WorldWideScience

Sample records for hybrid codes developed

  1. Development and Benchmarking of a Hybrid PIC Code For Dense Plasmas and Fast Ignition

    Energy Technology Data Exchange (ETDEWEB)

    Witherspoon, F. Douglas [HyperV Technologies Corp.; Welch, Dale R. [Voss Scientific, LLC; Thompson, John R. [FAR-TECH, Inc.; MacFarlane, Joeseph J. [Prism Computational Sciences Inc.; Phillips, Michael W. [Advanced Energy Systems, Inc.; Bruner, Nicki [Voss Scientific, LLC; Mostrom, Chris [Voss Scientific, LLC; Thoma, Carsten [Voss Scientific, LLC; Clark, R. E. [Voss Scientific, LLC; Bogatu, Nick [FAR-TECH, Inc.; Kim, Jin-Soo [FAR-TECH, Inc.; Galkin, Sergei [FAR-TECH, Inc.; Golovkin, Igor E. [Prism Computational Sciences, Inc.; Woodruff, P. R. [Prism Computational Sciences, Inc.; Wu, Linchun [HyperV Technologies Corp.; Messer, Sarah J. [HyperV Technologies Corp.

    2014-05-20

    Radiation processes play an important role in the study of both fast ignition and other inertial confinement schemes, such as plasma jet driven magneto-inertial fusion, both in their effect on energy balance, and in generating diagnostic signals. In the latter case, warm and hot dense matter may be produced by the convergence of a plasma shell formed by the merging of an assembly of high Mach number plasma jets. This innovative approach has the potential advantage of creating matter of high energy densities in voluminous amount compared with high power lasers or particle beams. An important application of this technology is as a plasma liner for the flux compression of magnetized plasma to create ultra-high magnetic fields and burning plasmas. HyperV Technologies Corp. has been developing plasma jet accelerator technology in both coaxial and linear railgun geometries to produce plasma jets of sufficient mass, density, and velocity to create such imploding plasma liners. An enabling tool for the development of this technology is the ability to model the plasma dynamics, not only in the accelerators themselves, but also in the resulting magnetized target plasma and within the merging/interacting plasma jets during transport to the target. Welch pioneered numerical modeling of such plasmas (including for fast ignition) using the LSP simulation code. Lsp is an electromagnetic, parallelized, plasma simulation code under development since 1995. It has a number of innovative features making it uniquely suitable for modeling high energy density plasmas including a hybrid fluid model for electrons that allows electrons in dense plasmas to be modeled with a kinetic or fluid treatment as appropriate. In addition to in-house use at Voss Scientific, several groups carrying out research in Fast Ignition (LLNL, SNL, UCSD, AWE (UK), and Imperial College (UK)) also use LSP. A collaborative team consisting of HyperV Technologies Corp., Voss Scientific LLC, FAR-TECH, Inc., Prism

  2. Analysis of Non-binary Hybrid LDPC Codes

    OpenAIRE

    Sassatelli, Lucile; Declercq, David

    2008-01-01

    In this paper, we analyse asymptotically a new class of LDPC codes called Non-binary Hybrid LDPC codes, which has been recently introduced. We use density evolution techniques to derive a stability condition for hybrid LDPC codes, and prove their threshold behavior. We study this stability condition to conclude on asymptotic advantages of hybrid LDPC codes compared to their non-hybrid counterparts.

  3. 1-D hybrid code for FRM start-up

    International Nuclear Information System (INIS)

    Stark, R.A.; Miley, G.H.

    1982-01-01

    A one-D hybrid has been developed to study the start-up of the FRM via neutral-beam injection. The code uses a multi-group numerical model originally developed by J. Willenberg to describe fusion product dynamics in a solenoidal plasma. Earlier we described such a model for use in determining self-consistent ion currents and magnetic fields in FRM start-up. However, consideration of electron dynamics during start-up indicate that the electron current will oppose the injected ion current and may even foil the attempt to achieve reversal. For this reason, we have combined the multi-group ion (model) with a fluid treatment for electron dynamics to form the hybrid code FROST (Field Reversed One-dimensional STart-up). The details of this merger, along with sample results of operation of FROST, are given

  4. Hybrid coded aperture and Compton imaging using an active mask

    International Nuclear Information System (INIS)

    Schultz, L.J.; Wallace, M.S.; Galassi, M.C.; Hoover, A.S.; Mocko, M.; Palmer, D.M.; Tornga, S.R.; Kippen, R.M.; Hynes, M.V.; Toolin, M.J.; Harris, B.; McElroy, J.E.; Wakeford, D.; Lanza, R.C.; Horn, B.K.P.; Wehe, D.K.

    2009-01-01

    The trimodal imager (TMI) images gamma-ray sources from a mobile platform using both coded aperture (CA) and Compton imaging (CI) modalities. In this paper we will discuss development and performance of image reconstruction algorithms for the TMI. In order to develop algorithms in parallel with detector hardware we are using a GEANT4 [J. Allison, K. Amako, J. Apostolakis, H. Araujo, P.A. Dubois, M. Asai, G. Barrand, R. Capra, S. Chauvie, R. Chytracek, G. Cirrone, G. Cooperman, G. Cosmo, G. Cuttone, G. Daquino, et al., IEEE Trans. Nucl. Sci. NS-53 (1) (2006) 270] based simulation package to produce realistic data sets for code development. The simulation code incorporates detailed detector modeling, contributions from natural background radiation, and validation of simulation results against measured data. Maximum likelihood algorithms for both imaging methods are discussed, as well as a hybrid imaging algorithm wherein CA and CI information is fused to generate a higher fidelity reconstruction.

  5. 1-D hybrid code for FRM dynamics

    International Nuclear Information System (INIS)

    Stark, R.A.; Miley, G.H.

    1985-01-01

    A 1-D radial hybrid code has been written to study the start-up of the FRM via neutral-beam injection. This code, named FROST (Field Reversed One-dimensional STart-up), models the plasma as azimuthal symmetric with no axial dependence. A multi-group method in energy and canonical angular momentum describes the large-orbit ions from the beam. This method is designed to be more efficient than those employing particle tracking, since the characteristic timescale of the simulation is the ion slowing down time, rather than the much shorter cyclotron period. A time-differentiated Grad-Shafranov equation couples the ion current to massless fluid equations describing electrons and low energy ions. Flux coordinates are used in this fluid model, in preference to an Eulerian framework, so that coupling of plasma at the two different radii of a closed flux surface may be treated with ease. Since a fluid treatment for electrons is invalid near a field null, a separate model for the electron current has been included for this region, a unique feature. Results of simulation of injection into a 2XIIB-like plasma are discussed. Electron currents are found to retard, but not prevent reversal of the magnetic field at the plasma center

  6. Hybrid 3D Fractal Coding with Neighbourhood Vector Quantisation

    Directory of Open Access Journals (Sweden)

    Zhen Yao

    2004-12-01

    Full Text Available A hybrid 3D compression scheme which combines fractal coding with neighbourhood vector quantisation for video and volume data is reported. While fractal coding exploits the redundancy present in different scales, neighbourhood vector quantisation, as a generalisation of translational motion compensation, is a useful method for removing both intra- and inter-frame coherences. The hybrid coder outperforms most of the fractal coders published to date while the algorithm complexity is kept relatively low.

  7. Non-binary Hybrid LDPC Codes: Structure, Decoding and Optimization

    OpenAIRE

    Sassatelli, Lucile; Declercq, David

    2007-01-01

    In this paper, we propose to study and optimize a very general class of LDPC codes whose variable nodes belong to finite sets with different orders. We named this class of codes Hybrid LDPC codes. Although efficient optimization techniques exist for binary LDPC codes and more recently for non-binary LDPC codes, they both exhibit drawbacks due to different reasons. Our goal is to capitalize on the advantages of both families by building codes with binary (or small finite set order) and non-bin...

  8. Development status of TUF code

    International Nuclear Information System (INIS)

    Liu, W.S.; Tahir, A.; Zaltsgendler

    1996-01-01

    An overview of the important development of the TUF code in 1995 is presented. The development in the following areas is presented: control of round-off error propagation, gas resolution and release models, and condensation induced water hammer. This development is mainly generated from station requests for operational support and code improvement. (author)

  9. Computer code MLCOSP for multiple-correlation and spectrum analysis with a hybrid computer

    International Nuclear Information System (INIS)

    Oguma, Ritsuo; Fujii, Yoshio; Usui, Hozumi; Watanabe, Koichi

    1975-10-01

    Usage of the computer code MLCOSP(Multiple Correlation and Spectrum) developed is described for a hybrid computer installed in JAERI Functions of the hybrid computer and its terminal devices are utilized ingeniously in the code to reduce complexity of the data handling which occurrs in analysis of the multivariable experimental data and to perform the analysis in perspective. Features of the code are as follows; Experimental data can be fed to the digital computer through the analog part of the hybrid computer by connecting with a data recorder. The computed results are displayed in figures, and hardcopies are taken when necessary. Series-messages to the code are shown on the terminal, so man-machine communication is possible. And further the data can be put in through a keyboard, so case study according to the results of analysis is possible. (auth.)

  10. Accident consequence assessment code development

    International Nuclear Information System (INIS)

    Homma, T.; Togawa, O.

    1991-01-01

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

  11. Scalable Simulation of Electromagnetic Hybrid Codes

    International Nuclear Information System (INIS)

    Perumalla, Kalyan S.; Fujimoto, Richard; Karimabadi, Dr. Homa

    2006-01-01

    New discrete-event formulations of physics simulation models are emerging that can outperform models based on traditional time-stepped techniques. Detailed simulation of the Earth's magnetosphere, for example, requires execution of sub-models that are at widely differing timescales. In contrast to time-stepped simulation which requires tightly coupled updates to entire system state at regular time intervals, the new discrete event simulation (DES) approaches help evolve the states of sub-models on relatively independent timescales. However, parallel execution of DES-based models raises challenges with respect to their scalability and performance. One of the key challenges is to improve the computation granularity to offset synchronization and communication overheads within and across processors. Our previous work was limited in scalability and runtime performance due to the parallelization challenges. Here we report on optimizations we performed on DES-based plasma simulation models to improve parallel performance. The net result is the capability to simulate hybrid particle-in-cell (PIC) models with over 2 billion ion particles using 512 processors on supercomputing platforms

  12. LiveCode mobile development

    CERN Document Server

    Lavieri, Edward D

    2013-01-01

    A practical guide written in a tutorial-style, ""LiveCode Mobile Development Hotshot"" walks you step-by-step through 10 individual projects. Every project is divided into sub tasks to make learning more organized and easy to follow along with explanations, diagrams, screenshots, and downloadable material.This book is great for anyone who wants to develop mobile applications using LiveCode. You should be familiar with LiveCode and have access to a smartphone. You are not expected to know how to create graphics or audio clips.

  13. Adaptive RD Optimized Hybrid Sound Coding

    NARCIS (Netherlands)

    Schijndel, N.H. van; Bensa, J.; Christensen, M.G.; Colomes, C.; Edler, B.; Heusdens, R.; Jensen, J.; Jensen, S.H.; Kleijn, W.B.; Kot, V.; Kövesi, B.; Lindblom, J.; Massaloux, D.; Niamut, O.A.; Nordén, F.; Plasberg, J.H.; Vafin, R.; Virette, D.; Wübbolt, O.

    2008-01-01

    Traditionally, sound codecs have been developed with a particular application in mind, their performance being optimized for specific types of input signals, such as speech or audio (music), and application constraints, such as low bit rate, high quality, or low delay. There is, however, an

  14. Spatial resolution enhancement residual coding using hybrid ...

    Indian Academy of Sciences (India)

    ble value of the parameters (ρ1,ρ2,θ) in their respective ranges. For each block ... Dirac was developed by BBC and aimed at a royalty-free, open technology (Dirac video codec [online]. ..... In: Acoustics, Speech, and Signal Processing, 1995.

  15. Coarse mesh code development

    Energy Technology Data Exchange (ETDEWEB)

    Lieberoth, J.

    1975-06-15

    The numerical solution of the neutron diffusion equation plays a very important role in the analysis of nuclear reactors. A wide variety of numerical procedures has been proposed, at which most of the frequently used numerical methods are fundamentally based on the finite- difference approximation where the partial derivatives are approximated by the finite difference. For complex geometries, typical of the practical reactor problems, the computational accuracy of the finite-difference method is seriously affected by the size of the mesh width relative to the neutron diffusion length and by the heterogeneity of the medium. Thus, a very large number of mesh points are generally required to obtain a reasonably accurate approximate solution of the multi-dimensional diffusion equation. Since the computation time is approximately proportional to the number of mesh points, a detailed multidimensional analysis, based on the conventional finite-difference method, is still expensive even with modern large-scale computers. Accordingly, there is a strong incentive to develop alternatives that can reduce the number of mesh-points and still retain accuracy. One of the promising alternatives is the finite element method, which consists of the expansion of the neutron flux by piecewise polynomials. One of the advantages of this procedure is its flexibility in selecting the locations of the mesh points and the degree of the expansion polynomial. The small number of mesh points of the coarse grid enables to store the results of several of the least outer iterations and to calculate well extrapolated values of them by comfortable formalisms. This holds especially if only one energy distribution of fission neutrons is assumed for all fission processes in the reactor, because the whole information of an outer iteration is contained in a field of fission rates which has the size of all mesh points of the coarse grid.

  16. Hybrid microscopic depletion model in nodal code DYN3D

    International Nuclear Information System (INIS)

    Bilodid, Y.; Kotlyar, D.; Shwageraus, E.; Fridman, E.; Kliem, S.

    2016-01-01

    Highlights: • A new hybrid method of accounting for spectral history effects is proposed. • Local concentrations of over 1000 nuclides are calculated using micro depletion. • The new method is implemented in nodal code DYN3D and verified. - Abstract: The paper presents a general hybrid method that combines the micro-depletion technique with correction of micro- and macro-diffusion parameters to account for the spectral history effects. The fuel in a core is subjected to time- and space-dependent operational conditions (e.g. coolant density), which cannot be predicted in advance. However, lattice codes assume some average conditions to generate cross sections (XS) for nodal diffusion codes such as DYN3D. Deviation of local operational history from average conditions leads to accumulation of errors in XS, which is referred as spectral history effects. Various methods to account for the spectral history effects, such as spectral index, burnup-averaged operational parameters and micro-depletion, were implemented in some nodal codes. Recently, an alternative method, which characterizes fuel depletion state by burnup and 239 Pu concentration (denoted as Pu-correction) was proposed, implemented in nodal code DYN3D and verified for a wide range of history effects. The method is computationally efficient, however, it has applicability limitations. The current study seeks to improve the accuracy and applicability range of Pu-correction method. The proposed hybrid method combines the micro-depletion method with a XS characterization technique similar to the Pu-correction method. The method was implemented in DYN3D and verified on multiple test cases. The results obtained with DYN3D were compared to those obtained with Monte Carlo code Serpent, which was also used to generate the XS. The observed differences are within the statistical uncertainties.

  17. Development of the DTNTES code

    International Nuclear Information System (INIS)

    Ortega Prieto, P.; Morales Dorado, M.D.; Alonso Santos, A.

    1987-01-01

    The DTNTES code has been developed in the Department of Nuclear Technology of the Polytechnical University in Madrid as a part of the Research Program on Quantitative Risk Analysis. DTNTES code calculates several time-dependent probabilistic characteristics of basic events, minimal cut sets and the top event of a fault tree. The code assumes that basic events are statistically independent, and they have failure and repair distributions. It computes the minimal cut upper bound approach for the top event unavailability, and the time-dependent unreliability of the top event by means of different methods, selected by the user. These methods are: expected number of system failures, failure rate, Barlow-Proschan bound, steady-state upper bound, and T* method. (author)

  18. Development of TIME2 code

    International Nuclear Information System (INIS)

    1986-02-01

    The paper reviews the progress on the development of a computer model TIME2, for modelling the long term evolution of shallow burial site environments for low- and intermediate-level radioactive waste disposal. The subject is discussed under the five topic headings: 1) background studies, including geomorphology, climate, human-induced effects, and seismicity, 2) development of the TIME2 code, 3) verification and testing, 4) documentation, and, 5) role of TIME2 in radiological risk assessment. (U.K.)

  19. 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.

  20. A hybrid gyrokinetic ion and isothermal electron fluid code for astrophysical plasma

    Science.gov (United States)

    Kawazura, Y.; Barnes, M.

