    The paper deals with the description of a Personal Computer oriented subchannel code, devoted to the steady state thermal hydraulic analysis of nuclear reactor fuel bundles. The development of a such code was made possible by two facts: first, the increase the computing power of the desk machines; secondly, the fact several years of experience into operate subchannels codes have shown how to simplify many of the physical models without a sensible loss of accuracy. For sake of validation, the developed code was compared with a traditional subchannel code, the COBRA one. The results of the comparison show a very good agreement between the two codes

  18. Recent improvements to TRIGLAV code

    International Nuclear Information System (INIS)

    Zagar, T.; Ravnik, M.; Persic, A.

    1998-01-01

    TRIGLAV code was developed for TRIGA research reactor calculations and is based on two-dimensional diffusion equation. The main purpose of the program is calculation of the fuel elements burn-up. Calculated core burn-up and excess reactivity results are compared with experimental values. New control rod model is introduced and tested in this paper. Calculated integral control rod worth and calculated integral reactivity curves are presented and compared with measured values. Comparison with measured fuel element worth values is presented as a test for two-dimensional flux distribution calculations.(author)

  19. User Effect on Code Application and Qualification Needs

    International Nuclear Information System (INIS)

    D'Auria, F.; Salah, A.B.

    2008-01-01

    Experience with some code assessment case studies and also additional ISPs have shown the dominant effect of the code user on the predicted system behavior. The general findings of the user effect investigations on some of the case studies indicate, specifically, that in addition to user effects, there are other reasons which affect the results of the calculations and are hidden under the general title of user effects. The specific characteristics of experimental facilities, i.e. limitations as far as code assessment is concerned; limitations of the used thermal-hydraulic codes to simulate certain system behavior or phenomena; limitations due to interpretation of experimental data by the code user, i.e. interpretation of experimental data base. On the basis of the discussions in this paper, the following conclusions and recommendations can be made: More dialogue appears to be necessary with the experimenters in the planning of code assessment calculations, e.g. ISPs.; User guidelines are not complete for the codes and the lack of sufficient and detailed user guidelines are observed with some of the case studies; More extensive user instruction and training, improved user guidelines, or quality assurance procedures may partially reduce some of the subjective user influence on the calculated results; The discrepancies between experimental data and code predictions are due both to the intrinsic code limit and to the so called user effects. There is a worthful need to quantify the percentage of disagreement due to the poor utilization of the code and due to the code itself. This need especially arises for the uncertainty evaluation studies (e.g. [18]) which do not take into account the mentioned user effects; A much focused investigation, based on the results of comparison calculations e.g. ISPs, analyzing the experimental data and the results of the specific code in order to evaluate the user effects and the related experimental aspects should be integral part of the

  20. NOVA: a nonvariational code for solving MHD stability of axisymmetric toroidal plasmas

    International Nuclear Information System (INIS)

    Cheng, C.Z.; Chance, M.S.

    1986-04-01

    A nonvariational approach for determining the ideal MHD stability of axisymmetric toroidal confinement systems is presented. The code (NOVA) employs cubic B-spline finite elements and Fourier expansion in a general flux coordinate (psi, theta, zeta) system. Better accuracy and faster convergence were obtained in comparison with the variational PEST and ERATO codes. The nonvariational approach can be extended to problems having non-Hermitian eigenmode equations where variational energy principles cannot be obtained

  1. Benchmark studies of the gyro-Landau-fluid code and gyro-kinetic codes on kinetic ballooning modes

    Energy Technology Data Exchange (ETDEWEB)

    Tang, T. F. [Dalian University of Technology, Dalian 116024 (China); Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Xu, X. Q. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Ma, C. H. [Fusion Simulation Center, School of Physics, Peking University, Beijing (China); Bass, E. M.; Candy, J. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); Holland, C. [University of California San Diego, La Jolla, California 92093-0429 (United States)

    2016-03-15

    A Gyro-Landau-Fluid (GLF) 3 + 1 model has been recently implemented in BOUT++ framework, which contains full Finite-Larmor-Radius effects, Landau damping, and toroidal resonance [Ma et al., Phys. Plasmas 22, 055903 (2015)]. A linear global beta scan has been conducted using the JET-like circular equilibria (cbm18 series), showing that the unstable modes are kinetic ballooning modes (KBMs). In this work, we use the GYRO code, which is a gyrokinetic continuum code widely used for simulation of the plasma microturbulence, to benchmark with GLF 3 + 1 code on KBMs. To verify our code on the KBM case, we first perform the beta scan based on “Cyclone base case parameter set.” We find that the growth rate is almost the same for two codes, and the KBM mode is further destabilized as beta increases. For JET-like global circular equilibria, as the modes localize in peak pressure gradient region, a linear local beta scan using the same set of equilibria has been performed at this position for comparison. With the drift kinetic electron module in the GYRO code by including small electron-electron collision to damp electron modes, GYRO generated mode structures and parity suggest that they are kinetic ballooning modes, and the growth rate is comparable to the GLF results. However, a radial scan of the pedestal for a particular set of cbm18 equilibria, using GYRO code, shows different trends for the low-n and high-n modes. The low-n modes show that the linear growth rate peaks at peak pressure gradient position as GLF results. However, for high-n modes, the growth rate of the most unstable mode shifts outward to the bottom of pedestal and the real frequency of what was originally the KBMs in ion diamagnetic drift direction steadily approaches and crosses over to the electron diamagnetic drift direction.

  2. The Protection of the Right to Life in the New Criminal Code

    Directory of Open Access Journals (Sweden)

    Lavinia Mihaela Vladila

    2011-05-01

    Full Text Available The present article illustrates general aspects of the actual regulation on the protection of the rightto life, in comparison with the new regulation, which shall be analyzed more carefully. The paper is based onthe study of the new Criminal Code, emphasizing the most important differences between the actualregulation and the new Criminal Code, and the few texts elaborated in this area. The approach of the subjectis a more practical one, because few texts were written about the new Criminal Code, at the moment of itsentrance into force, the doctrinaires will need to know the differences and the innovations brought by it. Theresult is meant to increase the understanding of the new text and to enrich the analysis and synthesis of thenew Criminal Code.

  3. A point-kernel shielding code for calculations of neutron and secondary gamma-ray 1cm dose equivalents: PKN

    International Nuclear Information System (INIS)

    Kotegawa, Hiroshi; Tanaka, Shun-ichi

    1991-09-01

    A point-kernel integral technique code, PKN, and the related data library have been developed to calculate neutron and secondary gamma-ray dose equivalents in water, concrete and iron shields for neutron sources in 3-dimensional geometry. The comparison between calculational results of the present code and those of the 1-dimensional transport code ANISN = JR, and the 2-dimensional transport code DOT4.2 showed a sufficient accuracy, and the availability of the PKN code has been confirmed. (author)

  4. An Optimal Linear Coding for Index Coding Problem

    OpenAIRE

    Pezeshkpour, Pouya

    2015-01-01

    An optimal linear coding solution for index coding problem is established. Instead of network coding approach by focus on graph theoric and algebraic methods a linear coding program for solving both unicast and groupcast index coding problem is presented. The coding is proved to be the optimal solution from the linear perspective and can be easily utilize for any number of messages. The importance of this work is lying mostly on the usage of the presented coding in the groupcast index coding ...

  5. Comparison of the PLCS malfunction event analyses by using the CESEC III and MEDUSA codes

    International Nuclear Information System (INIS)

    Oh, Myung Taek; Park, Chan Eok; Kim, Shin Whan; Seo, Jong Tae

    2004-01-01

    The MEDUSA is a code developed by Korea Power Engineering Company (KOPEC) for non-LOCA and LOCA analysis, providing two-fluid, three-field representation of two-phase flow. In this paper, as an effort to verify the MEDUSA code, comparative simulation for Pressurizer Level Control System (PLCS) malfunction with loss of off-site power is performed for UCN 3 and 4 plants. The results by the MEDUSA are compared with those calculated by the CESEC-III, a licensing analysis code used for Korean Standard Nuclear Power Plant (KSNP)

  6. A multiobjective approach to the genetic code adaptability problem.

    Science.gov (United States)

    de Oliveira, Lariza Laura; de Oliveira, Paulo S L; Tinós, Renato

    2015-02-19

    The organization of the canonical code has intrigued researches since it was first described. If we consider all codes mapping the 64 codes into 20 amino acids and one stop codon, there are more than 1.51×10(84) possible genetic codes. The main question related to the organization of the genetic code is why exactly the canonical code was selected among this huge number of possible genetic codes. Many researchers argue that the organization of the canonical code is a product of natural selection and that the code's robustness against mutations would support this hypothesis. In order to investigate the natural selection hypothesis, some researches employ optimization algorithms to identify regions of the genetic code space where best codes, according to a given evaluation function, can be found (engineering approach). The optimization process uses only one objective to evaluate the codes, generally based on the robustness for an amino acid property. Only one objective is also employed in the statistical approach for the comparison of the canonical code with random codes. We propose a multiobjective approach where two or more objectives are considered simultaneously to evaluate the genetic codes. In order to test our hypothesis that the multiobjective approach is useful for the analysis of the genetic code adaptability, we implemented a multiobjective optimization algorithm where two objectives are simultaneously optimized. Using as objectives the robustness against mutation with the amino acids properties polar requirement (objective 1) and robustness with respect to hydropathy index or molecular volume (objective 2), we found solutions closer to the canonical genetic code in terms of robustness, when compared with the results using only one objective reported by other authors. Using more objectives, more optimal solutions are obtained and, as a consequence, more information can be used to investigate the adaptability of the genetic code. The multiobjective approach

  7. Comparison of simulation with experiment in an RFQ

    International Nuclear Information System (INIS)

    Boicourt, G.P.; Sander, O.R.; Wangler, T.P.