    2018-05-01

    This paper describes a new code for simulating astrophysical plasmas that solves a hybrid model composed of gyrokinetic ions (GKI) and an isothermal electron fluid (ITEF) Schekochihin et al. (2009) [9]. This model captures ion kinetic effects that are important near the ion gyro-radius scale while electron kinetic effects are ordered out by an electron-ion mass ratio expansion. The code is developed by incorporating the ITEF approximation into AstroGK, an Eulerian δf gyrokinetics code specialized to a slab geometry Numata et al. (2010) [41]. The new code treats the linear terms in the ITEF equations implicitly while the nonlinear terms are treated explicitly. We show linear and nonlinear benchmark tests to prove the validity and applicability of the simulation code. Since the fast electron timescale is eliminated by the mass ratio expansion, the Courant-Friedrichs-Lewy condition is much less restrictive than in full gyrokinetic codes; the present hybrid code runs ∼ 2√{mi /me } ∼ 100 times faster than AstroGK with a single ion species and kinetic electrons where mi /me is the ion-electron mass ratio. The improvement of the computational time makes it feasible to execute ion scale gyrokinetic simulations with a high velocity space resolution and to run multiple simulations to determine the dependence of turbulent dynamics on parameters such as electron-ion temperature ratio and plasma beta.

  1. PEAK-TO-AVERAGE POWER RATIO REDUCTION USING CODING AND HYBRID TECHNIQUES FOR OFDM SYSTEM

    OpenAIRE

    Bahubali K. Shiragapur; Uday Wali

    2016-01-01

    In this article, the research work investigated is based on an error correction coding techniques are used to reduce the undesirable Peak-to-Average Power Ratio (PAPR) quantity. The Golay Code (24, 12), Reed-Muller code (16, 11), Hamming code (7, 4) and Hybrid technique (Combination of Signal Scrambling and Signal Distortion) proposed by us are used as proposed coding techniques, the simulation results shows that performance of Hybrid technique, reduces PAPR significantly as compared to Conve...

  2. The development of code benchmarks

    International Nuclear Information System (INIS)

    Glass, R.E.

    1986-01-01

    Sandia National Laboratories has undertaken a code benchmarking effort to define a series of cask-like problems having both numerical solutions and experimental data. The development of the benchmarks includes: (1) model problem definition, (2) code intercomparison, and (3) experimental verification. The first two steps are complete and a series of experiments are planned. The experiments will examine the elastic/plastic behavior of cylinders for both the end and side impacts resulting from a nine meter drop. The cylinders will be made from stainless steel and aluminum to give a range of plastic deformations. This paper presents the results of analyses simulating the model's behavior using materials properties for stainless steel and aluminum

  3. Development and validation of sodium fire codes

    International Nuclear Information System (INIS)

    Morii, Tadashi; Himeno Yoshiaki; Miyake, Osamu

    1989-01-01

    Development, verification, and validation of the spray fire code, SPRAY-3M, the pool fire codes, SOFIRE-M2 and SPM, the aerosol behavior code, ABC-INTG, and the simultaneous spray and pool fires code, ASSCOPS, are presented. In addition, the state-of-the-art of development of the multi-dimensional natural convection code, SOLFAS, for the analysis of heat-mass transfer during a fire, is presented. (author)

  4. Reactor safety computer code development at INEL

    International Nuclear Information System (INIS)

    Johnsen, G.W.

    1985-01-01

    This report provides a brief overview of the computer code development programs being conducted at EG and G Idaho, Inc. on behalf of US Nuclear Regulatory Commission and the Department of Energy, Idaho Operations Office. Included are descriptions of the codes being developed, their development status as of the date of this report, and resident code development expertise

  5. Influence of Code Size Variation on the Performance of 2D Hybrid ZCC/MD in OCDMA System

    Directory of Open Access Journals (Sweden)

    Matem Rima.

    2018-01-01

    Full Text Available Several two dimensional OCDMA have been developed in order to overcome many problems in optical network, enhancing cardinality, suppress Multiple Access Interference (MAI and mitigate Phase Induced Intensity Noise (PIIN. This paper propose a new 2D hybrid ZCC/MD code combining between 1D ZCC spectral encoding where M is its code length and 1D MD spatial spreading where N is its code length. The spatial spreading (N code length offers a good cardinality so it represents the main effect to enhance the performance of the system compared to the spectral (M code length according to the numerical results.

  6. Development of MCNP interface code in HFETR

    International Nuclear Information System (INIS)

    Qiu Liqing; Fu Rong; Deng Caiyu

    2007-01-01

    In order to describe the HFETR core with MCNP method, the interface code MCNPIP for HFETR and MCNP code is developed. This paper introduces the core DXSY and flowchart of MCNPIP code, and the handling of compositions of fuel elements and requirements on hardware and software. Finally, MCNPIP code is validated against the practical application. (authors)

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

    Directory of Open Access Journals (Sweden)

    Waqas A. Imtiaz

    2017-01-01

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

  8. Optical Code-Division Multiple-Access and Wavelength Division Multiplexing: Hybrid Scheme Review

    OpenAIRE

    P. Susthitha Menon; Sahbudin Shaari; Isaac A.M. Ashour; Hesham A. Bakarman

    2012-01-01

    Problem statement: Hybrid Optical Code-Division Multiple-Access (OCDMA) and Wavelength-Division Multiplexing (WDM) have flourished as successful schemes for expanding the transmission capacity as well as enhancing the security for OCDMA. However, a comprehensive review related to this hybrid system are lacking currently. Approach: The purpose of this paper is to review the literature on OCDMA-WDM overlay systems, including our hybrid approach of one-dimensional coding of SAC OCDMA with WDM si...

  9. Monte Carlo code development in Los Alamos

    International Nuclear Information System (INIS)

    Carter, L.L.; Cashwell, E.D.; Everett, C.J.; Forest, C.A.; Schrandt, R.G.; Taylor, W.M.; Thompson, W.L.; Turner, G.D.

    1974-01-01

    The present status of Monte Carlo code development at Los Alamos Scientific Laboratory is discussed. A brief summary is given of several of the most important neutron, photon, and electron transport codes. 17 references. (U.S.)

  10. Recent activities in accelerator code development

    International Nuclear Information System (INIS)

    Copper, R.K.; Ryne, R.D.

    1992-01-01

    In this paper we will review recent activities in the area of code development as it affects the accelerator community. We will first discuss the changing computing environment. We will review how the computing environment has changed in the last 10 years, with emphasis on computing power, operating systems, computer languages, graphics standards, and massively parallel processing. Then we will discuss recent code development activities in the areas of electromagnetics codes and beam dynamics codes

  11. Nuclear Characteristics of SPNDs and Preliminary Calculation of Hybrid Fixed Incore Detector with Monte Carlo Code

    International Nuclear Information System (INIS)

    Koo, Bon Seung; Lee, Kyung Hoon; Song, Jae Seung; Park, Sang Yoon

    2013-01-01

    In this paper, the basic nuclear characteristics of major emitter materials were surveyed. In addition, preliminary calculations of Cobalt-Vanadium fixed incore detector were performed using the Monte Carlo code. Calculational results were cross-checked by KARMA. KARMA is a two-dimensional multigroup transport theory code developed by the KAERI and approved by Korean regularity agency to be employed as a nuclear design tool for a Korean commercial pressurizer water reactor. The nuclear characteristics of the major emitter materials were surveyed, and preliminary calculations of the hybrid fixed incore detector were performed with the MCNP code. The eigenvalue and pin-by-pin fission power distributions were calculated and showed good agreement with the KARMA calculation results. As future work, gamma power distributions as well as several types of XS of the emitter, insulator, and collector regions for a Co-V ICI assembly will be evaluated and compared

  12. Status of SPACE Safety Analysis Code Development

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  13. Development and application of the waste code

    International Nuclear Information System (INIS)

    Morison, I.W.

    1984-01-01

    This paper discusses the objectives and general approach underlying the Australian Code of Practice on the Management of Radioactive Wastes arising from the Mining and Milling of Radioactive Ores 1982. Background to the development of the Code is provided and the guidelines which supplement the Code are considered

  14. Development of computer code in PNC, 3

    International Nuclear Information System (INIS)

    Ohtaki, Akira; Ohira, Hiroaki

    1990-01-01

    Super-COPD, a code which is integrated by calculation modules, has been developed in order to evaluate kinds of dynamics of LMFBR plant by improving COPD. The code involves all models and its advanced models of COPD in module structures. The code makes it possible to simulate the system dynamics of LMFBR plant of any configurations and components. (author)

  15. Collaborative Multi-Layer Network Coding in Hybrid Cellular Cognitive Radio Networks

    KAUST Repository

    Moubayed, Abdallah J.; Sorour, Sameh; Al-Naffouri, Tareq Y.; Alouini, Mohamed-Slim

    2015-01-01

    In this paper, as an extension to [1], we propose a prioritized multi-layer network coding scheme for collaborative packet recovery in hybrid (interweave and underlay) cellular cognitive radio networks. This scheme allows the uncoordinated

  16. Collaborative Multi-Layer Network Coding For Hybrid Cellular Cognitive Radio Networks

    KAUST Repository

    Moubayed, Abdallah J.

    2014-01-01

    In this thesis, as an extension to [1], we propose a prioritized multi-layer network coding scheme for collaborative packet recovery in hybrid (interweave and underlay) cellular cognitive radio networks. This scheme allows the uncoordinated

  17. Preliminary investigation study of code of developed country for developing Korean fuel cycle code

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Ko, Won Il; Lee, Ho Hee; Cho, Dong Keun; Park, Chang Je

    2012-01-01

    In order to develop Korean fuel cycle code, the analyses has been performed with the fuel cycle codes which are used in advanced country. Also, recommendations were proposed for future development. The fuel cycle codes are AS FLOOWS: VISTA which has been developed by IAEA, DANESS code which developed by ANL and LISTO, and VISION developed by INL for the Advanced Fuel Cycle Initiative (AFCI) system analysis. The recommended items were proposed for software, program scheme, material flow model, isotope decay model, environmental impact analysis model, and economics analysis model. The described things will be used for development of Korean nuclear fuel cycle code in future

  18. Low complexity source and channel coding for mm-wave hybrid fiber-wireless links

    DEFF Research Database (Denmark)

    Lebedev, Alexander; Vegas Olmos, Juan José; Pang, Xiaodan

    2014-01-01

    We report on the performance of channel and source coding applied for an experimentally realized hybrid fiber-wireless W-band link. Error control coding performance is presented for a wireless propagation distance of 3 m and 20 km fiber transmission. We report on peak signal-to-noise ratio perfor...

  19. NASA space radiation transport code development consortium

    International Nuclear Information System (INIS)

    Townsend, L. W.

    2005-01-01

    Recently, NASA established a consortium involving the Univ. of Tennessee (lead institution), the Univ. of Houston, Roanoke College and various government and national laboratories, to accelerate the development of a standard set of radiation transport computer codes for NASA human exploration applications. This effort involves further improvements of the Monte Carlo codes HETC and FLUKA and the deterministic code HZETRN, including developing nuclear reaction databases necessary to extend the Monte Carlo codes to carry out heavy ion transport, and extending HZETRN to three dimensions. The improved codes will be validated by comparing predictions with measured laboratory transport data, provided by an experimental measurements consortium, and measurements in the upper atmosphere on the balloon-borne Deep Space Test Bed (DSTB). In this paper, we present an overview of the consortium members and the current status and future plans of consortium efforts to meet the research goals and objectives of this extensive undertaking. (authors)

  20. Study on MPI/OpenMP hybrid parallelism for Monte Carlo neutron transport code

    International Nuclear Information System (INIS)

    Liang Jingang; Xu Qi; Wang Kan; Liu Shiwen

    2013-01-01

    Parallel programming with mixed mode of messages-passing and shared-memory has several advantages when used in Monte Carlo neutron transport code, such as fitting hardware of distributed-shared clusters, economizing memory demand of Monte Carlo transport, improving parallel performance, and so on. MPI/OpenMP hybrid parallelism was implemented based on a one dimension Monte Carlo neutron transport code. Some critical factors affecting the parallel performance were analyzed and solutions were proposed for several problems such as contention access, lock contention and false sharing. After optimization the code was tested finally. It is shown that the hybrid parallel code can reach good performance just as pure MPI parallel program, while it saves a lot of memory usage at the same time. Therefore hybrid parallel is efficient for achieving large-scale parallel of Monte Carlo neutron transport. (authors)

  1. CATHARE code development and assessment methodologies

    International Nuclear Information System (INIS)

    Micaelli, J.C.; Barre, F.; Bestion, D.

    1995-01-01

    The CATHARE thermal-hydraulic code has been developed jointly by Commissariat a l'Energie Atomique (CEA), Electricite de France (EdF), and Framatorne for safety analysis. Since the beginning of the project (September 1979), development and assessment activities have followed a methodology supported by two series of experimental tests: separate effects tests and integral effects tests. The purpose of this paper is to describe this methodology, the code assessment status, and the evolution to take into account two new components of this program: the modeling of three-dimensional phenomena and the requirements of code uncertainty evaluation

  2. Computer code development plant for SMART design

    International Nuclear Information System (INIS)

    Bae, Kyoo Hwan; Choi, S.; Cho, B.H.; Kim, K.K.; Lee, J.C.; Kim, J.P.; Kim, J.H.; Chung, M.; Kang, D.J.; Chang, M.H.

    1999-03-01

    In accordance with the localization plan for the nuclear reactor design driven since the middle of 1980s, various computer codes have been transferred into the korea nuclear industry through the technical transfer program from the worldwide major pressurized water reactor supplier or through the international code development program. These computer codes have been successfully utilized in reactor and reload core design works. As the results, design- related technologies have been satisfactorily accumulated. However, the activities for the native code development activities to substitute the some important computer codes of which usages are limited by the original technique owners have been carried out rather poorly. Thus, it is most preferentially required to secure the native techniques on the computer code package and analysis methodology in order to establish the capability required for the independent design of our own model of reactor. Moreover, differently from the large capacity loop-type commercial reactors, SMART (SYSTEM-integrated Modular Advanced ReacTor) design adopts a single reactor pressure vessel containing the major primary components and has peculiar design characteristics such as self-controlled gas pressurizer, helical steam generator, passive residual heat removal system, etc. Considering those peculiar design characteristics for SMART, part of design can be performed with the computer codes used for the loop-type commercial reactor design. However, most of those computer codes are not directly applicable to the design of an integral reactor such as SMART. Thus, they should be modified to deal with the peculiar design characteristics of SMART. In addition to the modification efforts, various codes should be developed in several design area. Furthermore, modified or newly developed codes should be verified their reliability through the benchmarking or the test for the object design. Thus, it is necessary to proceed the design according to the

  3. Computer code development plant for SMART design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Kyoo Hwan; Choi, S.; Cho, B.H.; Kim, K.K.; Lee, J.C.; Kim, J.P.; Kim, J.H.; Chung, M.; Kang, D.J.; Chang, M.H

    1999-03-01

    In accordance with the localization plan for the nuclear reactor design driven since the middle of 1980s, various computer codes have been transferred into the korea nuclear industry through the technical transfer program from the worldwide major pressurized water reactor supplier or through the international code development program. These computer codes have been successfully utilized in reactor and reload core design works. As the results, design- related technologies have been satisfactorily accumulated. However, the activities for the native code development activities to substitute the some important computer codes of which usages are limited by the original technique owners have been carried out rather poorly. Thus, it is most preferentially required to secure the native techniques on the computer code package and analysis methodology in order to establish the capability required for the independent design of our own model of reactor. Moreover, differently from the large capacity loop-type commercial reactors, SMART (SYSTEM-integrated Modular Advanced ReacTor) design adopts a single reactor pressure vessel containing the major primary components and has peculiar design characteristics such as self-controlled gas pressurizer, helical steam generator, passive residual heat removal system, etc. Considering those peculiar design characteristics for SMART, part of design can be performed with the computer codes used for the loop-type commercial reactor design. However, most of those computer codes are not directly applicable to the design of an integral reactor such as SMART. Thus, they should be modified to deal with the peculiar design characteristics of SMART. In addition to the modification efforts, various codes should be developed in several design area. Furthermore, modified or newly developed codes should be verified their reliability through the benchmarking or the test for the object design. Thus, it is necessary to proceed the design according to the

  4. Development of ADINA-J-integral code

    International Nuclear Information System (INIS)

    Kurihara, Ryoichi

    1988-07-01

    A general purpose finite element program ADINA (Automatic Dynamic Incremental Nonlinear Analysis), which was developed by Bathe et al., was revised to be able to calculate the J- and J-integral. This report introduced the numerical method to add this capability to the code, and the evaluation of the revised ADINA-J code by using a few of examples of the J estimation model, i.e. a compact tension specimen, a center cracked panel subjected to dynamic load, and a thick shell cylinder having inner axial crack subjected to thermal load. The evaluation testified the function of the revised code. (author)

  5. Development of HTGR plant dynamics simulation code

    International Nuclear Information System (INIS)

    Ohashi, Kazutaka; Tazawa, Yujiro; Mitake, Susumu; Suzuki, Katsuo.

    1987-01-01

    Plant dynamics simulation analysis plays an important role in the design work of nuclear power plant especially in the plant safety analysis, control system analysis, and transient condition analysis. The authors have developed the plant dynamics simulation code named VESPER, which is applicable to the design work of High Temperature Engineering Test Reactor, and have been improving the code corresponding to the design changes made in the subsequent design works. This paper describes the outline of VESPER code and shows its sample calculation results selected from the recent design work. (author)

  6. PEAK-TO-AVERAGE POWER RATIO REDUCTION USING CODING AND HYBRID TECHNIQUES FOR OFDM SYSTEM

    Directory of Open Access Journals (Sweden)

    Bahubali K. Shiragapur

    2016-03-01

    Full Text Available In this article, the research work investigated is based on an error correction coding techniques are used to reduce the undesirable Peak-to-Average Power Ratio (PAPR quantity. The Golay Code (24, 12, Reed-Muller code (16, 11, Hamming code (7, 4 and Hybrid technique (Combination of Signal Scrambling and Signal Distortion proposed by us are used as proposed coding techniques, the simulation results shows that performance of Hybrid technique, reduces PAPR significantly as compared to Conventional and Modified Selective mapping techniques. The simulation results are validated through statistical properties, for proposed technique’s autocorrelation value is maximum shows reduction in PAPR. The symbol preference is the key idea to reduce PAPR based on Hamming distance. The simulation results are discussed in detail, in this article.

  7. Discrete fracture network code development

    Energy Technology Data Exchange (ETDEWEB)

    Dershowitz, W.; Doe, T.; Shuttle, D.; Eiben, T.; Fox, A.; Emsley, S.; Ahlstrom, E. [Golder Associates Inc., Redmond, Washington (United States)

    1999-02-01

    This report presents the results of fracture flow model development and application performed by Golder Associates Inc. during the fiscal year 1998. The primary objective of the Golder Associates work scope was to provide theoretical and modelling support to the JNC performance assessment effort in fiscal year 2000. In addition, Golder Associates provided technical support to JNC for the Aespoe project. Major efforts for performance assessment support included extensive flow and transport simulations, analysis of pathway simplification, research on excavation damage zone effects, software verification and cross-verification, and analysis of confidence bounds on Monte Carlo simulations. In addition, a Fickian diffusion algorithm was implemented for Laplace Transform Galerkin solute transport. Support for the Aespoe project included predictive modelling of sorbing tracer transport in the TRUE-1 rock block, analysis of 1 km geochemical transport pathways for Task 5', and data analysis and experimental design for the TRUE Block Scale experiment. Technical information about Golder Associates support to JNC is provided in the appendices to this report. (author)

  8. 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

  9. 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

  10. A new hybrid code (CHIEF) implementing the inertial electron fluid equation without approximation

    Science.gov (United States)

    Muñoz, P. A.; Jain, N.; Kilian, P.; Büchner, J.

    2018-03-01

    We present a new hybrid algorithm implemented in the code CHIEF (Code Hybrid with Inertial Electron Fluid) for simulations of electron-ion plasmas. The algorithm treats the ions kinetically, modeled by the Particle-in-Cell (PiC) method, and electrons as an inertial fluid, modeled by electron fluid equations without any of the approximations used in most of the other hybrid codes with an inertial electron fluid. This kind of code is appropriate to model a large variety of quasineutral plasma phenomena where the electron inertia and/or ion kinetic effects are relevant. We present here the governing equations of the model, how these are discretized and implemented numerically, as well as six test problems to validate our numerical approach. Our chosen test problems, where the electron inertia and ion kinetic effects play the essential role, are: 0) Excitation of parallel eigenmodes to check numerical convergence and stability, 1) parallel (to a background magnetic field) propagating electromagnetic waves, 2) perpendicular propagating electrostatic waves (ion Bernstein modes), 3) ion beam right-hand instability (resonant and non-resonant), 4) ion Landau damping, 5) ion firehose instability, and 6) 2D oblique ion firehose instability. Our results reproduce successfully the predictions of linear and non-linear theory for all these problems, validating our code. All properties of this hybrid code make it ideal to study multi-scale phenomena between electron and ion scales such as collisionless shocks, magnetic reconnection and kinetic plasma turbulence in the dissipation range above the electron scales.

  11. Studies of Planet Formation using a Hybrid N-body + Planetesimal Code

    Science.gov (United States)

    Kenyon, Scott J.; Bromley, Benjamin C.; Salamon, Michael (Technical Monitor)

    2005-01-01

    The goal of our proposal was to use a hybrid multi-annulus planetesimal/n-body code to examine the planetesimal theory, one of the two main theories of planet formation. We developed this code to follow the evolution of numerous 1 m to 1 km planetesimals as they collide, merge, and grow into full-fledged planets. Our goal was to apply the code to several well-posed, topical problems in planet formation and to derive observational consequences of the models. We planned to construct detailed models to address two fundamental issues: 1) icy planets - models for icy planet formation will demonstrate how the physical properties of debris disks, including the Kuiper Belt in our solar system, depend on initial conditions and input physics; and 2) terrestrial planets - calculations following the evolution of 1-10 km planetesimals into Earth-mass planets and rings of dust will provide a better understanding of how terrestrial planets form and interact with their environment. During the past year, we made progress on each issue. Papers published in 2004 are summarized. Summaries of work to be completed during the first half of 2005 and work planned for the second half of 2005 are included.

  12. Noise suppression system of OCDMA with spectral/spatial 2D hybrid code

    Science.gov (United States)

    Matem, Rima; Aljunid, S. A.; Junita, M. N.; Rashidi, C. B. M.; Shihab Aqrab, Israa

    2017-11-01

    In this paper, we propose a novel 2D spectral/spatial hybrid code based on 1D ZCC and 1D MD where the both present a zero cross correlation property analyzed and the influence of the noise of optical as Phase Induced Intensity Noise (PIIN), shot and thermal noise. This new code is shown effectively to mitigate the PIIN and suppresses MAI. Using 2D ZCC/MD code the performance of the system can be improved in term of as well as to support more simultaneous users compared of the 2D FCC/MDW and 2D DPDC codes.

  13. Noise suppression system of OCDMA with spectral/spatial 2D hybrid code

    Directory of Open Access Journals (Sweden)

    Matem Rima

    2017-01-01

    Full Text Available In this paper, we propose a novel 2D spectral/spatial hybrid code based on 1D ZCC and 1D MD where the both present a zero cross correlation property analyzed and the influence of the noise of optical as Phase Induced Intensity Noise (PIIN, shot and thermal noise. This new code is shown effectively to mitigate the PIIN and suppresses MAI. Using 2D ZCC/MD code the performance of the system can be improved in term of as well as to support more simultaneous users compared of the 2D FCC/MDW and 2D DPDC codes.

  14. 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)

  15. Health Code Number (HCN) Development Procedure

    Energy Technology Data Exchange (ETDEWEB)

    Petrocchi, Rocky; Craig, Douglas K.; Bond, Jayne-Anne; Trott, Donna M.; Yu, Xiao-Ying

    2013-09-01

    This report provides the detailed description of health code numbers (HCNs) and the procedure of how each HCN is assigned. It contains many guidelines and rationales of HCNs. HCNs are used in the chemical mixture methodology (CMM), a method recommended by the department of energy (DOE) for assessing health effects as a result of exposures to airborne aerosols in an emergency. The procedure is a useful tool for proficient HCN code developers. Intense training and quality assurance with qualified HCN developers are required before an individual comprehends the procedure to develop HCNs for DOE.

  16. Development of Coolant Radioactivity Interpretation Code

    International Nuclear Information System (INIS)

    Kim, Kiyoung; Jung, Youngsuk; Kim, Kyounghyun; Kim, Jangwook

    2013-01-01

    In Korea, the coolant radioactivity analysis has been performed by using the computer codes of foreign companies such as CADE (Westinghouse), IODYNE and CESIUM (ABB-CE). However, these computer codes are too conservative and have involved considerable errors. Furthermore, since these codes are DOS-based program, their easy operability is not satisfactory. Therefore it is required development of an enhanced analysis algorithm applying an analytical method reflecting the change of operational environments of domestic nuclear power plants and a fuel failure evaluation software considering user' conveniences. We have developed a nuclear fuel failure evaluation code able to estimate the number of failed fuel rods and the burn-up of failed fuels during nuclear power plant operation cycle. A Coolant Radio-activity Interpretation Code (CRIC) for LWR has been developed as the output of the project 'Development of Fuel Reliability Enhanced Technique' organized by Korea Institute of Energy Technology Evaluation and Planning (KETEP). The CRIC is Windows based-software able to evaluate the number of failed fuel rods and the burn-up of failed fuel region by analyzing coolant radioactivity of LWR in operation. The CRIC is based on the model of fission products release commonly known as 'three region model' (pellet region, gap region, and coolant region), and we are verifying the CRIC results based on the cases of domestic fuel failures. CRIC users are able to estimate the number of failed fuel rods, burn-up and regions of failed fuel considered enrichment and power distribution of fuel region by using operational cycle data, coolant activity data, fuel loading pattern, Cs-134/Cs-137 ratio according to burn-up and U-235 enrichment provided in the code. Due to development of the CRIC, it is secured own unique fuel failure evaluation code. And, it is expected to have the following significant meaning. This is that the code reflecting a proprietary technique for quantitatively

  17. Hybrid real-code ant colony optimisation for constrained mechanical design

    Science.gov (United States)

    Pholdee, Nantiwat; Bureerat, Sujin

    2016-01-01

    This paper proposes a hybrid meta-heuristic based on integrating a local search simplex downhill (SDH) method into the search procedure of real-code ant colony optimisation (ACOR). This hybridisation leads to five hybrid algorithms where a Monte Carlo technique, a Latin hypercube sampling technique (LHS) and a translational propagation Latin hypercube design (TPLHD) algorithm are used to generate an initial population. Also, two numerical schemes for selecting an initial simplex are investigated. The original ACOR and its hybrid versions along with a variety of established meta-heuristics are implemented to solve 17 constrained test problems where a fuzzy set theory penalty function technique is used to handle design constraints. The comparative results show that the hybrid algorithms are the top performers. Using the TPLHD technique gives better results than the other sampling techniques. The hybrid optimisers are a powerful design tool for constrained mechanical design problems.

  18. Development of code PRETOR for stellarator simulation

    International Nuclear Information System (INIS)

    Dies, J.; Fontanet, J.; Fontdecaba, J.M.; Castejon, F.; Alejandre, C.

    1998-01-01

    The Department de Fisica i Enginyeria Nuclear (DFEN) of the UPC has some experience in the development of the transport code PRETOR. This code has been validated with shots of DIII-D, JET and TFTR, it has also been used in the simulation of operational scenarios of ITER fast burnt termination. Recently, the association EURATOM-CIEMAT has started the operation of the TJ-II stellarator. Due to the need of validating the results given by others transport codes applied to stellarators and because all of them made some approximations, as a averaging magnitudes in each magnetic surface, it was thought suitable to adapt the PRETOR code to devices without axial symmetry, like stellarators, which is very suitable for the specific needs of the study of TJ-II. Several modifications are required in PRETOR; the main concerns to the models of: magnetic equilibrium, geometry and transport of energy and particles. In order to solve the complex magnetic equilibrium geometry the powerful numerical code VMEC has been used. This code gives the magnetic surface shape as a Fourier series in terms of the harmonics (m,n). Most of the geometric magnitudes are also obtained from the VMEC results file. The energy and particle transport models will be replaced by other phenomenological models that are better adapted to stellarator simulation. Using the proposed models, it is pretended to reproduce experimental data available from present stellarators, given especial attention to the TJ-II of the association EURATOM-CIEMAT. (Author)

  19. Coupled geochemical and solute transport code development

    International Nuclear Information System (INIS)

    Morrey, J.R.; Hostetler, C.J.

    1985-01-01

    A number of coupled geochemical hydrologic codes have been reported in the literature. Some of these codes have directly coupled the source-sink term to the solute transport equation. The current consensus seems to be that directly coupling hydrologic transport and chemical models through a series of interdependent differential equations is not feasible for multicomponent problems with complex geochemical processes (e.g., precipitation/dissolution reactions). A two-step process appears to be the required method of coupling codes for problems where a large suite of chemical reactions must be monitored. Two-step structure requires that the source-sink term in the transport equation is supplied by a geochemical code rather than by an analytical expression. We have developed a one-dimensional two-step coupled model designed to calculate relatively complex geochemical equilibria (CTM1D). Our geochemical module implements a Newton-Raphson algorithm to solve heterogeneous geochemical equilibria, involving up to 40 chemical components and 400 aqueous species. The geochemical module was designed to be efficient and compact. A revised version of the MINTEQ Code is used as a parent geochemical code

  20. A HYDROCHEMICAL HYBRID CODE FOR ASTROPHYSICAL PROBLEMS. I. CODE VERIFICATION AND BENCHMARKS FOR A PHOTON-DOMINATED REGION (PDR)

    International Nuclear Information System (INIS)

    Motoyama, Kazutaka; Morata, Oscar; Hasegawa, Tatsuhiko; Shang, Hsien; Krasnopolsky, Ruben

    2015-01-01

    A two-dimensional hydrochemical hybrid code, KM2, is constructed to deal with astrophysical problems that would require coupled hydrodynamical and chemical evolution. The code assumes axisymmetry in a cylindrical coordinate system and consists of two modules: a hydrodynamics module and a chemistry module. The hydrodynamics module solves hydrodynamics using a Godunov-type finite volume scheme and treats included chemical species as passively advected scalars. The chemistry module implicitly solves nonequilibrium chemistry and change of energy due to thermal processes with transfer of external ultraviolet radiation. Self-shielding effects on photodissociation of CO and H 2 are included. In this introductory paper, the adopted numerical method is presented, along with code verifications using the hydrodynamics module and a benchmark on the chemistry module with reactions specific to a photon-dominated region (PDR). Finally, as an example of the expected capability, the hydrochemical evolution of a PDR is presented based on the PDR benchmark

  1. Developing an Australian code of construction ethics

    Directory of Open Access Journals (Sweden)

    Sean Francis McCarthy

    2012-05-01

    Full Text Available This article looks at the increasing need to consider the role of ethics in construction. The industry, historically, has been challenged by allegations of a serious shortfall in ethical standards. Only limited attempts to date in Australia have been made to address that concern. Any ethical analysis should consider the definition of ethics and its historical development. This paper considers major historical developments in ethical thinking as well as contemporary thinking on ethics for professional sub-sets. A code could be developed specific to construction. Current methods of addressing ethics in construction and in other industries are also reviewed. This paper argues that developing a code of ethics, supported by other measures is the way forward. The author’s aim is to promote further discussion and promote the drafting of a code. This paper includes a summary of other ethical codes that may provide a starting point. The time for reform is upon us, and there is an urgent need for an independent body to take the lead, for fear of floundering and having only found ‘another debating topic’ (Uff 2006.

  2. Multiple application coded switch development report

    International Nuclear Information System (INIS)

    Bernal, E.L.; Kestly, J.D.