    1985-01-01

    The accelerator test stand (ATS) RFQ has provided an opportunity to compare the predictions of the RFQ beam-dynamics code PARMTEQ with actual operation of an RFQ. For this comparison, the code was adapted to simulate the measured operation parameters, which are somewhat different from those of the ideal design. A Monte Carlo code was written to provide input to PARMTEQ, based on measured input beam distributions. With these refinements, the code has given results that are in good agreement with measurements and has provided information leading to an explanation of an unexpected set of measurements. This paper describes the method used to generate a pseudo particle beam based on the measured transverse properties of the RFQ input beam and describes some of the comparisons between simulation and experiment. An explanation is provided for the energy-spectrum structure observed in the RFQ output beam during low-voltage operation. 3 refs., 7 figs

  8. Development of Visual CINDER Code with Visual C⧣.NET

    International Nuclear Information System (INIS)

    Kim, Oyeon

    2016-01-01

    CINDER code, CINDER' 90 or CINDER2008 that is integrated with the Monte Carlo code, MCNPX, is widely used to calculate the inventory of nuclides in irradiated materials. The MCNPX code provides decay processes to the particle transport scheme that traditionally only covered prompt processes. The integration schemes serve not only the reactor community (MCNPX burnup) but also the accelerator community as well (residual production information). The big benefit for providing these options lies in the easy cross comparison of the transmutation codes since the calculations are based on exactly the same material, neutron flux and isotope production/destruction inputs. However, it is just frustratingly cumbersome to use. In addition, multiple human interventions may increase the possibility of making errors. The number of significant digits in the input data varies in steps, which may cause big errors for highly nonlinear problems. Thus, it is worthwhile to find a new way to wrap all the codes and procedures in one consistent package which can provide ease of use. The visual CINDER code development is underway with visual C .NET framework. It provides a few benefits for the atomic transmutation simulation with CINDER code. A few interesting and useful properties of visual C .NET framework are introduced. We also showed that the wrapper could make the simulation accurate for highly nonlinear transmutation problems and also increase the possibility of direct combination a radiation transport code MCNPX with CINDER code. Direct combination of CINDER with MCNPX in a wrapper will provide more functionalities for the radiation shielding and prevention study

  9. Development of Visual CINDER Code with Visual C⧣.NET

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Oyeon [Institute for Modeling and Simulation Convergence, Daegu (Korea, Republic of)

    2016-10-15

    CINDER code, CINDER' 90 or CINDER2008 that is integrated with the Monte Carlo code, MCNPX, is widely used to calculate the inventory of nuclides in irradiated materials. The MCNPX code provides decay processes to the particle transport scheme that traditionally only covered prompt processes. The integration schemes serve not only the reactor community (MCNPX burnup) but also the accelerator community as well (residual production information). The big benefit for providing these options lies in the easy cross comparison of the transmutation codes since the calculations are based on exactly the same material, neutron flux and isotope production/destruction inputs. However, it is just frustratingly cumbersome to use. In addition, multiple human interventions may increase the possibility of making errors. The number of significant digits in the input data varies in steps, which may cause big errors for highly nonlinear problems. Thus, it is worthwhile to find a new way to wrap all the codes and procedures in one consistent package which can provide ease of use. The visual CINDER code development is underway with visual C .NET framework. It provides a few benefits for the atomic transmutation simulation with CINDER code. A few interesting and useful properties of visual C .NET framework are introduced. We also showed that the wrapper could make the simulation accurate for highly nonlinear transmutation problems and also increase the possibility of direct combination a radiation transport code MCNPX with CINDER code. Direct combination of CINDER with MCNPX in a wrapper will provide more functionalities for the radiation shielding and prevention study.

  10. Comparison of HPGe detector response data for low energy photons using MCNP, EGS, and its codes

    International Nuclear Information System (INIS)

    Kim, Soon Young; Kim, Jong Kyung

    1995-01-01

    In this study, the photopeak efficiency, K α and K β escape fractions of HPGe detector(100mm 2 X 10mm) are calculated and tabulated as a function of incident X-ray energies from 12 to 60keV in 2-keV increments. Compton, elastic, and penetration fractions are not tabulated from this work since they are negligible amounts in this energy range. The results calculated from this work are compared with earlier Monte Carlo results which had been carried out by Chin-Tu Chen et al.. From the comparison, it is found that the results calculated from each code show a large difference when the incident photon energy approaches to 12keV as compared with energy ranges from 50 to 60keV. In X-ray dosimetry and diagnostic radiology, it is essential to have accurate knowledge of X-ray spectra for studies of patient dose and image quality. Being X-ray spectra measured with a detection system, some distortions due to the incomplete absorption of primary photon or escape before interacting with the detector which have finite dimension can take place

  11. The in-core fuel management code system for VVER reactors

    International Nuclear Information System (INIS)

    Cada, R.; Krysl, V.; Mikolas, P.; Sustek, J.; Svarny, J.

    2004-01-01

    The structure and methodology of a fuel management system for NPP VVER 1000 (NPP Temelin) and VVER 440 (NPP Dukovany) is described. It is under development in SKODA JS a.s. and is followed by practical applications. The general objectives of the system are maximization of end of cycle reactivity, the minimization of fresh fuel inventory for the minimization of fed enrichment and minimization of burnable poisons (BPs) inventory. They are also safety related constraints in witch minimization of power peaking plays a dominant role. General structure of the system consists in preparation of input data for macrocode calculation, algorithms (codes) for optimization of fuel loading, calculation of fuel enrichment and BPs assignment. At present core loading can be calculated (optimized) by Tabu search algorithm (code ATHENA), genetic algorithm (code Gen1) and hybrid algorithm - simplex procedure with application of Tabu search algorithm on binary shuffling (code OPAL B ). Enrichment search is realized by the application of simplex algorithm (OPAL B code) and BPs assignment by module BPASS and simplex algorithm in OPAL B code. Calculations of the real core loadings are presented and a comparison of different optimization methods is provided. (author)

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  13. 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)

  14. Use of the Spine Adverse Events Severity System (SAVES) in patients with traumatic spinal cord injury. A comparison with institutional ICD-10 coding for the identification of acute care adverse events.

    Science.gov (United States)

    Street, J T; Thorogood, N P; Cheung, A; Noonan, V K; Chen, J; Fisher, C G; Dvorak, M F

    2013-06-01

    Observational cohort comparison. To compare the previously validated Spine Adverse Events Severity system (SAVES) with International Classification of Diseases, Tenth Revision codes (ICD-10) codes for identifying adverse events (AEs) in patients with traumatic spinal cord injury (TSCI). Quaternary Care Spine Program. Patients discharged between 2006 and 2010 were identified from our prospective registry. Two consecutive cohorts were created based on the system used to record acute care AEs; one used ICD-10 coding by hospital coders and the other used SAVES data prospectively collected by a multidisciplinary clinical team. The ICD-10 codes were appropriately mapped to the SAVES. There were 212 patients in the ICD-10 cohort and 173 patients in the SAVES cohort. Analyses were adjusted to account for the different sample sizes, and the two cohorts were comparable based on age, gender and motor score. The SAVES system identified twice as many AEs per person as ICD-10 coding. Fifteen unique AEs were more reliably identified using SAVES, including neuropathic pain (32 × more; Ppatient age and severity of paralysis were more reliably correlated to AEs collected through SAVES than ICD-10. Implementation of the SAVES system for patients with TSCI captured more individuals experiencing AEs and more AEs per person compared with ICD-10 codes. This study demonstrates the utility of prospectively collecting AE data using validated tools.

  15. Improving the coding and classification of ambulance data through the application of International Classification of Disease 10th revision.

    Science.gov (United States)

    Cantwell, Kate; Morgans, Amee; Smith, Karen; Livingston, Michael; Dietze, Paul

    2014-02-01

    This paper aims to examine whether an adaptation of the International Classification of Disease (ICD) coding system can be applied retrospectively to final paramedic assessment data in an ambulance dataset with a view to developing more fine-grained, clinically relevant case definitions than are available through point-of-call data. Over 1.2 million case records were extracted from the Ambulance Victoria data warehouse. Data fields included dispatch code, cause (CN) and final primary assessment (FPA). Each FPA was converted to an ICD-10-AM code using word matching or best fit. ICD-10-AM codes were then converted into Major Diagnostic Categories (MDC). CN was aligned with the ICD-10-AM codes for external cause of morbidity and mortality. The most accurate results were obtained when ICD-10-AM codes were assigned using information from both FPA and CN. Comparison of cases coded as unconscious at point-of-call with the associated paramedic assessment highlighted the extra clinical detail obtained when paramedic assessment data are used. Ambulance paramedic assessment data can be aligned with ICD-10-AM and MDC with relative ease, allowing retrospective coding of large datasets. Coding of ambulance data using ICD-10-AM allows for comparison of not only ambulance service users but also with other population groups. WHAT IS KNOWN ABOUT THE TOPIC? There is no reliable and standard coding and categorising system for paramedic assessment data contained in ambulance service databases. WHAT DOES THIS PAPER ADD? This study demonstrates that ambulance paramedic assessment data can be aligned with ICD-10-AM and MDC with relative ease, allowing retrospective coding of large datasets. Representation of ambulance case types using ICD-10-AM-coded information obtained after paramedic assessment is more fine grained and clinically relevant than point-of-call data, which uses caller information before ambulance attendance. WHAT ARE THE IMPLICATIONS FOR PRACTITIONERS? This paper describes

  16. 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.

  17. LDPC Codes with Minimum Distance Proportional to Block Size

    Science.gov (United States)

    Divsalar, Dariush; Jones, Christopher; Dolinar, Samuel; Thorpe, Jeremy

    2009-01-01

    error floors as well as low decoding thresholds. As an example, the illustration shows the protograph (which represents the blueprint for overall construction) of one proposed code family for code rates greater than or equal to 1.2. Any size LDPC code can be obtained by copying the protograph structure N times, then permuting the edges. The illustration also provides Field Programmable Gate Array (FPGA) hardware performance simulations for this code family. In addition, the illustration provides minimum signal-to-noise ratios (Eb/No) in decibels (decoding thresholds) to achieve zero error rates as the code block size goes to infinity for various code rates. In comparison with the codes mentioned in the preceding article, these codes have slightly higher decoding thresholds.

  18. Development of a multi-grid FDTD code for three-dimensional simulation of large microwave sintering experiments

    Energy Technology Data Exchange (ETDEWEB)

    White, M.J.; Iskander, M.F. [Univ. of Utah, Salt Lake City, UT (United States). Electrical Engineering Dept.; Kimrey, H.D. [Oak Ridge National Lab., TN (United States)

    1996-12-31

    The Finite-Difference Time-Domain (FDTD) code available at the University of Utah has been used to simulate sintering of ceramics in single and multimode cavities, and many useful results have been reported in literature. More detailed and accurate results, specifically around and including the ceramic sample, are often desired to help evaluate the adequacy of the heating procedure. In electrically large multimode cavities, however, computer memory requirements limit the number of the mathematical cells, and the desired resolution is impractical to achieve due to limited computer resources. Therefore, an FDTD algorithm which incorporates multiple-grid regions with variable-grid sizes is required to adequately perform the desired simulations. In this paper the authors describe the development of a three-dimensional multi-grid FDTD code to help focus a large number of cells around the desired region. Test geometries were solved using a uniform-grid and the developed multi-grid code to help validate the results from the developed code. Results from these comparisons, as well as the results of comparisons between the developed FDTD code and other available variable-grid codes are presented. In addition, results from the simulation of realistic microwave sintering experiments showed improved resolution in critical sites inside the three-dimensional sintering cavity. With the validation of the FDTD code, simulations were performed for electrically large, multimode, microwave sintering cavities to fully demonstrate the advantages of the developed multi-grid FDTD code.