    1979-03-01

    The development of the Multiple Application Coded Switch (MACS) and its related controller are documented; the functional and electrical characteristics are described; the interface requirements defined, and a troubleshooting guide provided. The system was designed for the Safe Secure Trailer System used for secure transportation of nuclear material

  3. Parallel Computing Characteristics of CUPID code under MPI and Hybrid environment

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Ryong; Yoon, Han Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Jeon, Byoung Jin; Choi, Hyoung Gwon [Seoul National Univ. of Science and Technology, Seoul (Korea, Republic of)

    2014-05-15

    In this paper, a characteristic of parallel algorithm is presented for solving an elliptic type equation of CUPID via domain decomposition method using the MPI and the parallel performance is estimated in terms of a scalability which shows the speedup ratio. In addition, the time-consuming pattern of major subroutines is studied. Two different grid systems are taken into account: 40,000 meshes for coarse system and 320,000 meshes for fine system. Since the matrix of the CUPID code differs according to whether the flow is single-phase or two-phase, the effect of matrix shape is evaluated. Finally, the effect of the preconditioner for matrix solver is also investigated. Finally, the hybrid (OpenMP+MPI) parallel algorithm is introduced and discussed in detail for solving pressure solver. Component-scale thermal-hydraulics code, CUPID has been developed for two-phase flow analysis, which adopts a three-dimensional, transient, three-field model, and parallelized to fulfill a recent demand for long-transient and highly resolved multi-phase flow behavior. In this study, the parallel performance of the CUPID code was investigated in terms of scalability. The CUPID code was parallelized with domain decomposition method. The MPI library was adopted to communicate the information at the neighboring domain. For managing the sparse matrix effectively, the CSR storage format is used. To take into account the characteristics of the pressure matrix which turns to be asymmetric for two-phase flow, both single-phase and two-phase calculations were run. In addition, the effect of the matrix size and preconditioning was also investigated. The fine mesh calculation shows better scalability than the coarse mesh because the number of coarse mesh does not need to decompose the computational domain excessively. The fine mesh can be present good scalability when dividing geometry with considering the ratio between computation and communication time. For a given mesh, single-phase flow

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  5. Development of chemical equilibrium analysis code 'CHEEQ'

    International Nuclear Information System (INIS)

    Nagai, Shuichiro

    2006-08-01

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

  6. Hybrid code simulation on mode conversion in the second harmonic ICRF heating

    International Nuclear Information System (INIS)

    Sakai, K.; Takeuchi, S.; Matsumoto, M.; Sugihara, R.

    1985-01-01

    ICRF second harmonic heating of a single-species plasma is studied by using a 1-1/2 dimensional quasi-neutral hybrid code. Mode conversion, transmission and reflection of the magnetosonic waves are confirmed, both for the high- and low-field-side excitations. The ion heating by waves propagating perpendicularly to the static magnetic field is also observed

  7. Development of Evaluation Code for MUF Uncertainty

    International Nuclear Information System (INIS)

    Won, Byung Hee; Han, Bo Young; Shin, Hee Sung; Ahn, Seong-Kyu; Park, Geun-Il; Park, Se Hwan

    2015-01-01

    Material Unaccounted For (MUF) is the material balance evaluated by measured nuclear material in a Material Balance Area (MBA). Assuming perfect measurements and no diversion from a facility, one can expect a zero MUF. However, non-zero MUF is always occurred because of measurement uncertainty even though the facility is under normal operation condition. Furthermore, there are many measurements using different equipment at various Key Measurement Points (KMPs), and the MUF uncertainty is affected by errors of those measurements. Evaluating MUF uncertainty is essentially required to develop safeguards system including nuclear measurement system in pyroprocessing, which is being developed for reducing radioactive waste from spent fuel in Korea Atomic Energy Research Institute (KAERI). The evaluation code for analyzing MUF uncertainty has been developed and it was verified using sample problem from the IAEA reference. MUF uncertainty can be simply and quickly calculated by using this evaluation code which is made based on graphical user interface for user friendly. It is also expected that the code will make the sensitivity analysis on the MUF uncertainty for the various safeguards systems easy and more systematic. It is suitable for users who want to evaluate the conventional safeguards system as well as to develop a new system for developing facilities

  8. Development of Evaluation Code for MUF Uncertainty

    Energy Technology Data Exchange (ETDEWEB)

    Won, Byung Hee; Han, Bo Young; Shin, Hee Sung; Ahn, Seong-Kyu; Park, Geun-Il; Park, Se Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Material Unaccounted For (MUF) is the material balance evaluated by measured nuclear material in a Material Balance Area (MBA). Assuming perfect measurements and no diversion from a facility, one can expect a zero MUF. However, non-zero MUF is always occurred because of measurement uncertainty even though the facility is under normal operation condition. Furthermore, there are many measurements using different equipment at various Key Measurement Points (KMPs), and the MUF uncertainty is affected by errors of those measurements. Evaluating MUF uncertainty is essentially required to develop safeguards system including nuclear measurement system in pyroprocessing, which is being developed for reducing radioactive waste from spent fuel in Korea Atomic Energy Research Institute (KAERI). The evaluation code for analyzing MUF uncertainty has been developed and it was verified using sample problem from the IAEA reference. MUF uncertainty can be simply and quickly calculated by using this evaluation code which is made based on graphical user interface for user friendly. It is also expected that the code will make the sensitivity analysis on the MUF uncertainty for the various safeguards systems easy and more systematic. It is suitable for users who want to evaluate the conventional safeguards system as well as to develop a new system for developing facilities.

  9. Sensitivity analysis of the titan hybrid deterministic transport code for SPECT simulation

    International Nuclear Information System (INIS)

    Royston, Katherine K.; Haghighat, Alireza

    2011-01-01

    Single photon emission computed tomography (SPECT) has been traditionally simulated using Monte Carlo methods. The TITAN code is a hybrid deterministic transport code that has recently been applied to the simulation of a SPECT myocardial perfusion study. For modeling SPECT, the TITAN code uses a discrete ordinates method in the phantom region and a combined simplified ray-tracing algorithm with a fictitious angular quadrature technique to simulate the collimator and generate projection images. In this paper, we compare the results of an experiment with a physical phantom with predictions from the MCNP5 and TITAN codes. While the results of the two codes are in good agreement, they differ from the experimental data by ∼ 21%. In order to understand these large differences, we conduct a sensitivity study by examining the effect of different parameters including heart size, collimator position, collimator simulation parameter, and number of energy groups. (author)

  10. Development of steam explosion simulation code JASMINE

    Energy Technology Data Exchange (ETDEWEB)

    Moriyama, Kiyofumi; Yamano, Norihiro; Maruyama, Yu; Kudo, Tamotsu; Sugimoto, Jun [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nagano, Katsuhiro; Araki, Kazuhiro

    1995-11-01

    A steam explosion is considered as a phenomenon which possibly threatens the integrity of the containment vessel of a nuclear power plant in a severe accident condition. A numerical calculation code JASMINE (JAeri Simulator for Multiphase INteraction and Explosion) purposed to simulate the whole process of steam explosions has been developed. The premixing model is based on a multiphase flow simulation code MISTRAL by Fuji Research Institute Co. In JASMINE code, the constitutive equations and the flow regime map are modified for the simulation of premixing related phenomena. The numerical solution method of the original code is succeeded, i.e. the basic equations are discretized semi-implicitly, BCGSTAB method is used for the matrix solver to improve the stability and convergence, also TVD scheme is applied to capture a steep phase distribution accurately. Test calculations have been performed for the conditions correspond to the experiments by Gilbertson et al. and Angelini et al. in which mixing of solid particles and water were observed in iso-thermal condition and with boiling, respectively. (author).

  11. Development of a tritium dispersion code

    International Nuclear Information System (INIS)

    Bell, R.P.; Davis, M.W.; Joseph, S.; Wong, K.Y.

    1985-01-01

    This paper describes the development and verification of a computer code designed to calculate the radiation dose to man following acute or chronic atmospheric releases of tritium gas and oxide from a point source. The Ontario Hydro Tritium Dispersion Code calculates tritium concentrations in air, soil, and vegetation and doses to man resulting from inhalation/immersion and ingestion of food, milk meat and water. The deposition of HT to soil, conversion of HT to HTO by soil enzymes and resuspension of HTO to air have been incorporated into the terrestrial compartment model and are unique features of the code. Sensitivity analysis has identified the HT deposition velocity and the equivalent water depth of the vegetation compartment as two parameters which have a strong influence on dose calculations. Tritium concentrations in vegetation and soil calculated by the code were in reasonable agreement with experimental results. The radiological significance of including the mechanisms of HT to HTO conversion and resuspension of HTO to air is illustrated

  12. Development of steam explosion simulation code JASMINE

    International Nuclear Information System (INIS)

    Moriyama, Kiyofumi; Yamano, Norihiro; Maruyama, Yu; Kudo, Tamotsu; Sugimoto, Jun; Nagano, Katsuhiro; Araki, Kazuhiro.

    1995-11-01

    A steam explosion is considered as a phenomenon which possibly threatens the integrity of the containment vessel of a nuclear power plant in a severe accident condition. A numerical calculation code JASMINE (JAeri Simulator for Multiphase INteraction and Explosion) purposed to simulate the whole process of steam explosions has been developed. The premixing model is based on a multiphase flow simulation code MISTRAL by Fuji Research Institute Co. In JASMINE code, the constitutive equations and the flow regime map are modified for the simulation of premixing related phenomena. The numerical solution method of the original code is succeeded, i.e. the basic equations are discretized semi-implicitly, BCGSTAB method is used for the matrix solver to improve the stability and convergence, also TVD scheme is applied to capture a steep phase distribution accurately. Test calculations have been performed for the conditions correspond to the experiments by Gilbertson et al. and Angelini et al. in which mixing of solid particles and water were observed in iso-thermal condition and with boiling, respectively. (author)

  13. Towards Product Lining Model-Driven Development Code Generators

    OpenAIRE

    Roth, Alexander; Rumpe, Bernhard

    2015-01-01

    A code generator systematically transforms compact models to detailed code. Today, code generation is regarded as an integral part of model-driven development (MDD). Despite its relevance, the development of code generators is an inherently complex task and common methodologies and architectures are lacking. Additionally, reuse and extension of existing code generators only exist on individual parts. A systematic development and reuse based on a code generator product line is still in its inf...

  14. 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.

  15. Development and assessment of the CONTAIN hybrid flow solver

    International Nuclear Information System (INIS)

    Murata, K.K.; Stamps, D.W.

    1996-11-01

    A new gravitational head formulation for the treatment of stratified conditions has been developed for CONTAIN 1.2, a control volume code used primarily for the analyses of postulated accidents in nuclear power plants. The new CONTAIN formulation of gravitational heads, termed the hybrid formulation, is described. This method of calculating stratified conditions is compared with the old, average-density formulation used in code versions prior to CONTAIN 1.2. Both formulations are assessed in this report with experimental data from three large-scale experiments in which stratified conditions formed by injection of a buoyant gas were observed. In general, the hybrid formulation gives a substantially higher degree of stratification than the old formulation. For stable, fully developed stratifications, the hybrid formulation also gives much better agreement with the measured degree of stratification than the old formulation. In addition, the predicted degree of stratification is robust and not sensitive to nodalization, provided a set of nodalization guidelines are followed. However, for stratification behavior controlled by special physics not modeled in CONTAIN, such as momentum convection, plume entrainment, or bulk molecular diffusion, one should not expect good agreement with experiment unless special measures to accommodate the missing physics are taken

  16. Use of a hybrid code for global-scale plasma simulation

    International Nuclear Information System (INIS)

    Swift, D.W.

    1996-01-01

    This paper presents a demonstration of the use of a hybrid code to model the Earth's magnetosphere on a global scale. The typical hybrid code calculates the interaction of fully kinetic ions and a massless electron fluid with the magnetic field. This code also includes a fluid ion component to approximate the cold ionospheric plasma that spatially overlaps with the discrete particle component. Other innovative features of the code include a numerically generated curvilinear coordinate system and subcycling of the magnetic field update to the particle push. These innovations allow the code to accommodate disparate time and distance scales. The demonstration is a simulation of the noon meridian plane of the magnetosphere. The code exhibits the formation of fast and slow-mode shocks and tearing reconnection at the magnetopause. New results include particle acceleration in the cusp and nearly field aligned currents linking the cusp and polar ionosphere. The paper also describes a density depletion instability and measures to avoid it. 27 refs., 4 figs

  17. Using Coding Apps to Support Literacy Instruction and Develop Coding Literacy

    Science.gov (United States)

    Hutchison, Amy; Nadolny, Larysa; Estapa, Anne

    2016-01-01

    In this article the authors present the concept of Coding Literacy and describe the ways in which coding apps can support the development of Coding Literacy and disciplinary and digital literacy skills. Through detailed examples, we describe how coding apps can be integrated into literacy instruction to support learning of the Common Core English…

  18. Development of Nuclear Energy Security Code

    International Nuclear Information System (INIS)

    Shimamura, Takehisa; Suzuki, Atsuyuki; Okubo, Hiroo; Kikuchi, Masahiro.

    1990-01-01

    In establishing of the nuclear fuel cycle in Japan that have a vulnerability in own energy structure, an effectiveness of energy security should be taken into account as well as an economy based on the balance of supply and demand of nuclear fuels. NMCC develops the 'Nuclear Energy Security Code' which was able to evaluate the effectiveness of energy security. Evaluation method adopted in this code is 'Import Premium' which was proposed in 'World Oil', EMF Report 6. The viewpoints of evaluation are as follows: 1. How much uranium fuel quantity can be reduced by using plutonium fuel? 2. How much a sudden rise of fuel cost can be absorbed by establishing the plutonium cycle beforehand the energy crisis? (author)

  19. Advanced burnup calculation code system in a subcritical state with continuous-energy Monte Carlo code for fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Matsunaka, Masayuki; Ohta, Masayuki; Miyamaru, Hiroyuki; Murata, Isao

    2009-01-01

    The fusion-fission (FF) hybrid reactor is a promising energy source that is thought to act as a bridge between the existing fission reactor and the genuine fusion reactor in the future. The burnup calculation system that aims at precise burnup calculations of a subcritical system was developed for the detailed design of the FF hybrid reactor, and the system consists of MCNP, ORIGEN, and postprocess codes. In the present study, the calculation system was substantially modified to improve the calculation accuracy and at the same time the calculation speed as well. The reaction rate estimation can be carried out accurately with the present system that uses track-length (TL) data in the continuous-energy treatment. As for the speed-up of the reaction rate calculation, a new TL data bunching scheme was developed so that only necessary TL data are used as long as the accuracy of the point-wise nuclear data is conserved. With the present system, an example analysis result for our proposed FF hybrid reactor is described, showing that the computation time could really be saved with the same accuracy as before. (author)

  20. Neutron transport-burnup code MCORGS and its application in fusion fission hybrid blanket conceptual research

    Science.gov (United States)

    Shi, Xue-Ming; Peng, Xian-Jue

    2016-09-01

    Fusion science and technology has made progress in the last decades. However, commercialization of fusion reactors still faces challenges relating to higher fusion energy gain, irradiation-resistant material, and tritium self-sufficiency. Fusion Fission Hybrid Reactors (FFHR) can be introduced to accelerate the early application of fusion energy. Traditionally, FFHRs have been classified as either breeders or transmuters. Both need partition of plutonium from spent fuel, which will pose nuclear proliferation risks. A conceptual design of a Fusion Fission Hybrid Reactor for Energy (FFHR-E), which can make full use of natural uranium with lower nuclear proliferation risk, is presented. The fusion core parameters are similar to those of the International Thermonuclear Experimental Reactor. An alloy of natural uranium and zirconium is adopted in the fission blanket, which is cooled by light water. In order to model blanket burnup problems, a linkage code MCORGS, which couples MCNP4B and ORIGEN-S, is developed and validated through several typical benchmarks. The average blanket energy Multiplication and Tritium Breeding Ratio can be maintained at 10 and 1.15 respectively over tens of years of continuous irradiation. If simple reprocessing without separation of plutonium from uranium is adopted every few years, FFHR-E can achieve better neutronic performance. MCORGS has also been used to analyze the ultra-deep burnup model of Laser Inertial Confinement Fusion Fission Energy (LIFE) from LLNL, and a new blanket design that uses Pb instead of Be as the neutron multiplier is proposed. In addition, MCORGS has been used to simulate the fluid transmuter model of the In-Zinerater from Sandia. A brief comparison of LIFE, In-Zinerater, and FFHR-E will be given.

  1. High performance 3D neutron transport on peta scale and hybrid architectures within APOLLO3 code

    International Nuclear Information System (INIS)

    Jamelot, E.; Dubois, J.; Lautard, J-J.; Calvin, C.; Baudron, A-M.

    2011-01-01

    APOLLO3 code is a common project of CEA, AREVA and EDF for the development of a new generation system for core physics analysis. We present here the parallelization of two deterministic transport solvers of APOLLO3: MINOS, a simplified 3D transport solver on structured Cartesian and hexagonal grids, and MINARET, a transport solver based on triangular meshes on 2D and prismatic ones in 3D. We used two different techniques to accelerate MINOS: a domain decomposition method, combined with an accelerated algorithm using GPU. The domain decomposition is based on the Schwarz iterative algorithm, with Robin boundary conditions to exchange information. The Robin parameters influence the convergence and we detail how we optimized the choice of these parameters. MINARET parallelization is based on angular directions calculation using explicit message passing. Fine grain parallelization is also available for each angular direction using shared memory multithreaded acceleration. Many performance results are presented on massively parallel architectures using more than 103 cores and on hybrid architectures using some tens of GPUs. This work contributes to the HPC development in reactor physics at the CEA Nuclear Energy Division. (author)

  2. Development of Probabilistic Internal Dosimetry Computer Code

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Siwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kwon, Tae-Eun [Korea Institute of Radiological and Medical Sciences, Seoul (Korea, Republic of); Lee, Jai-Ki [Korean Association for Radiation Protection, Seoul (Korea, Republic of)

    2017-02-15

    Internal radiation dose assessment involves biokinetic models, the corresponding parameters, measured data, and many assumptions. Every component considered in the internal dose assessment has its own uncertainty, which is propagated in the intake activity and internal dose estimates. For research or scientific purposes, and for retrospective dose reconstruction for accident scenarios occurring in workplaces having a large quantity of unsealed radionuclides, such as nuclear power plants, nuclear fuel cycle facilities, and facilities in which nuclear medicine is practiced, a quantitative uncertainty assessment of the internal dose is often required. However, no calculation tools or computer codes that incorporate all the relevant processes and their corresponding uncertainties, i.e., from the measured data to the committed dose, are available. Thus, the objective of the present study is to develop an integrated probabilistic internal-dose-assessment computer code. First, the uncertainty components in internal dosimetry are identified, and quantitative uncertainty data are collected. Then, an uncertainty database is established for each component. In order to propagate these uncertainties in an internal dose assessment, a probabilistic internal-dose-assessment system that employs the Bayesian and Monte Carlo methods. Based on the developed system, we developed a probabilistic internal-dose-assessment code by using MATLAB so as to estimate the dose distributions from the measured data with uncertainty. Using the developed code, we calculated the internal dose distribution and statistical values (e.g. the 2.5{sup th}, 5{sup th}, median, 95{sup th}, and 97.5{sup th} percentiles) for three sample scenarios. On the basis of the distributions, we performed a sensitivity analysis to determine the influence of each component on the resulting dose in order to identify the major component of the uncertainty in a bioassay. The results of this study can be applied to various

  3. Development of disruption thermal analysis code DREAM

    Energy Technology Data Exchange (ETDEWEB)

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

    1989-07-01

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

  4. Development of Probabilistic Internal Dosimetry Computer Code

    International Nuclear Information System (INIS)

    Noh, Siwan; Kwon, Tae-Eun; Lee, Jai-Ki

    2017-01-01

    Internal radiation dose assessment involves biokinetic models, the corresponding parameters, measured data, and many assumptions. Every component considered in the internal dose assessment has its own uncertainty, which is propagated in the intake activity and internal dose estimates. For research or scientific purposes, and for retrospective dose reconstruction for accident scenarios occurring in workplaces having a large quantity of unsealed radionuclides, such as nuclear power plants, nuclear fuel cycle facilities, and facilities in which nuclear medicine is practiced, a quantitative uncertainty assessment of the internal dose is often required. However, no calculation tools or computer codes that incorporate all the relevant processes and their corresponding uncertainties, i.e., from the measured data to the committed dose, are available. Thus, the objective of the present study is to develop an integrated probabilistic internal-dose-assessment computer code. First, the uncertainty components in internal dosimetry are identified, and quantitative uncertainty data are collected. Then, an uncertainty database is established for each component. In order to propagate these uncertainties in an internal dose assessment, a probabilistic internal-dose-assessment system that employs the Bayesian and Monte Carlo methods. Based on the developed system, we developed a probabilistic internal-dose-assessment code by using MATLAB so as to estimate the dose distributions from the measured data with uncertainty. Using the developed code, we calculated the internal dose distribution and statistical values (e.g. the 2.5 th , 5 th , median, 95 th , and 97.5 th percentiles) for three sample scenarios. On the basis of the distributions, we performed a sensitivity analysis to determine the influence of each component on the resulting dose in order to identify the major component of the uncertainty in a bioassay. The results of this study can be applied to various situations. In cases

  5. Development of disruption thermal analysis code DREAM

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  6. Code development for nuclear reactor simulation

    International Nuclear Information System (INIS)

    Chauliac, C.; Verwaerde, D.; Pavageau, O.

    2006-01-01

    Full text of publication follows: Since several years, CEA, EDF and FANP have developed several numerical codes which are currently used for nuclear industry applications and will be remain in use for the coming years. Complementary to this set of codes and in order to better meet the present and future needs, a new system is being developed through a joint venture between CEA, EDF and FANP, with a ten year prospect and strong intermediate milestones. The focus is put on a multi-scale and multi-physics approach enabling to take into account phenomena from microscopic to macroscopic scale, and to describe interactions between various physical fields such as neutronics (DESCARTES), thermal-hydraulics (NEPTUNE) and fuel behaviour (PLEIADES). This approach is based on a more rational design of the softwares and uses a common integration platform providing pre-processing, supervision of computation and post-processing. This paper will describe the overall system under development and present the first results obtained. (authors)

  7. Frontier battery development for hybrid vehicles.

    Science.gov (United States)

    Lewis, Heather; Park, Haram; Paolini, Maion

    2012-04-23

    Interest in hybrid-electric vehicles (HEVs) has recently spiked, partly due to an increasingly negative view toward the U.S. foreign oil dependency and environmental concerns. Though HEVs are becoming more common, they have a significant price premium over gasoline-powered vehicles. One of the primary drivers of this "hybrid premium" is the cost of the vehicles' batteries. This paper focuses on these batteries used in hybrid vehicles, examines the types of batteries used for transportation applications and addresses some of the technological, environmental and political drivers in battery development and the deployment of HEVs. This paper examines the claim, often voiced by HEV proponents, that by taking into account savings on gasoline and vehicle maintenance, hybrid cars are cheaper than traditional gasoline cars. This is done by a quantitative benefit-cost analysis, in addition to qualitative benefit-cost analysis from political, technological and environmental perspectives. The quantitative benefit-cost analysis shows that, taking account of all costs for the life of the vehicle, hybrid cars are in fact more expensive than gasoline-powered vehicles; however, after five years, HEVs will break even with gasoline cars. Our results show that it is likely that after 5 years, using hybrid vehicles should be cheaper in effect and yield a positive net benefit to society. There are a number of externalities that could significantly impact the total social cost of the car. These externalities can be divided into four categories: environmental, industrial, R&D and political. Despite short-term implications and hurdles, increased HEV usage forecasts a generally favorable long-term net benefit to society. Most notably, increasing HEV usage could decrease greenhouse gas emissions, while also decreasing U.S. dependence on foreign oil.

  8. Frontier battery development for hybrid vehicles

    Directory of Open Access Journals (Sweden)

    Lewis Heather

    2012-04-01

    Full Text Available Abstract Background Interest in hybrid-electric vehicles (HEVs has recently spiked, partly due to an increasingly negative view toward the U.S. foreign oil dependency and environmental concerns. Though HEVs are becoming more common, they have a significant price premium over gasoline-powered vehicles. One of the primary drivers of this “hybrid premium” is the cost of the vehicles’ batteries. This paper focuses on these batteries used in hybrid vehicles, examines the types of batteries used for transportation applications and addresses some of the technological, environmental and political drivers in battery development and the deployment of HEVs. Methods This paper examines the claim, often voiced by HEV proponents, that by taking into account savings on gasoline and vehicle maintenance, hybrid cars are cheaper than traditional gasoline cars. This is done by a quantitative benefit-cost analysis, in addition to qualitative benefit-cost analysis from political, technological and environmental perspectives. Results The quantitative benefit-cost analysis shows that, taking account of all costs for the life of the vehicle, hybrid cars are in fact more expensive than gasoline-powered vehicles; however, after five years, HEVs will break even with gasoline cars. Conclusions Our results show that it is likely that after 5 years, using hybrid vehicles should be cheaper in effect and yield a positive net benefit to society. There are a number of externalities that could significantly impact the total social cost of the car. These externalities can be divided into four categories: environmental, industrial, R&D and political. Despite short-term implications and hurdles, increased HEV usage forecasts a generally favorable long-term net benefit to society. Most notably, increasing HEV usage could decrease greenhouse gas emissions, while also decreasing U.S. dependence on foreign oil.

  9. Frontier battery development for hybrid vehicles

    Science.gov (United States)

    2012-01-01

    Background Interest in hybrid-electric vehicles (HEVs) has recently spiked, partly due to an increasingly negative view toward the U.S. foreign oil dependency and environmental concerns. Though HEVs are becoming more common, they have a significant price premium over gasoline-powered vehicles. One of the primary drivers of this “hybrid premium” is the cost of the vehicles’ batteries. This paper focuses on these batteries used in hybrid vehicles, examines the types of batteries used for transportation applications and addresses some of the technological, environmental and political drivers in battery development and the deployment of HEVs. Methods This paper examines the claim, often voiced by HEV proponents, that by taking into account savings on gasoline and vehicle maintenance, hybrid cars are cheaper than traditional gasoline cars. This is done by a quantitative benefit-cost analysis, in addition to qualitative benefit-cost analysis from political, technological and environmental perspectives. Results The quantitative benefit-cost analysis shows that, taking account of all costs for the life of the vehicle, hybrid cars are in fact more expensive than gasoline-powered vehicles; however, after five years, HEVs will break even with gasoline cars. Conclusions Our results show that it is likely that after 5 years, using hybrid vehicles should be cheaper in effect and yield a positive net benefit to society. There are a number of externalities that could significantly impact the total social cost of the car. These externalities can be divided into four categories: environmental, industrial, R&D and political. Despite short-term implications and hurdles, increased HEV usage forecasts a generally favorable long-term net benefit to society. Most notably, increasing HEV usage could decrease greenhouse gas emissions, while also decreasing U.S. dependence on foreign oil. PMID:22540987

  10. Development of Premacy Hydrogen RE Hybrid

    Energy Technology Data Exchange (ETDEWEB)

    Wakayama, N. [Mazda Motor Corporation, Hiroshima (Japan)

    2010-07-01

    Hydrogen powered ICE (internal combustion engine) vehicles can play an important role as an automotive power source in the future, because of its higher reliability and cost performance than those of fuel cell vehicles. Combined with hydrogen, Mazda's unique rotary engine (RE) has merits such as a prevention of hydrogen pre-ignition. Mazda has been developing hydrogen vehicles with the hydrogen RE from the early 1990s. Premacy (Mazda5) Hydrogen RE Hybrid was developed and launched in 2009, following RX-8 Hydrogen RE delivered in 2006. A series hybrid system was adopted in Premacy Hydrogen RE Hybrid. A traction motor switches its windings while the vehicle is moving. This switching technology allows the motor to be small and high-efficient. The lithium-ion high voltage battery, which has excellent input-output characteristics, was installed. These features extend the hydrogen fuel driving range to 200 km and obtain excellent acceleration performance. The hydrogen RE can be also operated by gasoline (Dual Fuel System). The additional gasoline operation makes hydrogen vehicles possible to drive in non-hydrogen station area. With approval from the Japanese Ministry of Land Infrastructure and Transport, Mazda Premacy Hydrogen RE Hybrid was delivered successfully to the Japanese market in the form of leasing. (orig.)

  11. Review of Hybrid (Deterministic/Monte Carlo) Radiation Transport Methods, Codes, and Applications at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Wagner, John C.; Peplow, Douglas E.; Mosher, Scott W.; Evans, Thomas M.

    2010-01-01

    This paper provides a review of the hybrid (Monte Carlo/deterministic) radiation transport methods and codes used at the Oak Ridge National Laboratory and examples of their application for increasing the efficiency of real-world, fixed-source Monte Carlo analyses. The two principal hybrid methods are (1) Consistent Adjoint Driven Importance Sampling (CADIS) for optimization of a localized detector (tally) region (e.g., flux, dose, or reaction rate at a particular location) and (2) Forward Weighted CADIS (FW-CADIS) for optimizing distributions (e.g., mesh tallies over all or part of the problem space) or multiple localized detector regions (e.g., simultaneous optimization of two or more localized tally regions). The two methods have been implemented and automated in both the MAVRIC sequence of SCALE 6 and ADVANTG, a code that works with the MCNP code. As implemented, the methods utilize the results of approximate, fast-running 3-D discrete ordinates transport calculations (with the Denovo code) to generate consistent space- and energy-dependent source and transport (weight windows) biasing parameters. These methods and codes have been applied to many relevant and challenging problems, including calculations of PWR ex-core thermal detector response, dose rates throughout an entire PWR facility, site boundary dose from arrays of commercial spent fuel storage casks, radiation fields for criticality accident alarm system placement, and detector response for special nuclear material detection scenarios and nuclear well-logging tools. Substantial computational speed-ups, generally O(10 2-4 ), have been realized for all applications to date. This paper provides a brief review of the methods, their implementation, results of their application, and current development activities, as well as a considerable list of references for readers seeking more information about the methods and/or their applications.

  12. Review of Hybrid (Deterministic/Monte Carlo) Radiation Transport Methods, Codes, and Applications at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Wagner, John C.; Peplow, Douglas E.; Mosher, Scott W.; Evans, Thomas M.

    2010-01-01

    This paper provides a review of the hybrid (Monte Carlo/deterministic) radiation transport methods and codes used at the Oak Ridge National Laboratory and examples of their application for increasing the efficiency of real-world, fixed-source Monte Carlo analyses. The two principal hybrid methods are (1) Consistent Adjoint Driven Importance Sampling (CADIS) for optimization of a localized detector (tally) region (e.g., flux, dose, or reaction rate at a particular location) and (2) Forward Weighted CADIS (FW-CADIS) for optimizing distributions (e.g., mesh tallies over all or part of the problem space) or multiple localized detector regions (e.g., simultaneous optimization of two or more localized tally regions). The two methods have been implemented and automated in both the MAVRIC sequence of SCALE 6 and ADVANTG, a code that works with the MCNP code. As implemented, the methods utilize the results of approximate, fast-running 3-D discrete ordinates transport calculations (with the Denovo code) to generate consistent space- and energy-dependent source and transport (weight windows) biasing parameters. These methods and codes have been applied to many relevant and challenging problems, including calculations of PWR ex-core thermal detector response, dose rates throughout an entire PWR facility, site boundary dose from arrays of commercial spent fuel storage casks, radiation fields for criticality accident alarm system placement, and detector response for special nuclear material detection scenarios and nuclear well-logging tools. Substantial computational speed-ups, generally O(102-4), have been realized for all applications to date. This paper provides a brief review of the methods, their implementation, results of their application, and current development activities, as well as a considerable list of references for readers seeking more information about the methods and/or their applications.

  13. Review of hybrid (deterministic/Monte Carlo) radiation transport methods, codes, and applications at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Wagner, J.C.; Peplow, D.E.; Mosher, S.W.; Evans, T.M.

    2010-01-01

    This paper provides a review of the hybrid (Monte Carlo/deterministic) radiation transport methods and codes used at the Oak Ridge National Laboratory and examples of their application for increasing the efficiency of real-world, fixed-source Monte Carlo analyses. The two principal hybrid methods are (1) Consistent Adjoint Driven Importance Sampling (CADIS) for optimization of a localized detector (tally) region (e.g., flux, dose, or reaction rate at a particular location) and (2) Forward Weighted CADIS (FW-CADIS) for optimizing distributions (e.g., mesh tallies over all or part of the problem space) or multiple localized detector regions (e.g., simultaneous optimization of two or more localized tally regions). The two methods have been implemented and automated in both the MAVRIC sequence of SCALE 6 and ADVANTG, a code that works with the MCNP code. As implemented, the methods utilize the results of approximate, fast-running 3-D discrete ordinates transport calculations (with the Denovo code) to generate consistent space- and energy-dependent source and transport (weight windows) biasing parameters. These methods and codes have been applied to many relevant and challenging problems, including calculations of PWR ex-core thermal detector response, dose rates throughout an entire PWR facility, site boundary dose from arrays of commercial spent fuel storage casks, radiation fields for criticality accident alarm system placement, and detector response for special nuclear material detection scenarios and nuclear well-logging tools. Substantial computational speed-ups, generally O(10 2-4 ), have been realized for all applications to date. This paper provides a brief review of the methods, their implementation, results of their application, and current development activities, as well as a considerable list of references for readers seeking more information about the methods and/or their applications. (author)

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

    Science.gov (United States)

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

    2016-10-01

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

  15. FDA Developments: Food Code 2013 and Proposed Trans Fat Determination

    NARCIS (Netherlands)

    Grossman, M.R.