  19. Comparative simulation of Stirling and Sibling cycle cryocoolers with two codes

    International Nuclear Information System (INIS)

    Mitchell, M.P.; Wilson, K.J.; Bauwens, L.

    1989-01-01

    The authors present a comparative analysis of Stirling and Sibling Cycle cryocoolers conducted with two different computer simulation codes. One code (CRYOWEISS) performs an initial analysis on the assumption of isothermal conditions in the machines and adjusts that result with decoupled loss calculations. The other code (MS*2) models fluid flows and heat transfers more realistically but ignores significant loss mechanisms, including flow friction and heat conduction through the metal of the machines. Surprisingly, MS*2 is less optimistic about performance of all machines even though it ignores losses that are modelled by CRYOWEISS. Comparison between constant-bore Stirling and Sibling machines shows that their performance is generally comparable over a range of temperatures, pressures and operating speeds. No machine was consistently superior or inferior according to both codes over the whole range of conditions studied

  20. Computer codes incorporating pre-equilibrium decay

    International Nuclear Information System (INIS)

    Prince, A.

    1980-01-01

    After establishing the need to describe the high-energy particle spectrum which is evident in the experimental data, the various models used in the interpretation are presented. This includes the following: a) Cascade Model; b) Fermi-Gas Relaxation Model; c) Exciton Model; d) Hybrid and Geometry-Dependent Model. The codes description and preparation of input data for STAPRE was presented (Dr. Strohmaier). A simulated output was employed for a given input and comparison with experimental data substantiated the rather sophisticated treatment. (author)

  1. 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)

  2. MCOR - Monte Carlo depletion code for reference LWR calculations

    Energy Technology Data Exchange (ETDEWEB)

    Puente Espel, Federico, E-mail: fup104@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Tippayakul, Chanatip, E-mail: cut110@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Ivanov, Kostadin, E-mail: kni1@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Misu, Stefan, E-mail: Stefan.Misu@areva.com [AREVA, AREVA NP GmbH, Erlangen (Germany)

    2011-04-15

    , several interesting studies performed with MCOR are explained; its validation against MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN; sensitivity studies with different voids and cross-section libraries; detailed comparison against measurement. Finally, the MCOR is used as a reference tool for benchmarking deterministic codes on the example of qualification of the spectral codes (APOLLO2-A and CASMO-4).

  3. MCOR - Monte Carlo depletion code for reference LWR calculations

    International Nuclear Information System (INIS)

    Puente Espel, Federico; Tippayakul, Chanatip; Ivanov, Kostadin; Misu, Stefan

    2011-01-01

    . Additionally, several interesting studies performed with MCOR are explained; its validation against MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN; sensitivity studies with different voids and cross-section libraries; detailed comparison against measurement. Finally, the MCOR is used as a reference tool for benchmarking deterministic codes on the example of qualification of the spectral codes (APOLLO2-A and CASMO-4).

  4. The VENUS-7 benchmarks. Results from state-of-the-art transport codes and nuclear data

    International Nuclear Information System (INIS)

    Zwermann, Winfried; Pautz, Andreas; Timm, Wolf

    2010-01-01

    For the validation of both nuclear data and computational methods, comparisons with experimental data are necessary. Most advantageous are assemblies where not only the multiplication factors or critical parameters were measured, but also additional quantities like reactivity differences or pin-wise fission rate distributions have been assessed. Currently there is a comprehensive activity to evaluate such measure-ments and incorporate them in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. A large number of such experiments was performed at the VENUS zero power reactor at SCK/CEN in Belgium in the sixties and seventies. The VENUS-7 series was specified as an international benchmark within the OECD/NEA Working Party on Scientific Issues of Reactor Systems (WPRS), and results obtained with various codes and nuclear data evaluations were summarized. In the present paper, results of high-accuracy transport codes with full spatial resolution with up-to-date nuclear data libraries from the JEFF and ENDF/B evaluations are presented. The comparisons of the results, both code-to-code and with the measured data are augmented by uncertainty and sensitivity analyses with respect to nuclear data uncertainties. For the multiplication factors, these are performed with the TSUNAMI-3D code from the SCALE system. In addition, uncertainties in the reactivity differences are analyzed with the TSAR code which is available from the current SCALE-6 version. (orig.)

  5. Application of the subchannel analysis code COBRA III C for liquid sodium

    International Nuclear Information System (INIS)

    Nissen, K.L.

    1981-01-01

    The subchannel-analysis code COBRA III C was developed to gain knowledge of mass flow and temperature distribution in rod bundles of light water reactors. A comparison of experimental results for the temperature distribution in a 19 rod bundle with calculations done by the computer program shows the capability of COBRA III C to handle liquid sodium cooling. The code needs sodium properties as well as changed correlations for turbulent mixing and heat transfer at the rod. (orig.) [de

  6. The implementation of a toroidal limiter model into the gyrokinetic code ELMFIRE

    Energy Technology Data Exchange (ETDEWEB)

    Leerink, S.; Janhunen, S.J.; Kiviniemi, T.P.; Nora, M. [Euratom-Tekes Association, Helsinki University of Technology (Finland); Heikkinen, J.A. [Euratom-Tekes Association, VTT, P.O. Box 1000, FI-02044 VTT (Finland); Ogando, F. [Universidad Nacional de Educacion a Distancia, Madrid (Spain)

    2008-03-15

    The ELMFIRE full nonlinear gyrokinetic simulation code has been developed for calculations of plasma evolution and dynamics of turbulence in tokamak geometry. The code is applicable for calculations of strong perturbations in particle distribution function, rapid transients and steep gradients in plasma. Benchmarking against experimental reflectometry data from the FT2 tokamak is being discussed and in this paper a model for comparison and studying poloidal velocity is presented. To make the ELMFIRE code suitable for scrape-off layer simulations a simplified toroidal limiter model has been implemented. The model is be discussed and first results are presented. (copyright 2008 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  7. Cardinality enhancement utilizing Sequential Algorithm (SeQ code in OCDMA system

    Directory of Open Access Journals (Sweden)

    Fazlina C. A. S.

    2017-01-01

    Full Text Available Optical Code Division Multiple Access (OCDMA has been important with increasing demand for high capacity and speed for communication in optical networks because of OCDMA technique high efficiency that can be achieved, hence fibre bandwidth is fully used. In this paper we will focus on Sequential Algorithm (SeQ code with AND detection technique using Optisystem design tool. The result revealed SeQ code capable to eliminate Multiple Access Interference (MAI and improve Bit Error Rate (BER, Phase Induced Intensity Noise (PIIN and orthogonally between users in the system. From the results, SeQ shows good performance of BER and capable to accommodate 190 numbers of simultaneous users contrast with existing code. Thus, SeQ code have enhanced the system about 36% and 111% of FCC and DCS code. In addition, SeQ have good BER performance 10-25 at 155 Mbps in comparison with 622 Mbps, 1 Gbps and 2 Gbps bit rate. From the plot graph, 155 Mbps bit rate is suitable enough speed for FTTH and LAN networks. Resolution can be made based on the superior performance of SeQ code. Thus, these codes will give an opportunity in OCDMA system for better quality of service in an optical access network for future generation's usage

  8. Cardinality enhancement utilizing Sequential Algorithm (SeQ) code in OCDMA system

    Science.gov (United States)

    Fazlina, C. A. S.; Rashidi, C. B. M.; Rahman, A. K.; Aljunid, S. A.

    2017-11-01

    Optical Code Division Multiple Access (OCDMA) has been important with increasing demand for high capacity and speed for communication in optical networks because of OCDMA technique high efficiency that can be achieved, hence fibre bandwidth is fully used. In this paper we will focus on Sequential Algorithm (SeQ) code with AND detection technique using Optisystem design tool. The result revealed SeQ code capable to eliminate Multiple Access Interference (MAI) and improve Bit Error Rate (BER), Phase Induced Intensity Noise (PIIN) and orthogonally between users in the system. From the results, SeQ shows good performance of BER and capable to accommodate 190 numbers of simultaneous users contrast with existing code. Thus, SeQ code have enhanced the system about 36% and 111% of FCC and DCS code. In addition, SeQ have good BER performance 10-25 at 155 Mbps in comparison with 622 Mbps, 1 Gbps and 2 Gbps bit rate. From the plot graph, 155 Mbps bit rate is suitable enough speed for FTTH and LAN networks. Resolution can be made based on the superior performance of SeQ code. Thus, these codes will give an opportunity in OCDMA system for better quality of service in an optical access network for future generation's usage

  9. Analysis of the OECD main steam line break benchmark using ANC-K/MIDAC code

    International Nuclear Information System (INIS)

    Aoki, Shigeaki; Tahara, Yoshihisa; Suemura, Takayuki; Ogawa, Junto

    2004-01-01

    A three-dimensional (3D) neutronics and thermal-and-hydraulics (T/H) coupling code ANC-K/MIDAC has been developed. It is the combination of the 3D nodal kinetic code ANC-K and the 3D drift flux thermal hydraulic code MIDAC. In order to verify the adequacy of this code, we have performed several international benchmark problems. In this paper, we show the calculation results of ''OECD Main Steam Line Break Benchmark (MSLB benchmark)'', which gives the typical local power peaking problem. And we calculated the return-to-power scenario of the Phase II problem. The comparison of the results shows the very good agreement of important core parameters between the ANC-K/MIDAC and other participant codes. (author)

  10. COMPARISON OF ELM PULSE PROPAGATION IN THE DIII-D SOL AND DIVERTORS WITH AN ION CONVECTION MODEL

    International Nuclear Information System (INIS)

    FENSTERMACHER, ME; PORTER, GD; LEONARD, AW; BROOKS, NH; BOEDO, JA; COLCHIN, RJ; GRAY, DS; GROEBNER, RJ; GROTH, M; HOGAN, JT; HOLLMANN, EM; LASNIER, CJ; OSBORNE, TH; PETRIE, TW; RUDAKOV, DL; SNYDER, PB; TAKAHASHI, H; WATKINS, JG; ZENG, L; DIII-D TEAM

    2003-01-01

    OAK-B135 Results from dedicated ELM experiments, performed in DIII-D with fast diagnostics to measure the evolution of Type-I ELM effects in the SOL and divertor, are compared with a simple ion convection model and with initial time-dependent UEDGE simulations. Delays between ELM effects observed in the inner versus the outer divertor regions in the experiments scale, as a function of density, with the difference in ion convection time along field lines from the outer midplane to the divertor targets. The ELM perturbation was modeled as an instantaneous radially uniform increase of diffusion coefficients from the top of the pedestal to the outer SOL. The perturbation was confined to a low field side poloidal zone ± 40 o from the outer midplane. The delays in the simulations are similar to those observed in the experiments

  11. An in vitro comparison of the accuracy of implant impressions with coded healing abutments and different implant angulations.