    2014-01-01

    268 Reports EFFL 4|2014 USA FDA Developments: Food Code 2013 and Proposed Trans Fat Determination Margaret Rosso Grossman* I. Food Code 2013 and Food Code Reference System Since 1993, the US Food and Drug Administration has published a Food Code, now updated every four years. In November 2013, the

  16. TRAC code development status and plans

    International Nuclear Information System (INIS)

    Spore, J.W.; Liles, D.R.; Nelson, R.A.

    1986-01-01

    This report summarizes the characteristics and current status of the TRAC-PF1/MOD1 computer code. Recent error corrections and user-convenience features are described, and several user enhancements are identified. Current plans for the release of the TRAC-PF1/MOD2 computer code and some preliminary MOD2 results are presented. This new version of the TRAC code implements stability-enhancing two-step numerics into the 3-D vessel, using partial vectorization to obtain a code that has run 400% faster than the MOD1 code

  17. Theoretical Atomic Physics code development II: ACE: Another collisional excitation code

    International Nuclear Information System (INIS)

    Clark, R.E.H.; Abdallah, J. Jr.; Csanak, G.; Mann, J.B.; Cowan, R.D.

    1988-12-01

    A new computer code for calculating collisional excitation data (collision strengths or cross sections) using a variety of models is described. The code uses data generated by the Cowan Atomic Structure code or CATS for the atomic structure. Collisional data are placed on a random access file and can be displayed in a variety of formats using the Theoretical Atomic Physics Code or TAPS. All of these codes are part of the Theoretical Atomic Physics code development effort at Los Alamos. 15 refs., 10 figs., 1 tab

  18. Developing a Blended Type Course of Introduction to Hybrid Vehicles

    OpenAIRE

    Na Zhu

    2016-01-01

    An innovative course of introduction to hybrid vehicles is developed for both associate and bachelor degree programs for engineering technology with automotive/mechanical concentration. The hybrid vehicle course content includes several topics, such as the rational of pure electric vehicle and hybrid vehicle, hybrid vehicle propulsion systems, fundamentals of motor/generator systems, fundamentals of battery and energy management system, and introduction to various configurations of hybrid veh...

  19. The PARTRAC code: Status and recent developments

    Science.gov (United States)

    Friedland, Werner; Kundrat, Pavel

    Biophysical modeling is of particular value for predictions of radiation effects due to manned space missions. PARTRAC is an established tool for Monte Carlo-based simulations of radiation track structures, damage induction in cellular DNA and its repair [1]. Dedicated modules describe interactions of ionizing particles with the traversed medium, the production and reactions of reactive species, and score DNA damage determined by overlapping track structures with multi-scale chromatin models. The DNA repair module describes the repair of DNA double-strand breaks (DSB) via the non-homologous end-joining pathway; the code explicitly simulates the spatial mobility of individual DNA ends in parallel with their processing by major repair enzymes [2]. To simulate the yields and kinetics of radiation-induced chromosome aberrations, the repair module has been extended by tracking the information on the chromosome origin of ligated fragments as well as the presence of centromeres [3]. PARTRAC calculations have been benchmarked against experimental data on various biological endpoints induced by photon and ion irradiation. The calculated DNA fragment distributions after photon and ion irradiation reproduce corresponding experimental data and their dose- and LET-dependence. However, in particular for high-LET radiation many short DNA fragments are predicted below the detection limits of the measurements, so that the experiments significantly underestimate DSB yields by high-LET radiation [4]. The DNA repair module correctly describes the LET-dependent repair kinetics after (60) Co gamma-rays and different N-ion radiation qualities [2]. First calculations on the induction of chromosome aberrations have overestimated the absolute yields of dicentrics, but correctly reproduced their relative dose-dependence and the difference between gamma- and alpha particle irradiation [3]. Recent developments of the PARTRAC code include a model of hetero- vs euchromatin structures to enable

  20. Recent Developments in Abrasive Hybrid Manufacturing Processes

    Directory of Open Access Journals (Sweden)

    Ruszaj Adam

    2017-06-01

    Full Text Available Recent dynamic development of abrasive hybrid manufacturing processes results from application of a new difficult for machining materials and improvement of technological indicators of manufacturing processes already applied in practice. This tendency also occurs in abrasive machining processes which are often supported by ultrasonic vibrations, electrochemical dissolution or by electrical discharges. In the paper we present the review of new results of investigations and new practical applications of Abrasive Electrodischarge (AEDM and Electrochemical (AECM Machining.

  1. Human Resource Development in Hybrid Libraries

    OpenAIRE

    Prakasan, E. R.; Swarna, T.; Vijai Kumar, *

    2000-01-01

    This paper explores the human resources and development implications in hybrid libraries. Due to technological changes in libraries, which is a result of the proliferation of electronic resources, there has been a shift in workloads and workflow, requiring staff with different skills and educational backgrounds. Training of staff at all levels in information technology is the key to manage change, alleviate anxiety in the workplace and assure quality service in the libraries. Staff developmen...

  2. ER@CEBAF: Modeling code developments

    Energy Technology Data Exchange (ETDEWEB)

    Meot, F. [Brookhaven National Lab. (BNL), Upton, NY (United States); Thomas Jefferson National Accelerator Facility (TJNAF), Newport News, VA (United States); Roblin, Y. [Brookhaven National Lab. (BNL), Upton, NY (United States); Thomas Jefferson National Accelerator Facility (TJNAF), Newport News, VA (United States)

    2016-04-13

    A proposal for a multiple-pass, high-energy, energy-recovery experiment using CEBAF is under preparation in the frame of a JLab-BNL collaboration. In view of beam dynamics investigations regarding this project, in addition to the existing model in use in Elegant a version of CEBAF is developed in the stepwise ray-tracing code Zgoubi, Beyond the ER experiment, it is also planned to use the latter for the study of polarization transport in the presence of synchrotron radiation, down to Hall D line where a 12 GeV polarized beam can be delivered. This Note briefly reports on the preliminary steps, and preliminary outcomes, based on an Elegant to Zgoubi translation.

  3. The FLUKA code for space applications Recent developments

    CERN Document Server

    Andersen, V; Battistoni, G; Campanella, M; Carboni, M; Cerutti, F; Empl, A; Fassò, A; Ferrari, A; Gadioli, E; Garzelli, M V; Lee, K; Ottolenghi, A; Pelliccioni, M; Pinsky, L S; Ranft, J; Roesler, S; Sala, P R; Wilson, T L

    2004-01-01

    The FLUKA Monte Carlo transport code is widely used for fundamental research, radioprotection and dosimetry, hybrid nuclear energy system and cosmic ray calculations. The validity of its physical models has been benchmarked against a variety of experimental data over a wide range of energies, ranging from accelerator data to cosmic ray showers in the earth atmosphere. The code is presently undergoing several developments in order to better fit the needs of space applications. The generation of particle spectra according to up-to- date cosmic ray data as well as the effect of the solar and geomagnetic modulation have been implemented and already successfully applied to a variety of problems. The implementation of suitable models for heavy ion nuclear interactions has reached an operational stage. At medium/high energy FLUKA is using the DPMJET model. The major task of incorporating heavy ion interactions from a few GeV/n down to the threshold for inelastic collisions is also progressing and promising results h...

  4. Developing multiplatform apps : using hybrid solutions

    OpenAIRE

    Kristbergsson, Ellert

    2015-01-01

    The aim of this thesis is to describe the process of developing multiplatform apps for mobile phones by discussing and comparing the different methods of developing apps for all platforms. The purpose is to ease decision making of smaller companies that are starting to develop their own product for smartphones. The paper analyses tools based on the Cordova framework and the Qt framework. The difference between how the frameworks deliver the code to the machine, Cordova compiles JavaScript...

  5. Hybrid digital-analog coding with bandwidth expansion for correlated Gaussian sources under Rayleigh fading

    Science.gov (United States)

    Yahampath, Pradeepa

    2017-12-01

    Consider communicating a correlated Gaussian source over a Rayleigh fading channel with no knowledge of the channel signal-to-noise ratio (CSNR) at the transmitter. In this case, a digital system cannot be optimal for a range of CSNRs. Analog transmission however is optimal at all CSNRs, if the source and channel are memoryless and bandwidth matched. This paper presents new hybrid digital-analog (HDA) systems for sources with memory and channels with bandwidth expansion, which outperform both digital-only and analog-only systems over a wide range of CSNRs. The digital part is either a predictive quantizer or a transform code, used to achieve a coding gain. Analog part uses linear encoding to transmit the quantization error which improves the performance under CSNR variations. The hybrid encoder is optimized to achieve the minimum AMMSE (average minimum mean square error) over the CSNR distribution. To this end, analytical expressions are derived for the AMMSE of asymptotically optimal systems. It is shown that the outage CSNR of the channel code and the analog-digital power allocation must be jointly optimized to achieve the minimum AMMSE. In the case of HDA predictive quantization, a simple algorithm is presented to solve the optimization problem. Experimental results are presented for both Gauss-Markov sources and speech signals.

  6. 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)

  7. Software development infrastructure for the HYBRID modeling and simulation project

    International Nuclear Information System (INIS)

    Epiney, Aaron S.; Kinoshita, Robert A.; Kim, Jong Suk; Rabiti, Cristian; Greenwood, M. Scott

    2016-01-01

    One of the goals of the HYBRID modeling and simulation project is to assess the economic viability of hybrid systems in a market that contains renewable energy sources like wind. The idea is that it is possible for the nuclear plant to sell non-electric energy cushions, which absorb (at least partially) the volatility introduced by the renewable energy sources. This system is currently modeled in the Modelica programming language. To assess the economics of the system, an optimization procedure is trying to find the minimal cost of electricity production. The RAVEN code is used as a driver for the whole problem. It is assumed that at this stage, the HYBRID modeling and simulation framework can be classified as non-safety “research and development” software. The associated quality level is Quality Level 3 software. This imposes low requirements on quality control, testing and documentation. The quality level could change as the application development continues.Despite the low quality requirement level, a workflow for the HYBRID developers has been defined that include a coding standard and some documentation and testing requirements. The repository performs automated unit testing of contributed models. The automated testing is achieved via an open-source python script called BuildingsP from Lawrence Berkeley National Lab. BuildingsPy runs Modelica simulation tests using Dymola in an automated manner and generates and runs unit tests from Modelica scripts written by developers. In order to assure effective communication between the different national laboratories a biweekly videoconference has been set-up, where developers can report their progress and issues. In addition, periodic face-face meetings are organized intended to discuss high-level strategy decisions with management. A second means of communication is the developer email list. This is a list to which everybody can send emails that will be received by the collective of the developers and managers

  8. Software development infrastructure for the HYBRID modeling and simulation project

    Energy Technology Data Exchange (ETDEWEB)

    Epiney, Aaron S. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinoshita, Robert A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kim, Jong Suk [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Greenwood, M. Scott [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    One of the goals of the HYBRID modeling and simulation project is to assess the economic viability of hybrid systems in a market that contains renewable energy sources like wind. The idea is that it is possible for the nuclear plant to sell non-electric energy cushions, which absorb (at least partially) the volatility introduced by the renewable energy sources. This system is currently modeled in the Modelica programming language. To assess the economics of the system, an optimization procedure is trying to find the minimal cost of electricity production. The RAVEN code is used as a driver for the whole problem. It is assumed that at this stage, the HYBRID modeling and simulation framework can be classified as non-safety “research and development” software. The associated quality level is Quality Level 3 software. This imposes low requirements on quality control, testing and documentation. The quality level could change as the application development continues.Despite the low quality requirement level, a workflow for the HYBRID developers has been defined that include a coding standard and some documentation and testing requirements. The repository performs automated unit testing of contributed models. The automated testing is achieved via an open-source python script called BuildingsP from Lawrence Berkeley National Lab. BuildingsPy runs Modelica simulation tests using Dymola in an automated manner and generates and runs unit tests from Modelica scripts written by developers. In order to assure effective communication between the different national laboratories a biweekly videoconference has been set-up, where developers can report their progress and issues. In addition, periodic face-face meetings are organized intended to discuss high-level strategy decisions with management. A second means of communication is the developer email list. This is a list to which everybody can send emails that will be received by the collective of the developers and managers

  9. Hydrogen hybrid vehicle engine development: Experimental program

    Energy Technology Data Exchange (ETDEWEB)

    Van Blarigan, P. [Sandia National Lab., Livermore, CA (United States)

    1995-09-01

    A hydrogen fueled engine is being developed specifically for the auxiliary power unit (APU) in a series type hybrid vehicle. Hydrogen is different from other internal combustion (IC) engine fuels, and hybrid vehicle IC engine requirements are different from those of other IC vehicle engines. Together these differences will allow a new engine design based on first principles that will maximize thermal efficiency while minimizing principal emissions. The experimental program is proceeding in four steps: (1) Demonstration of the emissions and the indicated thermal efficiency capability of a standard CLR research engine modified for higher compression ratios and hydrogen fueled operation. (2) Design and test a new combustion chamber geometry for an existing single cylinder research engine, in an attempt to improve on the baseline indicated thermal efficiency of the CLR engine. (3) Design and build, in conjunction with an industrial collaborator, a new full scale research engine designed to maximize brake thermal efficiency. Include a full complement of combustion diagnostics. (4) Incorporate all of the knowledge thus obtained in the design and fabrication, by an industrial collaborator, of the hydrogen fueled engine for the hybrid vehicle power train illustrator. Results of the CLR baseline engine testing are presented, as well as preliminary data from the new combustion chamber engine. The CLR data confirm the low NOx produced by lean operation. The preliminary indicated thermal efficiency data from the new combustion chamber design engine show an improvement relative to the CLR engine. Comparison with previous high compression engine results shows reasonable agreement.

  10. An integrated PCR colony hybridization approach to screen cDNA libraries for full-length coding sequences.

    Science.gov (United States)

    Pollier, Jacob; González-Guzmán, Miguel; Ardiles-Diaz, Wilson; Geelen, Danny; Goossens, Alain

    2011-01-01

    cDNA-Amplified Fragment Length Polymorphism (cDNA-AFLP) is a commonly used technique for genome-wide expression analysis that does not require prior sequence knowledge. Typically, quantitative expression data and sequence information are obtained for a large number of differentially expressed gene tags. However, most of the gene tags do not correspond to full-length (FL) coding sequences, which is a prerequisite for subsequent functional analysis. A medium-throughput screening strategy, based on integration of polymerase chain reaction (PCR) and colony hybridization, was developed that allows in parallel screening of a cDNA library for FL clones corresponding to incomplete cDNAs. The method was applied to screen for the FL open reading frames of a selection of 163 cDNA-AFLP tags from three different medicinal plants, leading to the identification of 109 (67%) FL clones. Furthermore, the protocol allows for the use of multiple probes in a single hybridization event, thus significantly increasing the throughput when screening for rare transcripts. The presented strategy offers an efficient method for the conversion of incomplete expressed sequence tags (ESTs), such as cDNA-AFLP tags, to FL-coding sequences.

  11. Benchmark of coupling codes (ALOHA, TOPLHA and GRILL3D) with ITER-relevant Lower Hybrid antenna

    International Nuclear Information System (INIS)

    Milanesio, D.; Hillairet, J.; Panaccione, L.; Maggiora, R.; Artaud, J.F.; Bae, Y.S.; Barbera, A.M.A.; Belo, J.; Berger-By, G.; Bernard, J.M.; Cara, Ph.; Cardinali, A.; Castaldo, C.; Ceccuzzi, S.; Cesario, R.; Decker, J.; Delpech, L.; Ekedahl, A.; Garcia, J.; Garibaldi, P.

    2011-01-01

    In order to assist the design of the future ITER Lower Hybrid launcher, coupling codes ALOHA, from CEA/IRFM, TOPLHA, from Politecnico di Torino, and GRILL3D, developed by Dr. Mikhail Irzak (A.F. Ioffe Physico-Technical Institute, St. Petersburg, Russia) and operated by ENEA Frascati, have been compared with the updated (six modules with four active waveguides per module) Passive-Active Multi-junction (PAM) Lower Hybrid antennas. Both ALOHA and GRILL3D formulate the problem in terms of rectangular waveguides modes, while TOPLHA is based on boundary-value problem with the adoption of a triangular cell-mesh to represent the relevant waveguides surfaces. Several plasma profiles, with varying edge density and density increase, have been adopted to provide a complete description of the simulated launcher in terms of reflection coefficient, computed at the beginning of each LH module, and of power spectra. Good agreement can be observed among codes for all the simulated profiles.

  12. Development of a definitive internal dosimetry code

    International Nuclear Information System (INIS)

    Miller, G.; Inkret, W.C.; Schillaci, M.E.

    1996-01-01

    Internal dosimetry may be divided into tow main problems: (1) the forward (scientific) problem of determining biokinetics models that describe how radionuclides are taken into the body, distributed in body tissues, and excreted, and (2) the inverse (mathematical) problem: given the measured amounts in excreta and assuming a biokinetic model, to determine the times and amounts of intakes into the body. The inverse problem of internal dosimetry is, in fact, a generic problem studied in other fields (e.g., image reconstruction, spectral deconvulution, and model parameter fitting). We have developed a code for plutonium internal dosimetry using the maximum entropy method, a method for solving underdetermined inverse problems with a positivity constraint. Within the framework of Bayesian statistics, we believe the definitive approach is to examine the Bayesian posterior probability describing the probability of an intake scenario (X i ) read ( ... ) as open-quotes the set of,close quotes where X i denotes the intake amount that occurs on the with day. For plutonium, for a worker with a long employment history, this is a very high dimensional probability space, since there may be on the order of 10,000 days when intakes may have occurred. Within this high dimensional space, we calculate the mean intake scenario as i > where denotes the expectation value over the posterior probability distribution. Similarly, we calculate uncertainties and other relevant quantities, such as X 2 , as expectation values over the posterior distribution. Thanks to a recent breakthrough in describing the mathematical structure of the intake process (a Poisson sum representation of intakes), we have developed the initial version of a Bayesian expectation-value algorithm for internal dosimetry reconstructions

  13. Hybrid threshold adaptable quantum secret sharing scheme with reverse Huffman-Fibonacci-tree coding.

    Science.gov (United States)

    Lai, Hong; Zhang, Jun; Luo, Ming-Xing; Pan, Lei; Pieprzyk, Josef; Xiao, Fuyuan; Orgun, Mehmet A

    2016-08-12

    With prevalent attacks in communication, sharing a secret between communicating parties is an ongoing challenge. Moreover, it is important to integrate quantum solutions with classical secret sharing schemes with low computational cost for the real world use. This paper proposes a novel hybrid threshold adaptable quantum secret sharing scheme, using an m-bonacci orbital angular momentum (OAM) pump, Lagrange interpolation polynomials, and reverse Huffman-Fibonacci-tree coding. To be exact, we employ entangled states prepared by m-bonacci sequences to detect eavesdropping. Meanwhile, we encode m-bonacci sequences in Lagrange interpolation polynomials to generate the shares of a secret with reverse Huffman-Fibonacci-tree coding. The advantages of the proposed scheme is that it can detect eavesdropping without joint quantum operations, and permits secret sharing for an arbitrary but no less than threshold-value number of classical participants with much lower bandwidth. Also, in comparison with existing quantum secret sharing schemes, it still works when there are dynamic changes, such as the unavailability of some quantum channel, the arrival of new participants and the departure of participants. Finally, we provide security analysis of the new hybrid quantum secret sharing scheme and discuss its useful features for modern applications.

  14. Development of codes for physical calculations of WWER

    International Nuclear Information System (INIS)

    Novikov, A.N.

    2000-01-01

    A package of codes for physical calculations of WWER reactors, used at the RRC 'Kurchatov Institute' is discussed including the purpose of these codes, approximations used, degree of data verification, possibilities of automation of calculations and presentation of results, trends of further development of the codes. (Authors)

  15. Color-coded Live Imaging of Heterokaryon Formation and Nuclear Fusion of Hybridizing Cancer Cells.

    Science.gov (United States)

    Suetsugu, Atsushi; Matsumoto, Takuro; Hasegawa, Kosuke; Nakamura, Miki; Kunisada, Takahiro; Shimizu, Masahito; Saji, Shigetoyo; Moriwaki, Hisataka; Bouvet, Michael; Hoffman, Robert M

    2016-08-01

    Fusion of cancer cells has been studied for over half a century. However, the steps involved after initial fusion between cells, such as heterokaryon formation and nuclear fusion, have been difficult to observe in real time. In order to be able to visualize these steps, we have established cancer-cell sublines from the human HT-1080 fibrosarcoma, one expressing green fluorescent protein (GFP) linked to histone H2B in the nucleus and a red fluorescent protein (RFP) in the cytoplasm and the other subline expressing RFP in the nucleus (mCherry) linked to histone H2B and GFP in the cytoplasm. The two reciprocal color-coded sublines of HT-1080 cells were fused using the Sendai virus. The fused cells were cultured on plastic and observed using an Olympus FV1000 confocal microscope. Multi-nucleate (heterokaryotic) cancer cells, in addition to hybrid cancer cells with single-or multiple-fused nuclei, including fused mitotic nuclei, were observed among the fused cells. Heterokaryons with red, green, orange and yellow nuclei were observed by confocal imaging, even in single hybrid cells. The orange and yellow nuclei indicate nuclear fusion. Red and green nuclei remained unfused. Cell fusion with heterokaryon formation and subsequent nuclear fusion resulting in hybridization may be an important natural phenomenon between cancer cells that may make them more malignant. The ability to image the complex processes following cell fusion using reciprocal color-coded cancer cells will allow greater understanding of the genetic basis of malignancy. Copyright© 2016 International Institute of Anticancer Research (Dr. John G. Delinassios), All rights reserved.

  16. Hybrid pump for Pazflor deepwater development

    Energy Technology Data Exchange (ETDEWEB)

    Bibet, Pierre-Jean

    2010-07-01

    Subsea Processing gives engineers permanent technical challenges due to the comprehensive specifications which often stretch the proven limits of technology. It is particularly true for pumps that have to boost a multiphase effluent, because they must, often at the same time, be tolerant to free gas, able to handle very viscous, generate a high ?P, and be efficient. The Pazflor project decided to base the development on full utilization of subsea pumps. This paper presents the state of the art of the pumps that are designed to be installed on the sea floor. It further describes why existing technology could not match the technical requirements for the Pazflor project. It finally presents the pump specifically developed for this world first full field's development application, the so called 'Hybrid Pump', with the description of the associated qualification program and results. (Author)

  17. 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

  18. Development of time dependent safety analysis code for plasma anomaly events in fusion reactors

    International Nuclear Information System (INIS)

    Honda, Takuro; Okazaki, Takashi; Bartels, H.W.; Uckan, N.A.; Seki, Yasushi.

    1997-01-01

    A safety analysis code SAFALY has been developed to analyze plasma anomaly events in fusion reactors, e.g., a loss of plasma control. The code is a hybrid code comprising a zero-dimensional plasma dynamics and a one-dimensional thermal analysis of in-vessel components. The code evaluates the time evolution of plasma parameters and temperature distributions of in-vessel components. As the plasma-safety interface model, we proposed a robust plasma physics model taking into account updated data for safety assessment. For example, physics safety guidelines for beta limit, density limit and H-L mode confinement transition threshold power, etc. are provided in the model. The model of the in-vessel components are divided into twenty temperature regions in the poloidal direction taking account of radiative heat transfer between each surface of each region. This code can also describe the coolant behavior under hydraulic accidents with the results by hydraulics code and treat vaporization (sublimation) from plasma facing components (PFCs). Furthermore, the code includes the model of impurity transport form PFCs by using a transport probability and a time delay. Quantitative analysis based on the model is possible for a scenario of plasma passive shutdown. We examined the possibility of the code as a safety analysis code for plasma anomaly events in fusion reactors and had a prospect that it would contribute to the safety analysis of the International Thermonuclear Experimental Reactor (ITER). (author)

  19. Code Development and Analysis Program: developmental checkout of the BEACON/MOD2A code

    International Nuclear Information System (INIS)

    Ramsthaler, J.A.; Lime, J.F.; Sahota, M.S.

    1978-12-01

    A best-estimate transient containment code, BEACON, is being developed by EG and G Idaho, Inc. for the Nuclear Regulatory Commission's reactor safety research program. This is an advanced, two-dimensional fluid flow code designed to predict temperatures and pressures in a dry PWR containment during a hypothetical loss-of-coolant accident. The most recent version of the code, MOD2A, is presently in the final stages of production prior to being released to the National Energy Software Center. As part of the final code checkout, seven sample problems were selected to be run with BEACON/MOD2A

  20. A NEW HYBRID N-BODY-COAGULATION CODE FOR THE FORMATION OF GAS GIANT PLANETS

    International Nuclear Information System (INIS)

    Bromley, Benjamin C.; Kenyon, Scott J.

    2011-01-01

    We describe an updated version of our hybrid N-body-coagulation code for planet formation. In addition to the features of our 2006-2008 code, our treatment now includes algorithms for the one-dimensional evolution of the viscous disk, the accretion of small particles in planetary atmospheres, gas accretion onto massive cores, and the response of N-bodies to the gravitational potential of the gaseous disk and the swarm of planetesimals. To validate the N-body portion of the algorithm, we use a battery of tests in planetary dynamics. As a first application of the complete code, we consider the evolution of Pluto-mass planetesimals in a swarm of 0.1-1 cm pebbles. In a typical evolution time of 1-3 Myr, our calculations transform 0.01-0.1 M sun disks of gas and dust into planetary systems containing super-Earths, Saturns, and Jupiters. Low-mass planets form more often than massive planets; disks with smaller α form more massive planets than disks with larger α. For Jupiter-mass planets, masses of solid cores are 10-100 M + .

  1. Development of authentication code for multi-access optical code division multiplexing based quantum key distribution

    Science.gov (United States)

    Taiwo, Ambali; Alnassar, Ghusoon; Bakar, M. H. Abu; Khir, M. F. Abdul; Mahdi, Mohd Adzir; Mokhtar, M.

    2018-05-01

    One-weight authentication code for multi-user quantum key distribution (QKD) is proposed. The code is developed for Optical Code Division Multiplexing (OCDMA) based QKD network. A unique address assigned to individual user, coupled with degrading probability of predicting the source of the qubit transmitted in the channel offer excellent secure mechanism against any form of channel attack on OCDMA based QKD network. Flexibility in design as well as ease of modifying the number of users are equally exceptional quality presented by the code in contrast to Optical Orthogonal Code (OOC) earlier implemented for the same purpose. The code was successfully applied to eight simultaneous users at effective key rate of 32 bps over 27 km transmission distance.

  2. Theoretical atomic physics code development III TAPS: A display code for atomic physics data

    International Nuclear Information System (INIS)

    Clark, R.E.H.; Abdallah, J. Jr.; Kramer, S.P.

    1988-12-01

    A large amount of theoretical atomic physics data is becoming available through use of the computer codes CATS and ACE developed at Los Alamos National Laboratory. A new code, TAPS, has been written to access this data, perform averages over terms and configurations, and display information in graphical or text form. 7 refs., 13 figs., 1 tab

  3. Hybrid MPI-OpenMP Parallelism in the ONETEP Linear-Scaling Electronic Structure Code: Application to the Delamination of Cellulose Nanofibrils.

    Science.gov (United States)

    Wilkinson, Karl A; Hine, Nicholas D M; Skylaris, Chris-Kriton

    2014-11-11

    We present a hybrid MPI-OpenMP implementation of Linear-Scaling Density Functional Theory within the ONETEP code. We illustrate its performance on a range of high performance computing (HPC) platforms comprising shared-memory nodes with fast interconnect. Our work has focused on applying OpenMP parallelism to the routines which dominate the computational load, attempting where possible to parallelize different loops from those already parallelized within MPI. This includes 3D FFT box operations, sparse matrix algebra operations, calculation of integrals, and Ewald summation. While the underlying numerical methods are unchanged, these developments represent significant changes to the algorithms used within ONETEP to distribute the workload across CPU cores. The new hybrid code exhibits much-improved strong scaling relative to the MPI-only code and permits calculations with a much higher ratio of cores to atoms. These developments result in a significantly shorter time to solution than was possible using MPI alone and facilitate the application of the ONETEP code to systems larger than previously feasible. We illustrate this with benchmark calculations from an amyloid fibril trimer containing 41,907 atoms. We use the code to study the mechanism of delamination of cellulose nanofibrils when undergoing sonification, a process which is controlled by a large number of interactions that collectively determine the structural properties of the fibrils. Many energy evaluations were needed for these simulations, and as these systems comprise up to 21,276 atoms this would not have been feasible without the developments described here.

  4. MIDAS/PK code development using point kinetics model

    International Nuclear Information System (INIS)

    Song, Y. M.; Park, S. H.

    1999-01-01

    In this study, a MIDAS/PK code has been developed for analyzing the ATWS (Anticipated Transients Without Scram) which can be one of severe accident initiating events. The MIDAS is an integrated computer code based on the MELCOR code to develop a severe accident risk reduction strategy by Korea Atomic Energy Research Institute. In the mean time, the Chexal-Layman correlation in the current MELCOR, which was developed under a BWR condition, is appeared to be inappropriate for a PWR. So as to provide ATWS analysis capability to the MIDAS code, a point kinetics module, PKINETIC, has first been developed as a stand-alone code whose reference model was selected from the current accident analysis codes. In the next step, the MIDAS/PK code has been developed via coupling PKINETIC with the MIDAS code by inter-connecting several thermal hydraulic parameters between the two codes. Since the major concern in the ATWS analysis is the primary peak pressure during the early few minutes into the accident, the peak pressure from the PKINETIC module and the MIDAS/PK are compared with the RETRAN calculations showing a good agreement between them. The MIDAS/PK code is considered to be valuable for analyzing the plant response during ATWS deterministically, especially for the early domestic Westinghouse plants which rely on the operator procedure instead of an AMSAC (ATWS Mitigating System Actuation Circuitry) against ATWS. This capability of ATWS analysis is also important from the view point of accident management and mitigation

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

    International Nuclear Information System (INIS)

    Forster, R.A.; Parsons, K.

    1997-01-01

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

  6. Develop of a quantum electromechanical hybrid system

    Science.gov (United States)

    Hao, Yu; Rouxinol, Francisco; Brito, Frederico; Caldeira, Amir; Irish, Elinor; Lahaye, Matthew

    In this poster, we will show our results from measurements of a hybrid quantum system composed of a superconducting transmon qubit-coupled and ultra-high frequency nano-mechanical resonator, embedded in a superconducting cavity. The transmon is capacitively coupled to a 3.4GHz nanoresonator and a T-filter-biased high-Q transmission line cavity. Single-tone and two-tone transmission spectroscopy measurements are used to probe the interactions between the cavity, qubit and mechanical resonator. These measurements are in good agreement with numerical simulations based upon a master equation for the tripartite system including dissipation. The results indicate that this system may be developed to serve as a platform for more advanced measurements with nanoresonators, including quantum state measurement, the exploration of nanoresonator quantum noise, and reservoir engineering.

  7. Hardware in the loop platform development for hybrid vehicles

    Energy Technology Data Exchange (ETDEWEB)

    Wilhelm, E. [ETH Zurich, Zurich (Switzerland); Fowler, E.; Stevens, M.B. [Waterloo Univ., ON (Canada). Dept. of Chemical Engineering; Fraser, M.W. [Waterloo Univ., ON (Canada). Dept. of Mechanical Engineering

    2007-07-01

    This paper described a hardware-in-the-loop (HIL) validation simulation system designed to evaluate hybrid control strategies. The system was designed to reduce development costs and improve the safety of hybrid vehicle control systems. Model-based design processes for power trains typically include a series of processes to assess the real time and physical limitations of control systems prior to in-vehicle testing. The study used a 70 kW nickel metal hydride battery; a 67 kW 3-phase induction traction motor; and, a high voltage DC-DC converter within a fuel cell Chevrolet Equinox. Two physical vehicle controllers were used to interface with the virtual vehicle simulation in real time. System performance was monitored with a supervisory computer. A software in the loop (SIL) process was conducted to assess torque control and regenerative braking algorithm validation. An analysis of the controller code showed that a Simulink-native integrator block was updating too slowly. A custom integration term calculation was written. The charge control was then validated and tuned. It was concluded that use of the HIL system mitigated the risk of component damage through the identification and correction of unstable control logic. 10 refs., 2 tabs., 10 figs.

  8. Frontier battery development for hybrid vehicles

    OpenAIRE

    Lewis, Heather; Park, Haram; Paolini, Maion

    2012-01-01

    Abstract Background Interest in hybrid-electric vehicles (HEVs) has recently spiked, partly due to an increasingly negative view toward the U.S. foreign oil dependency and environmental concerns. Though HEVs are becoming more common, they have a significant price premium over gasoline-powered vehicles. One of the primary drivers of this “hybrid premium” is the cost of the vehicles’ batteries. This paper focuses on these batteries used in hybrid vehicles, examines the types of batteries used f...