    Science.gov (United States)

    Al-Abdullah, Khaled; Zandparsa, Roya; Finkelman, Matthew; Hirayama, Hiroshi

    2013-08-01

    Fabricating implant definitive casts with CAD/CAM technology (Robocasts) from coded healing abutment impressions represents a simpler and innovative alternative to conventional implant impression techniques. However, information about the accuracy of the impressions and the resultant definitive casts is limited. The purpose of the study was to evaluate the accuracy of the Robocasts and compare them to those definitive casts fabricated with conventional implant impression techniques (open tray with splinted impression copings technique). A reference epoxy resin cast was fabricated and shaped to simulate a dental arch. Two regular platform implant replicas (Biomet 3i Certain, 4.1 mm diameter and 15 mm length) with internal connections were placed 10 mm apart with a 10-degree convergence for one side of the reference resin cast and a 30-degree convergence for the other. Coded healing abutments (Encode) were placed at 3 different heights above the level of the soft tissue replication material (approximately 1, 2, and 4 mm) and served as test groups (E1, E2, and E4), and open trays with splinted impression copings (OTSC) served as a control group. The control group was compared to the impressions of the coded healing abutments by using a standardized measurement protocol. Impressions were made for each group (n=18) and poured with vacuum mixed (100 g powder/20 mL water) Type IV dental stone. The vertical discrepancy (Z axis) between 2 prefabricated passively fitting titanium reference frameworks and the platforms of the implant replicas was measured with an optical comparator applying the 1 screw test. Data were analyzed with Kruskal-Wallis and post-hoc Mann-Whitney U tests, as well as the Wilcoxon signed-rank tests. The Bonferroni correction was used to account for multiple comparisons. The significance level (α) used in a given set of tests was equal to .05 divided by the number of tests performed in that set. The median vertical discrepancy of each coded healing

  12. Cross-code gyrokinetic verification and benchmark on the linear collisionless dynamics of the geodesic acoustic mode

    Science.gov (United States)

    Biancalani, A.; Bottino, A.; Ehrlacher, C.; Grandgirard, V.; Merlo, G.; Novikau, I.; Qiu, Z.; Sonnendrücker, E.; Garbet, X.; Görler, T.; Leerink, S.; Palermo, F.; Zarzoso, D.

    2017-06-01

    The linear properties of the geodesic acoustic modes (GAMs) in tokamaks are investigated by means of the comparison of analytical theory and gyrokinetic numerical simulations. The dependence on the value of the safety factor, finite-orbit-width of the ions in relation to the radial mode width, magnetic-flux-surface shaping, and electron/ion mass ratio are considered. Nonuniformities in the plasma profiles (such as density, temperature, and safety factor), electro-magnetic effects, collisions, and the presence of minority species are neglected. Also, only linear simulations are considered, focusing on the local dynamics. We use three different gyrokinetic codes: the Lagrangian (particle-in-cell) code ORB5, the Eulerian code GENE, and semi-Lagrangian code GYSELA. One of the main aims of this paper is to provide a detailed comparison of the numerical results and analytical theory, in the regimes where this is possible. This helps understanding better the behavior of the linear GAM dynamics in these different regimes, the behavior of the codes, which is crucial in the view of a future work where more physics is present, and the regimes of validity of each specific analytical dispersion relation.

  13. 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.

  14. 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

  15. IM (Integrity Management) software must show flexibility to local codes

    Energy Technology Data Exchange (ETDEWEB)

    Brors, Markus [ROSEN Technology and Research Center GmbH (Germany); Diggory, Ian [Macaw Engineering Ltd., Northumberland (United Kingdom)

    2009-07-01

    There are many internationally recognized codes and standards, such as API 1160 and ASME B31.8S, which help pipeline operators to manage and maintain the integrity of their pipeline networks. However, operators in many countries still use local codes that often reflect the history of pipeline developments in their region and are based on direct experience and research on their pipelines. As pipeline companies come under increasing regulatory and financial pressures to maintain the integrity of their networks, it is important that operators using regional codes are able to benchmark their integrity management schemes against these international standards. Any comprehensive Pipeline Integrity Management System (PIMS) software package should therefore not only incorporate industry standards for pipeline integrity assessment but also be capable of implementing regional codes for comparison purposes. This paper describes the challenges and benefits of incorporating one such set of regional pipeline standards into ROSEN Asset Integrity Management Software (ROAIMS). (author)

  16. A proposed methodology for computational fluid dynamics code verification, calibration, and validation

    Science.gov (United States)

    Aeschliman, D. P.; Oberkampf, W. L.; Blottner, F. G.

    Verification, calibration, and validation (VCV) of Computational Fluid Dynamics (CFD) codes is an essential element of the code development process. The exact manner in which code VCV activities are planned and conducted, however, is critically important. It is suggested that the way in which code validation, in particular, is often conducted--by comparison to published experimental data obtained for other purposes--is in general difficult and unsatisfactory, and that a different approach is required. This paper describes a proposed methodology for CFD code VCV that meets the technical requirements and is philosophically consistent with code development needs. The proposed methodology stresses teamwork and cooperation between code developers and experimentalists throughout the VCV process, and takes advantage of certain synergisms between CFD and experiment. A novel approach to uncertainty analysis is described which can both distinguish between and quantify various types of experimental error, and whose attributes are used to help define an appropriate experimental design for code VCV experiments. The methodology is demonstrated with an example of laminar, hypersonic, near perfect gas, 3-dimensional flow over a sliced sphere/cone of varying geometrical complexity.

  17. Study of reactor analysis codes available at IPEN and their application to problems involving the diffusion theory

    International Nuclear Information System (INIS)

    Mendonca, A.G.

    1980-01-01

    Two computer codes that are available at IPEN for analyses of static neutron diffusion problems are studied and applied. The CITATION code is animed at analyses of criticality, fuel burnup, flux and power distributions etc, in one, two, and three spatial dimensions in multigroup. The EXTERMINATOR code can be used for the same purposes as for CITATION with a limitation to one or two spatial dimensions. Basic theories and numerical techniques used in the codes are studied and summarized. Benchmark problems have been solved using the codes. Comparisons of the results show that both codes can be used with confidence in the analyses of nuclear reactor problems. (author) [pt

  18. ASPECTS CONCERNING THE PRIVATE OWNERSHIP RIGHT WITHIN THE CONTEXT OF THE NEW CIVIL CODE

    Directory of Open Access Journals (Sweden)

    Ana-Maria Lupulescu

    2012-11-01

    Full Text Available The new Civil Code introduces several important changes and clarifications regarding the ownership right in general, and the private ownership right, in particular, so that it becomes necessary, for both the analyst in law and the practitioner, to make a comparison between the old regulation contained in the Civil Code of 1864 and the current regulation provided by the new Civil Code. At least in theory, the new legal framework in this area shows greater consistency and legal precision, although it is not entirely safe from any criticism.

  19. Field-based tests of geochemical modeling codes usign New Zealand hydrothermal systems

    International Nuclear Information System (INIS)

    Bruton, C.J.; Glassley, W.E.; Bourcier, W.L.

    1994-06-01

    Hydrothermal systems in the Taupo Volcanic Zone, North Island, New Zealand are being used as field-based modeling exercises for the EQ3/6 geochemical modeling code package. Comparisons of the observed state and evolution of the hydrothermal systems with predictions of fluid-solid equilibria made using geochemical modeling codes will determine how the codes can be used to predict the chemical and mineralogical response of the environment to nuclear waste emplacement. Field-based exercises allow us to test the models on time scales unattainable in the laboratory. Preliminary predictions of mineral assemblages in equilibrium with fluids sampled from wells in the Wairakei and Kawerau geothermal field suggest that affinity-temperature diagrams must be used in conjunction with EQ6 to minimize the effect of uncertainties in thermodynamic and kinetic data on code predictions

  20. Restructuring of the driver for TASS/SMR, SMART analysis code

    International Nuclear Information System (INIS)

    Lee, G. H.; Kim, S. H.; Whang, Y. D.; Kim, H. C.

    2003-01-01

    TASS/SMR is under development by implementing the newly developed thermal hydraulic models for the integral reactor, SMART. The data structure of TASS utilizes external files, which could result in difficulties for the code maintenance and an easy understanding of the code. The driver for TASS/SMR was rewritten for the use of Fortran 90 advanced features, such as dynamic memory management and user defined derived type. The input structure of TASS changed from the interactive method to card numbered system. Also the output process of TASS reformed by removing the utility program. The validation of the TASS/SMR driver was performed by the comparison of results. The use of the revised TASS/SMR driver is expected to accelerate code development process as well as improve user friendliness

  1. Software requirements specification document for the AREST code development

    International Nuclear Information System (INIS)

    Engel, D.W.; McGrail, B.P.; Whitney, P.D.; Gray, W.J.; Williford, R.E.; White, M.D.; Eslinger, P.W.; Altenhofen, M.K.

    1993-11-01

    The Analysis of the Repository Source Term (AREST) computer code was selected in 1992 by the U.S. Department of Energy. The AREST code will be used to analyze the performance of an underground high level nuclear waste repository. The AREST code is being modified by the Pacific Northwest Laboratory (PNL) in order to evaluate the engineered barrier and waste package designs, model regulatory compliance, analyze sensitivities, and support total systems performance assessment modeling. The current version of the AREST code was developed to be a very useful tool for analyzing model uncertainties and sensitivities to input parameters. The code has also been used successfully in supplying source-terms that were used in a total systems performance assessment. The current version, however, has been found to be inadequate for the comparison and selection of a design for the waste package. This is due to the assumptions and simplifications made in the selection of the process and system models. Thus, the new version of the AREST code will be designed to focus on the details of the individual processes and implementation of more realistic models. This document describes the requirements of the new models that will be implemented. Included in this document is a section describing the near-field environmental conditions for this waste package modeling, description of the new process models that will be implemented, and a description of the computer requirements for the new version of the AREST code

  2. Rate-adaptive BCH codes for distributed source coding

    DEFF Research Database (Denmark)

    Salmistraro, Matteo; Larsen, Knud J.; Forchhammer, Søren

    2013-01-01

    This paper considers Bose-Chaudhuri-Hocquenghem (BCH) codes for distributed source coding. A feedback channel is employed to adapt the rate of the code during the decoding process. The focus is on codes with short block lengths for independently coding a binary source X and decoding it given its...... strategies for improving the reliability of the decoded result are analyzed, and methods for estimating the performance are proposed. In the analysis, noiseless feedback and noiseless communication are assumed. Simulation results show that rate-adaptive BCH codes achieve better performance than low...... correlated side information Y. The proposed codes have been analyzed in a high-correlation scenario, where the marginal probability of each symbol, Xi in X, given Y is highly skewed (unbalanced). Rate-adaptive BCH codes are presented and applied to distributed source coding. Adaptive and fixed checking...

  3. International nuclear model and code comparison on pre-equilibrium effects

    International Nuclear Information System (INIS)

    Gruppelaar, H.; van der Kamp, H.A.J.; Nagel, P.

    1983-01-01

    This paper gives the specification of an intercomparison of statistical nuclear models and codes with emphasis on pre-equilibrium effects. It is partly based upon the conclusions of a meeting of an ad-hoc working group on this subject. The parameters studied are: masses, Q values, level scheme data, optical model parameters, X-ray competition parameters, total level-density specifications, for 86 Rb, 89 Sr, 90 Y, 92 Y, 92 Zr, 93 Zr, 89 Y, 91 Nb, 92 Nb and 93 Nb

  4. A comparison of FEMAXI-III code calculations with irradiation experiments

    International Nuclear Information System (INIS)

    Ito, K.; Sogame, M.; Ichikawa, M.; Nakajima, T.

    1981-01-01

    The FEMAXI-III code calculations were compared with in-pile diameter measurements in the Halden Boiling Water Reactor, in order to check the ability to analyse the pellet-cladding mechanical interaction. The results showed generally good agreement between calculations and measurements. The Studsvik INTER-RAMP Experiments were also analysed to examine the predictability of fuel rod failures. Good agreement was obtained between calculated and measured fission gas x release. The threshold stress to cause failure was estimated by means of FEMAXI-III. (author)

  5. Self-complementary circular codes in coding theory.

    Science.gov (United States)

    Fimmel, Elena; Michel, Christian J; Starman, Martin; Strüngmann, Lutz

    2018-04-01

    Self-complementary circular codes are involved in pairing genetic processes. A maximal [Formula: see text] self-complementary circular code X of trinucleotides was identified in genes of bacteria, archaea, eukaryotes, plasmids and viruses (Michel in Life 7(20):1-16 2017, J Theor Biol 380:156-177, 2015; Arquès and Michel in J Theor Biol 182:45-58 1996). In this paper, self-complementary circular codes are investigated using the graph theory approach recently formulated in Fimmel et al. (Philos Trans R Soc A 374:20150058, 2016). A directed graph [Formula: see text] associated with any code X mirrors the properties of the code. In the present paper, we demonstrate a necessary condition for the self-complementarity of an arbitrary code X in terms of the graph theory. The same condition has been proven to be sufficient for codes which are circular and of large size [Formula: see text] trinucleotides, in particular for maximal circular codes ([Formula: see text] trinucleotides). For codes of small-size [Formula: see text] trinucleotides, some very rare counterexamples have been constructed. Furthermore, the length and the structure of the longest paths in the graphs associated with the self-complementary circular codes are investigated. It has been proven that the longest paths in such graphs determine the reading frame for the self-complementary circular codes. By applying this result, the reading frame in any arbitrary sequence of trinucleotides is retrieved after at most 15 nucleotides, i.e., 5 consecutive trinucleotides, from the circular code X identified in genes. Thus, an X motif of a length of at least 15 nucleotides in an arbitrary sequence of trinucleotides (not necessarily all of them belonging to X) uniquely defines the reading (correct) frame, an important criterion for analyzing the X motifs in genes in the future.

  6. Development of the SPIKE code for analysis of the sodium-water reaction

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Sung Tai; Park, Jin Ho; Choi, Jong Hyeun; Kim, Tae Joon [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-08-01

    In the secondary loop of liquid metal reactors, including SG, water leak into sodium causes the sudden increase of pressure by the H{sub 2} and heat generated from reaction. At few miliseconds after leak, a sharp and short-lived increase of pressure is generated and its propagation depends on the acoustic constraint characteristics of secondary loop. As increasing leak amount of water, another pressure increase is caused by H{sub 2} and its transients depends on the resistance of pressure opening system, such as rupture disc. For prediction of the transients of initial spike pressure, a code of SPIKE was developed. The code was based on the following simplifications and assumptions: combination of total and half release of H{sub 2} rate, spherical shape of H{sub 2} bubble, compressible and Newtonian fluid for sodium. The program was built in FOTRAN language and consisted of 5 modules. Several sample calculations were performed to test the code and to determine the scale down factor of experimental facilities for experimental verification of the code: parameter study of the variables in chemical reaction model, comparison study with results calculated by superposition methods for simple piping structures, comparison study with results calculated by previous researchers, and calculation for KALIMER models of various size. With these calculation results, the generally predicted phenomena of sodium water reaction can be explained and the calculated ones by SPIKE code were well agreed with the previous study. And the scale down factor can be determined. (author). 88 refs., 99 figs., 39 tabs.

  7. Diagonal Eigenvalue Unity (DEU) code for spectral amplitude coding-optical code division multiple access

    Science.gov (United States)

    Ahmed, Hassan Yousif; Nisar, K. S.

    2013-08-01

    Code with ideal in-phase cross correlation (CC) and practical code length to support high number of users are required in spectral amplitude coding-optical code division multiple access (SAC-OCDMA) systems. SAC systems are getting more attractive in the field of OCDMA because of its ability to eliminate the influence of multiple access interference (MAI) and also suppress the effect of phase induced intensity noise (PIIN). In this paper, we have proposed new Diagonal Eigenvalue Unity (DEU) code families with ideal in-phase CC based on Jordan block matrix with simple algebraic ways. Four sets of DEU code families based on the code weight W and number of users N for the combination (even, even), (even, odd), (odd, odd) and (odd, even) are constructed. This combination gives DEU code more flexibility in selection of code weight and number of users. These features made this code a compelling candidate for future optical communication systems. Numerical results show that the proposed DEU system outperforms reported codes. In addition, simulation results taken from a commercial optical systems simulator, Virtual Photonic Instrument (VPI™) shown that, using point to multipoint transmission in passive optical network (PON), DEU has better performance and could support long span with high data rate.

  8. Code Assessment of SPACE 2.19 using LSTF 10% Main Steam-Line-Break Test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minhee; Kim, Seyun [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The Safety and Performance Analysis Code for Nuclear Power Plants (SPACE) has been developed in recent years by the Korea Hydro and Nuclear Power Co. through collaborative works with other Korean nuclear industries and research institutes. As a result of the development, the 2.19 version of the code was released through the successive various verification and validation works. In this study, results produced by the SPACE 2.19 code were compared with the experimental data from JAERI's LSTF Test Run SBSL- 01 for a 10% main steam line break transient in a pressurized water reactor. The LSTF 10% main steam line break test were simulated using the SPACE 2.19 for code V and V work. The overall comparisons between the SPACE 2.19 code prediction and the LSTF Test Run SB-SL-01 experimental data are reasonably satisfactory. The comparisons were conducted in terms of the variations of mass flow rate, void fraction, pressure, collapsed liquid level, temperature, and system flow rate for the transient. In addition, the input model was modified for simulation accuracy of PZR pressure based on the calculated results. The correction of PORV setpoint affects to simulate the PORV open and close phenomena similarly with experiments. From the modification, the computed results show a reasonable agreement with experimental data in overall transient time.

  9. SimProp: a simulation code for ultra high energy cosmic ray propagation

    International Nuclear Information System (INIS)

    Aloisio, R.; Grillo, A.F.; Boncioli, D.; Petrera, S.; Salamida, F.

    2012-01-01

    A new Monte Carlo simulation code for the propagation of Ultra High Energy Cosmic Rays is presented. The results of this simulation scheme are tested by comparison with results of another Monte Carlo computation as well as with the results obtained by directly solving the kinetic equation for the propagation of Ultra High Energy Cosmic Rays. A short comparison with the latest flux published by the Pierre Auger collaboration is also presented

  10. List Decoding of Matrix-Product Codes from nested codes: an application to Quasi-Cyclic codes

    DEFF Research Database (Denmark)

    Hernando, Fernando; Høholdt, Tom; Ruano, Diego

    2012-01-01

    A list decoding algorithm for matrix-product codes is provided when $C_1,..., C_s$ are nested linear codes and $A$ is a non-singular by columns matrix. We estimate the probability of getting more than one codeword as output when the constituent codes are Reed-Solomon codes. We extend this list...... decoding algorithm for matrix-product codes with polynomial units, which are quasi-cyclic codes. Furthermore, it allows us to consider unique decoding for matrix-product codes with polynomial units....

  11. Comparison of the LLL HANDYL76 code with three Stanford Linac experiments

    International Nuclear Information System (INIS)

    Carson, J.K.

    1976-06-01

    HANDYL76, the Lawrence Livermore Laboratory version of SANDYL, the multidimensional Monte Carlo electron/photon computer code, was compared with experimental work performed on the Stanford Mark III Linear Accelerator. The three pieces of experimental work involved bombardment of Al, C, Cu, Sn, and Pb by using 185-MeV electrons, electron-induced cascade showers in Cu, Sn, and Pb at 900 MeV, and cascades in H 2 O and Al at 1 GeV. Agreement between experiment and code calculations were good for the 900-MeV work, but only the aluminum at 185 MeV was within associated uncertainties for the major part of the target. Although good agreement was attained at 1 GeV in aluminum, the H 2 O at 1 GeV produced moderate variations. Possible reasons for the differences obtained are discussed. 89 figures, 16 tables

  12. Performance Comparison of Assorted Color Spaces for Multilevel Block Truncation Coding based Face Recognition

    OpenAIRE

    H.B. Kekre; Sudeep Thepade; Karan Dhamejani; Sanchit Khandelwal; Adnan Azmi

    2012-01-01

    The paper presents a performance analysis of Multilevel Block Truncation Coding based Face Recognition among widely used color spaces. In [1], Multilevel Block Truncation Coding was applied on the RGB color space up to four levels for face recognition. Better results were obtained when the proposed technique was implemented using Kekre’s LUV (K’LUV) color space [25]. This was the motivation to test the proposed technique using assorted color spaces. For experimental analysis, two face databas...

  13. COSA II Further benchmark exercises to compare geomechanical computer codes for salt

    International Nuclear Information System (INIS)

    Lowe, M.J.S.; Knowles, N.C.

    1989-01-01

    Project COSA (COmputer COdes COmparison for SAlt) was a benchmarking exercise involving the numerical modelling of the geomechanical behaviour of heated rock salt. Its main objective was to assess the current European capability to predict the geomechanical behaviour of salt, in the context of the disposal of heat-producing radioactive waste in salt formations. Twelve organisations participated in the exercise in which their solutions to a number of benchmark problems were compared. The project was organised in two distinct phases: The first, from 1984-1986, concentrated on the verification of the computer codes. The second, from 1986-1988 progressed to validation, using three in-situ experiments at the Asse research facility in West Germany as a basis for comparison. This document reports the activities of the second phase of the project and presents the results, assessments and conclusions

  14. The calculation of coolant leak rate through the cracks using RELAP5 code

    International Nuclear Information System (INIS)

    Krungeleviciute, V.; Kaliatka, A.

    2001-01-01

    For reason to choose method of leak detection first of all it is necessary to perform evaluating thermal-hydraulic calculations. These calculations allow to determine flow rate of discharged coolant. For coolant leak rate calculations through possible cracks in Ignalina NPP pipes SQUIRT and RELAP5 thermal-hydraulic codes were used. SQUIRT is well known as computer program that predicts the leakage for cracked pipes in NPP. As this code calculates only water (at subcooled or saturated conditions) leak rate, RELAP5 code model, that calculates water and steam leak rate, was created. For model validation comparison of SQUIRT, RELAP5 and experimental results was performed. Analysis shows RELAP5 code model suitability for calculations of leak through through-wall cracks in pipes. (author)

  15. The management-retrieval code of nuclear level density sub-library (CENPL-NLD)

    International Nuclear Information System (INIS)

    Ge Zhigang; Su Zongdi; Huang Zhongfu; Dong Liaoyuan

    1995-01-01

    The management-retrieval code of the Nuclear Level Density (NLD) is presented. It contains two retrieval ways: single nucleus (SN) and neutron reaction (NR). The latter contains four kinds of retrieval types. This code not only can retrieve level density parameter and the data related to the level density, but also can calculate the relevant data by using different level density parameters and do comparison of the calculated results with related data in order to help user to select level density parameters

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

    Energy Technology Data Exchange (ETDEWEB)

    Sawamura, Sadashi [Hokkaido Univ., Sapporo (Japan)

    2001-03-01

    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 study is to obtain the benchmark data of skyshine and to investigate the calculation code for skyshine. In this fiscal year, the followings are investigated. (1) A {sup 3}He detector and some instruments are added to the former detection system to increase the detection sensitivity in pulsed neutron measurements. Using the new detection system, the skyshine of neutrons from 45 MeV LINAC facility are measured in the distance up to 350 m. (2) To estimate the spectrum of leakage neutron from the facility, {sup 3}He detector with moderators is constructed and the response functions of the detector are calculated using the MCNP simulation code. The leakage spectrum in the facility are measured and unfolded using the SAND-II code. (3) Using the EGS code and/or MCNP code, neutron yields by the photo-nuclear reaction in the lead target are calculated. Then, the neutron fluence at some points including the duct (from which neutrons leaks and is considered to be a skyshine source) is simulated by MCNP MONTE CARLO code. (4) In the distance up to 350 m from the facility, neutron fluence due to the skyshine process are calculated and compared with the experimental results. The comparison gives a fairly good agreement. (author)

  17. Coding Partitions

    Directory of Open Access Journals (Sweden)

    Fabio Burderi

    2007-05-01

    Full Text Available Motivated by the study of decipherability conditions for codes weaker than Unique Decipherability (UD, we introduce the notion of coding partition. Such a notion generalizes that of UD code and, for codes that are not UD, allows to recover the ``unique decipherability" at the level of the classes of the partition. By tacking into account the natural order between the partitions, we define the characteristic partition of a code X as the finest coding partition of X. This leads to introduce the canonical decomposition of a code in at most one unambiguouscomponent and other (if any totally ambiguouscomponents. In the case the code is finite, we give an algorithm for computing its canonical partition. This, in particular, allows to decide whether a given partition of a finite code X is a coding partition. This last problem is then approached in the case the code is a rational set. We prove its decidability under the hypothesis that the partition contains a finite number of classes and each class is a rational set. Moreover we conjecture that the canonical partition satisfies such a hypothesis. Finally we consider also some relationships between coding partitions and varieties of codes.

  18. Verification of the code DYN3D/R with the help of international benchmarks

    International Nuclear Information System (INIS)

    Grundmann, U.; Rohde, U.

    1997-10-01

    Different benchmarks for reactors with quadratic fuel assemblies were calculated with the code DYN3D/R. In this report comparisons with the results of the reference solutions are carried out. The results of DYN3D/R and the reference calculation for the eigenvalue k eff and the power distribution are shown for the steady-state 3-dimensional IAEA-Benchmark. The results of NEACRP-Benchmarks on control rod ejections in a standard PWR were compared with the reference solutions published by the NEA Data Bank. For assessing the accuracy of DYN3D/R results in comparison to other codes the deviations to the reference solutions are considered. Detailed comparisons with the published reference solutions of the NEA-NSC Benchmarks on uncontrolled withdrawal of control rods are made. The influence of the axial nodalization is also investigated. All in all, a good agreement of the DYN3D/R results with the reference solutions can be seen for the considered benchmark problems. (orig.) [de

  19. Consistency and accuracy of diagnostic cancer codes generated by automated registration: comparison with manual registration

    Directory of Open Access Journals (Sweden)

    Codazzi Tiziana

    2006-09-01

    Full Text Available Abstract Background Automated procedures are increasingly used in cancer registration, and it is important that the data produced are systematically checked for consistency and accuracy. We evaluated an automated procedure for cancer registration adopted by the Lombardy Cancer Registry in 1997, comparing automatically-generated diagnostic codes with those produced manually over one year (1997. Methods The automatically generated cancer cases were produced by Open Registry algorithms. For manual registration, trained staff consulted clinical records, pathology reports and death certificates. The social security code, present and checked in both databases in all cases, was used to match the files in the automatic and manual databases. The cancer cases generated by the two methods were compared by manual revision. Results The automated procedure generated 5027 cases: 2959 (59% were accepted automatically and 2068 (41% were flagged for manual checking. Among the cases accepted automatically, discrepancies in data items (surname, first name, sex and date of birth constituted 8.5% of cases, and discrepancies in the first three digits of the ICD-9 code constituted 1.6%. Among flagged cases, cancers of female genital tract, hematopoietic system, metastatic and ill-defined sites, and oropharynx predominated. The usual reasons were use of specific vs. generic codes, presence of multiple primaries, and use of extranodal vs. nodal codes for lymphomas. The percentage of automatically accepted cases ranged from 83% for breast and thyroid cancers to 13% for metastatic and ill-defined cancer sites. Conclusion Since 59% of cases were accepted automatically and contained relatively few, mostly trivial discrepancies, the automatic procedure is efficient for routine case generation effectively cutting the workload required for routine case checking by this amount. Among cases not accepted automatically, discrepancies were mainly due to variations in coding practice.

  20. A PC version of the Monte Carlo criticality code OMEGA

    International Nuclear Information System (INIS)

    Seifert, E.

    1996-05-01

    A description of the PC version of the Monte Carlo criticality code OMEGA is given. The report contains a general description of the code together with a detailed input description. Furthermore, some examples are given illustrating the generation of an input file. The main field of application is the calculation of the criticality of arrangements of fissionable material. Geometrically complicated arrangements that often appear inside and outside a reactor, e.g. in a fuel storage or transport container, can be considered essentially without geometrical approximations. For example, the real geometry of assemblies containing hexagonal or square lattice structures can be described in full detail. Moreover, the code can be used for special investigations in the field of reactor physics and neutron transport. Many years of practical experience and comparison with reference cases have shown that the code together with the built-in data libraries gives reliable results. OMEGA is completely independent on other widely used criticality codes (KENO, MCNP, etc.), concerning programming and the data base. It is a good practice to run difficult criticality safety problems by different independent codes in order to mutually verify the results. In this way, OMEGA can be used as a redundant code within the family of criticality codes. An advantage of OMEGA is the short calculation time: A typical criticality safety application takes only a few minutes on a Pentium PC. Therefore, the influence of parameter variations can simply be investigated by running many variants of a problem. (orig.)

  1. Some benchmark calculations for VVER-1000 assemblies by WIMS-7B code

    International Nuclear Information System (INIS)

    Sultanov, N.V.

    2001-01-01

    Our aim in this report is to compare of calculation results, obtained with the use of different libraries, which are in the variant of the WIMS7B code. We had the three libraries: the 1986 library is based on the UKNDL files, the two 1996 libraries are based on the JEF-2.2 files, the one having the 69 group approximation, the other having the 172 group approximation. We wanted also to have some acquaintance with the new option of WIMS-7B - CACTUS. The variant of WIMS-7B was placed at our disposal by the code authors for a temporal use for 9 months. It was natural to make at comparisons with analogous values of TVS-M, MCU, Apollo-2, Casmo-4, Conkemo, MCNP, HELIOS codes, where the other different libraries were used. In accordance with our aims the calculations of unprofiled and profiled assemblies of the VVER-1000 reactor have been carried out by the option CACTUS. This option provides calculations by the characteristics method. The calculation results have been compared with the K ∞ values obtained by other codes in work. The conclusion from this analysis is such: the methodical parts of errors of these codes have nearly the same values. Spacing for K eff values can be explained of the library microsections differences mainly. Nevertheless, the more detailed analysis of the results obtained is required. In conclusion the calculation of a depletion of VVER-1000 cell has been carried out. The comparison of the dependency of the multiply factor from the depletion obtained by WIMS-7B with different libraries and by the TVS-M, MCU, HELIOS and WIMS-ABBN codes in work has been performed. (orig.)

  2. LDGM Codes for Channel Coding and Joint Source-Channel Coding of Correlated Sources

    Directory of Open Access Journals (Sweden)

    Javier Garcia-Frias

    2005-05-01

    Full Text Available We propose a coding scheme based on the use of systematic linear codes with low-density generator matrix (LDGM codes for channel coding and joint source-channel coding of multiterminal correlated binary sources. In both cases, the structures of the LDGM encoder and decoder are shown, and a concatenated scheme aimed at reducing the error floor is proposed. Several decoding possibilities are investigated, compared, and evaluated. For different types of noisy channels and correlation models, the resulting performance is very close to the theoretical limits.

  3. Development of seismic analysis model for HTGR core on commercial FEM code

    International Nuclear Information System (INIS)

    Tsuji, Nobumasa; Ohashi, Kazutaka

    2015-01-01

    The aftermath of the Great East Japan Earthquake prods to revise the design basis earthquake intensity severely. In aseismic design of block-type HTGR, the securement of structural integrity of core blocks and other structures which are made of graphite become more important. For the aseismic design of block-type HTGR, it is necessary to predict the motion of core blocks which are collided with adjacent blocks. Some seismic analysis codes have been developed in 1970s, but these codes are special purpose-built codes and have poor collaboration with other structural analysis code. We develop the vertical 2 dimensional analytical model on multi-purpose commercial FEM code, which take into account the multiple impacts and friction between block interfaces and rocking motion on contact with dowel pins of the HTGR core by using contact elements. This model is verified by comparison with the experimental results of 12 column vertical slice vibration test. (author)

  4. High-dynamic range compressive spectral imaging by grayscale coded aperture adaptive filtering

    Directory of Open Access Journals (Sweden)

    Nelson Eduardo Diaz

    2015-09-01

    Full Text Available The coded aperture snapshot spectral imaging system (CASSI is an imaging architecture which senses the three dimensional informa-tion of a scene with two dimensional (2D focal plane array (FPA coded projection measurements. A reconstruction algorithm takes advantage of the compressive measurements sparsity to recover the underlying 3D data cube. Traditionally, CASSI uses block-un-block coded apertures (BCA to spatially modulate the light. In CASSI the quality of the reconstructed images depends on the design of these coded apertures and the FPA dynamic range. This work presents a new CASSI architecture based on grayscaled coded apertu-res (GCA which reduce the FPA saturation and increase the dynamic range of the reconstructed images. The set of GCA is calculated in a real-time adaptive manner exploiting the information from the FPA compressive measurements. Extensive simulations show the attained improvement in the quality of the reconstructed images when GCA are employed.  In addition, a comparison between traditional coded apertures and GCA is realized with respect to noise tolerance.

  5. The Initial Atmospheric Transport (IAT) Code: Description and Validation

    Energy Technology Data Exchange (ETDEWEB)

    Morrow, Charles W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bartel, Timothy James [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-10-01

    The Initial Atmospheric Transport (IAT) computer code was developed at Sandia National Laboratories as part of their nuclear launch accident consequences analysis suite of computer codes. The purpose of IAT is to predict the initial puff/plume rise resulting from either a solid rocket propellant or liquid rocket fuel fire. The code generates initial conditions for subsequent atmospheric transport calculations. The Initial Atmospheric Transfer (IAT) code has been compared to two data sets which are appropriate to the design space of space launch accident analyses. The primary model uncertainties are the entrainment coefficients for the extended Taylor model. The Titan 34D accident (1986) was used to calibrate these entrainment settings for a prototypic liquid propellant accident while the recent Johns Hopkins University Applied Physics Laboratory (JHU/APL, or simply APL) large propellant block tests (2012) were used to calibrate the entrainment settings for prototypic solid propellant accidents. North American Meteorology (NAM )formatted weather data profiles are used by IAT to determine the local buoyancy force balance. The IAT comparisons for the APL solid propellant tests illustrate the sensitivity of the plume elevation to the weather profiles; that is, the weather profile is a dominant factor in determining the plume elevation. The IAT code performed remarkably well and is considered validated for neutral weather conditions.

  6. Verification test calculations for the Source Term Code Package

    International Nuclear Information System (INIS)

    Denning, R.S.; Wooton, R.O.; Alexander, C.A.; Curtis, L.A.; Cybulskis, P.; Gieseke, J.A.; Jordan, H.; Lee, K.W.; Nicolosi, S.L.

    1986-07-01

    The purpose of this report is to demonstrate the reasonableness of the Source Term Code Package (STCP) results. Hand calculations have been performed spanning a wide variety of phenomena within the context of a single accident sequence, a loss of all ac power with late containment failure, in the Peach Bottom (BWR) plant, and compared with STCP results. The report identifies some of the limitations of the hand calculation effort. The processes involved in a core meltdown accident are complex and coupled. Hand calculations by their nature must deal with gross simplifications of these processes. Their greatest strength is as an indicator that a computer code contains an error, for example that it doesn't satisfy basic conservation laws, rather than in showing the analysis accurately represents reality. Hand calculations are an important element of verification but they do not satisfy the need for code validation. The code validation program for the STCP is a separate effort. In general the hand calculation results show that models used in the STCP codes (e.g., MARCH, TRAP-MELT, VANESA) obey basic conservation laws and produce reasonable results. The degree of agreement and significance of the comparisons differ among the models evaluated. 20 figs., 26 tabs

  7. Studies of fast reactor disassembly using a Bethe-Tait computer code

    International Nuclear Information System (INIS)

    Ludwig, J.C.

    1978-10-01

    The advantages of the fast reactor are given and the general design outlined. Loss of Flow and Transient Overpower faults are possible; the potential consequences of such incidents are analysed using a deterministic approach. The course of an incident is split into several stages; of these only predisassembly and disassembly are considered. Predisassembly computer codes are described in general, and several particular codes are examined in more detail, based on a literature survey. The results and implications of disassembly calculations using the code EXTRA are presented. Here, the effects of several factors, such as the presence of retained fission gases and possible restraints on fuel motion, are investigated. Some comparisons are made with published results from the VENUS-II disassembly code. A general conclusion is that under some circumstances, the yield predicted during disassembly is relatively insensitive to modelling assumptions, and a simple code such as EXTRA may prove adequate if explicit core displacements are not required. A major factor in determining the yield of the disassembly phase is confirmed as being the rate of reactivity insertion during disassembly, as predicted by a predisassembly code. (U.K.)

  8. Hardware-efficient bosonic quantum error-correcting codes based on symmetry operators

    Science.gov (United States)

    Niu, Murphy Yuezhen; Chuang, Isaac L.; Shapiro, Jeffrey H.

    2018-03-01

    We establish a symmetry-operator framework for designing quantum error-correcting (QEC) codes based on fundamental properties of the underlying system dynamics. Based on this framework, we propose three hardware-efficient bosonic QEC codes that are suitable for χ(2 )-interaction based quantum computation in multimode Fock bases: the χ(2 ) parity-check code, the χ(2 ) embedded error-correcting code, and the χ(2 ) binomial code. All of these QEC codes detect photon-loss or photon-gain errors by means of photon-number parity measurements, and then correct them via χ(2 ) Hamiltonian evolutions and linear-optics transformations. Our symmetry-operator framework provides a systematic procedure for finding QEC codes that are not stabilizer codes, and it enables convenient extension of a given encoding to higher-dimensional qudit bases. The χ(2 ) binomial code is of special interest because, with m ≤N identified from channel monitoring, it can correct m -photon-loss errors, or m -photon-gain errors, or (m -1 )th -order dephasing errors using logical qudits that are encoded in O (N ) photons. In comparison, other bosonic QEC codes require O (N2) photons to correct the same degree of bosonic errors. Such improved photon efficiency underscores the additional error-correction power that can be provided by channel monitoring. We develop quantum Hamming bounds for photon-loss errors in the code subspaces associated with the χ(2 ) parity-check code and the χ(2 ) embedded error-correcting code, and we prove that these codes saturate their respective bounds. Our χ(2 ) QEC codes exhibit hardware efficiency in that they address the principal error mechanisms and exploit the available physical interactions of the underlying hardware, thus reducing the physical resources required for implementing their encoding, decoding, and error-correction operations, and their universal encoded-basis gate sets.

  9. Software Certification - Coding, Code, and Coders

    Science.gov (United States)

    Havelund, Klaus; Holzmann, Gerard J.

    2011-01-01

    We describe a certification approach for software development that has been adopted at our organization. JPL develops robotic spacecraft for the exploration of the solar system. The flight software that controls these spacecraft is considered to be mission critical. We argue that the goal of a software certification process cannot be the development of "perfect" software, i.e., software that can be formally proven to be correct under all imaginable and unimaginable circumstances. More realistically, the goal is to guarantee a software development process that is conducted by knowledgeable engineers, who follow generally accepted procedures to control known risks, while meeting agreed upon standards of workmanship. We target three specific issues that must be addressed in such a certification procedure: the coding process, the code that is developed, and the skills of the coders. The coding process is driven by standards (e.g., a coding standard) and tools. The code is mechanically checked against the standard with the help of state-of-the-art static source code analyzers. The coders, finally, are certified in on-site training courses that include formal exams.

  10. Validation and comparison of dispersion models of RTARC DSS

    International Nuclear Information System (INIS)

    Duran, J.; Pospisil, M.

    2004-01-01

    RTARC DSS (Real Time Accident Release Consequences - Decision Support System) is a computer code developed at the VUJE Trnava, Inc. (Stubna, M. et al, 1993). The code calculations include atmospheric transport and diffusion, dose assessment, evaluation and displaying of the affected zones, evaluation of the early health effects, concentration and dose rate time dependence in the selected sites etc. The simulation of the protective measures (sheltering, iodine administration) is involved. The aim of this paper is to present the process of validation of the RTARC dispersion models. RTARC includes models for calculations of release for very short (Method Monte Carlo - MEMOC), short (Gaussian Straight-Line Model) and long distances (Puff Trajectory Model - PTM). Validation of the code RTARC was performed using the results of comparisons and experiments summarized in the Table 1.: 1. Experiments and comparisons in the process of validation of the system RTARC - experiments or comparison - distance - model. Wind tunnel experiments (Universitaet der Bundeswehr, Muenchen) - Area of NPP - Method Monte Carlo. INEL (Idaho National Engineering Laboratory) - short/medium - Gaussian model and multi tracer atmospheric experiment - distances - PTM. Model Validation Kit - short distances - Gaussian model. STEP II.b 'Realistic Case Studies' - long distances - PTM. ENSEMBLE comparison - long distances - PTM (orig.)

  11. Comparison of EGS4 and MCNP Monte Carlo codes when calculating radiotherapy depth doses.

    Science.gov (United States)

    Love, P A; Lewis, D G; Al-Affan, I A; Smith, C W

    1998-05-01

    The Monte Carlo codes EGS4 and MCNP have been compared when calculating radiotherapy depth doses in water. The aims of the work were to study (i) the differences between calculated depth doses in water for a range of monoenergetic photon energies and (ii) the relative efficiency of the two codes for different electron transport energy cut-offs. The depth doses from the two codes agree with each other within the statistical uncertainties of the calculations (1-2%). The relative depth doses also agree with data tabulated in the British Journal of Radiology Supplement 25. A discrepancy in the dose build-up region may by attributed to the different electron transport algorithims used by EGS4 and MCNP. This discrepancy is considerably reduced when the improved electron transport routines are used in the latest (4B) version of MCNP. Timing calculations show that EGS4 is at least 50% faster than MCNP for the geometries used in the simulations.

  12. The adjoint sensitivity method, a contribution to the code uncertainty evaluation

    International Nuclear Information System (INIS)

    Ounsy, A.; Crecy, F. de; Brun, B.

    1993-01-01

    The application of the ASM (Adjoint Sensitivity Method) to thermohydraulic codes, is examined. The advantage of the method is to be very few CPU time consuming in comparison with usual approach requiring one complete code run per sensitivity determination. The mathematical aspects of the problem are first described, and the applicability of the method of the functional-type response of a thermalhydraulic model is demonstrated. On a simple example of non linear hyperbolic equation (Burgers equation) the problem has been analyzed. It is shown that the formalism used in the literature treating this subject is not appropriate. A new mathematical formalism circumventing the problem is proposed. For the discretized form of the problem, two methods are possible: the Continuous ASM and the Discrete ASM. The equivalence of both methods is demonstrated; nevertheless only the DASM constitutes a practical solution for thermalhydraulic codes. The application of the DASM to the thermalhydraulic safety code CATHARE is then presented for two examples. They demonstrate that ASM constitutes an efficient tool for the analysis of code sensitivity. (authors) 7 figs., 5 tabs., 8 refs

  13. The adjoint sensitivity method. A contribution to the code uncertainty evaluation

    International Nuclear Information System (INIS)

    Ounsy, A.; Brun, B.

    1993-01-01

    The application of the ASM (Adjoint Sensitivity Method) to thermohydraulic codes, is examined. The advantage of the method is to be very few CPU time consuming in comparison with usual approach requiring one complete code run per sensitivity determination. The mathematical aspects of the problem are first described, and the applicability of the method of the functional-type response of a thermalhydraulic model is demonstrated. On a simple example of non linear hyperbolic equation (Burgers equation) the problem has been analyzed. It is shown that the formalism used in the literature treating this subject is not appropriate. A new mathematical formalism circumventing the problem is proposed. For the discretized form of the problem, two methods are possible: the Continuous ASM and the Discrete ASM. The equivalence of both methods is demonstrated; nevertheless only the DASM constitutes a practical solution for thermalhydraulic codes. The application of the DASM to the thermalhydraulic safety code CATHARE is then presented for two examples. They demonstrate that ASM constitutes an efficient tool for the analysis of code sensitivity. (authors) 7 figs., 5 tabs., 8 refs

  14. Instantly Decodable Network Coding: From Centralized to Device-to-Device Communications

    KAUST Repository

    Douik, Ahmed S.

    2015-05-01

    From its introduction to its quindecennial, network coding have built a strong reputation in enhancing packet recovery process and achieving maximum information flow in both wires and wireless networks. Traditional studies focused on optimizing the throughput of the network by proposing complex schemes that achieve optimal delay. With the shift toward distributed computing at mobile devices, throughput and complexity become both critical factors that affect the efficiency of a coding scheme. Instantly decodable network coding imposed itself as a new paradigm in network coding that trades off this two aspects. This paper presents a survey of instantly decodable network coding schemes that are proposed in the literature. The various schemes are identified, categorized and evaluated. Two categories can be distinguished namely the conventional centralized schemes and the distributed or cooperative schemes. For each scheme, the comparison is carried out in terms of reliability, performance, complexity and packet selection methodology. Although the performance is generally inversely proportional to the computation complexity, numerous successful schemes from both the performance and complexity viewpoint are identified.

  15. 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)

  16. The adjoint sensitivity method. A contribution to the code uncertainty evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Ounsy, A; Brun, B

    1994-12-31

    The application of the ASM (Adjoint Sensitivity Method) to thermohydraulic codes, is examined. The advantage of the method is to be very few CPU time consuming in comparison with usual approach requiring one complete code run per sensitivity determination. The mathematical aspects of the problem are first described, and the applicability of the method of the functional-type response of a thermalhydraulic model is demonstrated. On a simple example of non linear hyperbolic equation (Burgers equation) the problem has been analyzed. It is shown that the formalism used in the literature treating this subject is not appropriate. A new mathematical formalism circumventing the problem is proposed. For the discretized form of the problem, two methods are possible: the Continuous ASM and the Discrete ASM. The equivalence of both methods is demonstrated; nevertheless only the DASM constitutes a practical solution for thermalhydraulic codes. The application of the DASM to the thermalhydraulic safety code CATHARE is then presented for two examples. They demonstrate that ASM constitutes an efficient tool for the analysis of code sensitivity. (authors) 7 figs., 5 tabs., 8 refs.

  17. The adjoint sensitivity method, a contribution to the code uncertainty evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Ounsy, A; Crecy, F de; Brun, B

    1994-12-31

    The application of the ASM (Adjoint Sensitivity Method) to thermohydraulic codes, is examined. The advantage of the method is to be very few CPU time consuming in comparison with usual approach requiring one complete code run per sensitivity determination. The mathematical aspects of the problem are first described, and the applicability of the method of the functional-type response of a thermalhydraulic model is demonstrated. On a simple example of non linear hyperbolic equation (Burgers equation) the problem has been analyzed. It is shown that the formalism used in the literature treating this subject is not appropriate. A new mathematical formalism circumventing the problem is proposed. For the discretized form of the problem, two methods are possible: the Continuous ASM and the Discrete ASM. The equivalence of both methods is demonstrated; nevertheless only the DASM constitutes a practical solution for thermalhydraulic codes. The application of the DASM to the thermalhydraulic safety code CATHARE is then presented for two examples. They demonstrate that ASM constitutes an efficient tool for the analysis of code sensitivity. (authors) 7 figs., 5 tabs., 8 refs.

  18. Instantly Decodable Network Coding: From Centralized to Device-to-Device Communications

    KAUST Repository

    Douik, Ahmed S.

    2015-01-01

    From its introduction to its quindecennial, network coding have built a strong reputation in enhancing packet recovery process and achieving maximum information flow in both wires and wireless networks. Traditional studies focused on optimizing the throughput of the network by proposing complex schemes that achieve optimal delay. With the shift toward distributed computing at mobile devices, throughput and complexity become both critical factors that affect the efficiency of a coding scheme. Instantly decodable network coding imposed itself as a new paradigm in network coding that trades off this two aspects. This paper presents a survey of instantly decodable network coding schemes that are proposed in the literature. The various schemes are identified, categorized and evaluated. Two categories can be distinguished namely the conventional centralized schemes and the distributed or cooperative schemes. For each scheme, the comparison is carried out in terms of reliability, performance, complexity and packet selection methodology. Although the performance is generally inversely proportional to the computation complexity, numerous successful schemes from both the performance and complexity viewpoint are identified.

  19. Coding efficiency of AVS 2.0 for CBAC and CABAC engines

    Science.gov (United States)

    Cui, Jing; Choi, Youngkyu; Chae, Soo-Ik

    2015-12-01

    In this paper we compare the coding efficiency of AVS 2.0[1] for engines of the Context-based Binary Arithmetic Coding (CBAC)[2] in the AVS 2.0 and the Context-Adaptive Binary Arithmetic Coder (CABAC)[3] in the HEVC[4]. For fair comparison, the CABAC is embedded in the reference code RD10.1 because the CBAC is in the HEVC in our previous work[5]. The rate estimation table is employed only for RDOQ in the RD code. To reduce the computation complexity of the video encoder, therefore we modified the RD code so that the rate estimation table is employed for all RDO decision. Furthermore, we also simplify the complexity of rate estimation table by reducing the bit depth of its fractional part to 2 from 8. The simulation result shows that the CABAC has the BD-rate loss of about 0.7% compared to the CBAC. It seems that the CBAC is a little more efficient than that the CABAC in the AVS 2.0.

  20. Validation of DRAGON code in connection with WIMS-AECL/RFSP code system based on ENDF/B-VI library and two group model

    International Nuclear Information System (INIS)

    Hong, In Seob; Suk, Ho Chun; Kim, Soon Young; Jo, Chang Keun

    2002-06-01

    The major objective of this research is to validate the incremental cross section property of DRAGON code in connection with WIMS-AECL/DRAGON/RFSP code system with ENDF/B-VI library and full 2G calculation model. The direct comparison between the incremental cross section results calculated by DRAGON with ENDF/B-VI and ENDF/B-V and MULTICELL with ENDF/B-V indicate that there are not much differences between the incremental cross sections of DRAGON with ENDF/B-V and ENDF/B-VI, but there exists large discrepancies between the results of DRAGON and those of MULTICELL. In the analysis of the difference between calculated and measured reactivity worths of various types of control devices during Phase-B Post-Simulation of Wolsong Units 2, 3 and 4, WIMS-AECL/DRAGON/RFSP analysis well agrees with those of previous WIMS-AECL /MULTICELL/RFSP analysis within very small differences. From those results, we can conclude that DRAGON code can be used as a general purpose incremental cross section generation tool for not only the natural uranium fuel but also slightly enriched fuel such as RU or SEU, to cover the shortcomings of natural uranium based MULTICELL code

Some links on this page may take you to non-federal websites. Their policies may differ from this site.