  9. Development and Application of a Code for Internal Exposure (CINEX) based on the CINDY code

    International Nuclear Information System (INIS)

    Kravchik, T.; Duchan, N.; Sarah, R.; Gabay, Y.; Kol, R.

    2004-01-01

    Internal exposure to radioactive materials at the NRCN is evaluated using the CINDY (Code for Internal Dosimetry) Package. The code was developed by the Pacific Northwest Laboratory to assist the interpretation of bioassay data, provide bioassay projections and evaluate committed and calendar-year doses from intake or bioassay measurement data. It provides capabilities to calculate organ dose and effective dose equivalents using the International Commission on Radiological Protection (ICRP) 30 approach. The CINDY code operates under DOS operating system and consequently its operation needs a relatively long procedure which also includes a lot of manual typing that can lead to personal human mistakes. A new code has been developed at the NRCN, the CINEX (Code for Internal Exposure), which is an Excel application and leads to a significant reduction in calculation time (in the order of 5-10 times) and in the risk of personal human mistakes. The code uses a database containing tables which were constructed by the CINDY and contain the bioassay values predicted by the ICRP30 model after an intake of an activity unit of each isotope. Using the database, the code than calculates the appropriate intake and consequently the committed effective dose and organ dose. Calculations with the CINEX code were compared to similar calculations with the CINDY code. The discrepancies were less than 5%, which is the rounding error of the CINDY code. Attached is a table which compares parameters calculated with the CINEX and the CINDY codes (for a class Y uranium). The CINEX is now used at the NRCN to calculate occupational intakes and doses to workers with radioactive materials

  10. Computer-assisted Particle-in-Cell code development

    International Nuclear Information System (INIS)

    Kawata, S.; Boonmee, C.; Teramoto, T.; Drska, L.; Limpouch, J.; Liska, R.; Sinor, M.

    1997-12-01

    This report presents a new approach for an electromagnetic Particle-in-Cell (PIC) code development by a computer: in general PIC codes have a common structure, and consist of a particle pusher, a field solver, charge and current density collections, and a field interpolation. Because of the common feature, the main part of the PIC code can be mechanically developed on a computer. In this report we use the packages FIDE and GENTRAN of the REDUCE computer algebra system for discretizations of field equations and a particle equation, and for an automatic generation of Fortran codes. The approach proposed is successfully applied to the development of 1.5-dimensional PIC code. By using the generated PIC code the Weibel instability in a plasma is simulated. The obtained growth rate agrees well with the theoretical value. (author)

  11. Development of Traction Drive Motors for the Toyota Hybrid System

    Science.gov (United States)

    Kamiya, Munehiro

    Toyota Motor Corporation developed in 2005 a new hybrid system for a large SUV. This system included the new development of a high-speed traction drive motor achieving a significant increase in power weight ratio. This paper provides an overview of the hybrid system, discusses the characteristics required of a traction drive motor, and presents the technologies employed in the developed motor.

  12. Integrated code development for studying laser driven plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Takabe, Hideaki; Nagatomo, Hideo; Sunahara, Atsusi; Ohnishi, Naofumi; Naruo, Syuji; Mima, Kunioki [Osaka Univ., Suita (Japan). Inst. of Laser Engineering

    1998-03-01

    Present status and plan for developing an integrated implosion code are briefly explained by focusing on motivation, numerical scheme and issues to be developed more. Highly nonlinear stage of Rayleigh-Taylor instability of ablation front by laser irradiation has been simulated so as to be compared with model experiments. Improvement in transport and rezoning/remapping algorithms in ILESTA code is described. (author)

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

    International Nuclear Information System (INIS)

    Yang, Yanhua; Zhang, Hao

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-03-15

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

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

    International Nuclear Information System (INIS)

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

    2007-03-01

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

  17. GEOS Code Development Road Map - May, 2013

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, Scott [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Settgast, Randolph [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Fu, Pengcheng [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Antoun, Tarabay [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ryerson, F. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2013-05-03

    GEOS is a massively parallel computational framework designed to enable HPC-based simulations of subsurface reservoir stimulation activities with the goal of optimizing current operations and evaluating innovative stimulation methods. GEOS will enable coupling of different solvers associated with the various physical processes occurring during reservoir stimulation in unique and sophisticated ways, adapted to various geologic settings, materials and stimulation methods. The overall architecture of the framework includes consistent data structures and will allow incorporation of additional physical and materials models as demanded by future applications. Along with predicting the initiation, propagation and reactivation of fractures, GEOS will also generate a seismic source term that can be linked with seismic wave propagation codes to generate synthetic microseismicity at surface and downhole arrays. Similarly, the output from GEOS can be linked with existing fluid/thermal transport codes. GEOS can also be linked with existing, non-intrusive uncertainty quantification schemes to constrain uncertainty in its predictions and sensitivity to the various parameters describing the reservoir and stimulation operations. We anticipate that an implicit-explicit 3D version of GEOS, including a preliminary seismic source model, will be available for parametric testing and validation against experimental and field data by Oct. 1, 2013.

  18. Development of covariance capabilities in EMPIRE code

    Energy Technology Data Exchange (ETDEWEB)

    Herman,M.; Pigni, M.T.; Oblozinsky, P.; Mughabghab, S.F.; Mattoon, C.M.; Capote, R.; Cho, Young-Sik; Trkov, A.

    2008-06-24

    The nuclear reaction code EMPIRE has been extended to provide evaluation capabilities for neutron cross section covariances in the thermal, resolved resonance, unresolved resonance and fast neutron regions. The Atlas of Neutron Resonances by Mughabghab is used as a primary source of information on uncertainties at low energies. Care is taken to ensure consistency among the resonance parameter uncertainties and those for thermal cross sections. The resulting resonance parameter covariances are formatted in the ENDF-6 File 32. In the fast neutron range our methodology is based on model calculations with the code EMPIRE combined with experimental data through several available approaches. The model-based covariances can be obtained using deterministic (Kalman) or stochastic (Monte Carlo) propagation of model parameter uncertainties. We show that these two procedures yield comparable results. The Kalman filter and/or the generalized least square fitting procedures are employed to incorporate experimental information. We compare the two approaches analyzing results for the major reaction channels on {sup 89}Y. We also discuss a long-standing issue of unreasonably low uncertainties and link it to the rigidity of the model.

  19. Development of AGNES, a kinetics code for fissile solutions, 1

    International Nuclear Information System (INIS)

    Nakajima, Ken; Ohnishi, Nobuaki

    1986-01-01

    A kinetics code for fissile solutions, AGNES (Accidentally Generated Nuclear Excursion Simulation code), has been developed. This code calculates the radiolytic gas void effect as a reactivity feedback. Physical and calculative models of the radiolytic gas void are summarized and the usage of AGNES is described. In addition, some benchmark calculations were performed and results of calculations show good agreement with those of experiments. (author)

  20. Subchannel analysis code development for CANDU fuel channel

    International Nuclear Information System (INIS)

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

    1998-07-01

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

  1. Development of the New Light-Duty Hybrid Truck

    OpenAIRE

    Yamaguchi, Koichi

    2008-01-01

    Hino Motors, Ltd., developed the new light-duty hybrid truck whose traction motor, inverter, and traction battery were completely redesigned for maximizing output and efficiency. It also succeeds in balancing low fuel economy and low exhaust emissions by utilizing a combination of a new hybrid system control with a specially developed diesel engine.

  2. Development of bio-hybrid material based on Salmonella ...

    African Journals Online (AJOL)

    The immobilization of a whole microbial cell is an important process used in nanotechnology of biosensors and other related fields, especially the development of bio-hybrid materials based on live organisms and inorganic compounds. Here, we described an essay to develop a bio-hybrid material based on Salmonella ...

  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. Development of a nuclear power plant system analysis code

    International Nuclear Information System (INIS)

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

    1997-07-01

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

  5. HyDEn: A Hybrid Steganocryptographic Approach for Data Encryption Using Randomized Error-Correcting DNA Codes

    Directory of Open Access Journals (Sweden)

    Dan Tulpan

    2013-01-01

    Full Text Available This paper presents a novel hybrid DNA encryption (HyDEn approach that uses randomized assignments of unique error-correcting DNA Hamming code words for single characters in the extended ASCII set. HyDEn relies on custom-built quaternary codes and a private key used in the randomized assignment of code words and the cyclic permutations applied on the encoded message. Along with its ability to detect and correct errors, HyDEn equals or outperforms existing cryptographic methods and represents a promising in silico DNA steganographic approach.

  6. COSINE software development based on code generation technology

    International Nuclear Information System (INIS)

    Ren Hao; Mo Wentao; Liu Shuo; Zhao Guang

    2013-01-01

    The code generation technology can significantly improve the quality and productivity of software development and reduce software development risk. At present, the code generator is usually based on UML model-driven technology, which can not satisfy the development demand of nuclear power calculation software. The feature of scientific computing program was analyzed and the FORTRAN code generator (FCG) based on C# was developed in this paper. FCG can generate module variable definition FORTRAN code automatically according to input metadata. FCG also can generate memory allocation interface for dynamic variables as well as data access interface. FCG was applied to the core and system integrated engine for design and analysis (COSINE) software development. The result shows that FCG can greatly improve the development efficiency of nuclear power calculation software, and reduce the defect rate of software development. (authors)

  7. Hybrid Micro-Depletion method in the DYN3D code

    Energy Technology Data Exchange (ETDEWEB)

    Bilodid, Yurii [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Div. Reactor Safety

    2016-07-01

    A new method for accounting spectral history effects was developed and implemented in the reactor dynamics code DYN3D. Detailed nuclide content is calculated for each region of the reactor core and used to correct fuel properties. The new method demonstrates excellent results in test cases.

  8. Recent developments in the CONTAIN-LMR code

    International Nuclear Information System (INIS)

    Murata, K.K.

    1990-01-01

    Through an international collaborative effort, a special version of the CONTAIN code is being developed for integrated mechanistic analysis of the conditions in liquid metal reactor (LMR) containments during severe accidents. The capabilities of the most recent code version, CONTAIN LMR/1B-Mod.1, are discussed. These include new models for the treatment of two condensables, sodium condensation on aerosols, chemical reactions, hygroscopic aerosols, and concrete outgassing. This code version also incorporates all of the previously released LMR model enhancements. The results of an integral demonstration calculation of a sever core-melt accident scenario are given to illustrate the features of this code version. 11 refs., 7 figs., 1 tab

  9. Recent developments in KTF. Code optimization and improved numerics

    International Nuclear Information System (INIS)

    Jimenez, Javier; Avramova, Maria; Sanchez, Victor Hugo; Ivanov, Kostadin

    2012-01-01

    The rapid increase of computer power in the last decade facilitated the development of high fidelity simulations in nuclear engineering allowing a more realistic and accurate optimization as well as safety assessment of reactor cores and power plants compared to the legacy codes. Thermal hydraulic subchannel codes together with time dependent neutron transport codes are the options of choice for an accurate prediction of local safety parameters. Moreover, fast running codes with the best physical models are needed for high fidelity coupled thermal hydraulic / neutron kinetic solutions. Hence at KIT, different subchannel codes such as SUBCHANFLOW and KTF are being improved, validated and coupled with different neutron kinetics solutions. KTF is a subchannel code developed for best-estimate analysis of both Pressurized Water Reactor (PWR) and BWR. It is based on the Pennsylvania State University (PSU) version of COBRA-TF (Coolant Boling in Rod Arrays Two Fluids) named CTF. In this paper, the investigations devoted to the enhancement of the code numeric and informatics structure are presented and discussed. By some examples the gain on code speed-up will be demonstrated and finally an outlook of further activities concentrated on the code improvements will be given. (orig.)

  10. Recent developments in KTF. Code optimization and improved numerics

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, Javier; Avramova, Maria; Sanchez, Victor Hugo; Ivanov, Kostadin [Karlsruhe Institute of Technology (KIT) (Germany). Inst. for Neutron Physics and Reactor Technology (INR)

    2012-11-01

    The rapid increase of computer power in the last decade facilitated the development of high fidelity simulations in nuclear engineering allowing a more realistic and accurate optimization as well as safety assessment of reactor cores and power plants compared to the legacy codes. Thermal hydraulic subchannel codes together with time dependent neutron transport codes are the options of choice for an accurate prediction of local safety parameters. Moreover, fast running codes with the best physical models are needed for high fidelity coupled thermal hydraulic / neutron kinetic solutions. Hence at KIT, different subchannel codes such as SUBCHANFLOW and KTF are being improved, validated and coupled with different neutron kinetics solutions. KTF is a subchannel code developed for best-estimate analysis of both Pressurized Water Reactor (PWR) and BWR. It is based on the Pennsylvania State University (PSU) version of COBRA-TF (Coolant Boling in Rod Arrays Two Fluids) named CTF. In this paper, the investigations devoted to the enhancement of the code numeric and informatics structure are presented and discussed. By some examples the gain on code speed-up will be demonstrated and finally an outlook of further activities concentrated on the code improvements will be given. (orig.)

  11. Fluid-structure-interaction analyses of reactor vessel using improved hybrid Lagrangian Eulerian code ALICE-II

    Energy Technology Data Exchange (ETDEWEB)

    Wang, C.Y.

    1993-06-01

    This paper describes fluid-structure-interaction and structure response analyses of a reactor vessel subjected to loadings associated with postulated accidents, using the hybrid Lagrangian-Eulerian code ALICE-II. This code has been improved recently to accommodate many features associated with innovative designs of reactor vessels. Calculational capabilities have been developed to treat water in the reactor cavity outside the vessel, internal shield structures and internal thin shells. The objective of the present analyses is to study the cover response and potential for missile generation in response to a fuel-coolant interaction in the core region. Three calculations were performed using the cover weight as a parameter. To study the effect of the cavity water, vessel response calculations for both wet- and dry-cavity designs are compared. Results indicate that for all cases studied and for the design parameters assumed, the calculated cover displacements are all smaller than the bolts` ultimate displacement and no missile generation of the closure head is predicted. Also, solutions reveal that the cavity water of the wet-cavity design plays an important role of restraining the downward displacement of the bottom head. Based on these studies, the analyses predict that the structure integrity is maintained throughout the postulated accident for the wet-cavity design.

  12. Fluid-structure-interaction analyses of reactor vessel using improved hybrid Lagrangian Eulerian code ALICE-II

    Energy Technology Data Exchange (ETDEWEB)

    Wang, C.Y.

    1993-01-01

    This paper describes fluid-structure-interaction and structure response analyses of a reactor vessel subjected to loadings associated with postulated accidents, using the hybrid Lagrangian-Eulerian code ALICE-II. This code has been improved recently to accommodate many features associated with innovative designs of reactor vessels. Calculational capabilities have been developed to treat water in the reactor cavity outside the vessel, internal shield structures and internal thin shells. The objective of the present analyses is to study the cover response and potential for missile generation in response to a fuel-coolant interaction in the core region. Three calculations were performed using the cover weight as a parameter. To study the effect of the cavity water, vessel response calculations for both wet- and dry-cavity designs are compared. Results indicate that for all cases studied and for the design parameters assumed, the calculated cover displacements are all smaller than the bolts' ultimate displacement and no missile generation of the closure head is predicted. Also, solutions reveal that the cavity water of the wet-cavity design plays an important role of restraining the downward displacement of the bottom head. Based on these studies, the analyses predict that the structure integrity is maintained throughout the postulated accident for the wet-cavity design.

  13. Hybrid parallelization of the XTOR-2F code for the simulation of two-fluid MHD instabilities in tokamaks

    Science.gov (United States)

    Marx, Alain; Lütjens, Hinrich

    2017-03-01

    A hybrid MPI/OpenMP parallel version of the XTOR-2F code [Lütjens and Luciani, J. Comput. Phys. 229 (2010) 8130] solving the two-fluid MHD equations in full tokamak geometry by means of an iterative Newton-Krylov matrix-free method has been developed. The present work shows that the code has been parallelized significantly despite the numerical profile of the problem solved by XTOR-2F, i.e. a discretization with pseudo-spectral representations in all angular directions, the stiffness of the two-fluid stability problem in tokamaks, and the use of a direct LU decomposition to invert the physical pre-conditioner at every Krylov iteration of the solver. The execution time of the parallelized version is an order of magnitude smaller than the sequential one for low resolution cases, with an increasing speedup when the discretization mesh is refined. Moreover, it allows to perform simulations with higher resolutions, previously forbidden because of memory limitations.

  14. Development and Verification of Behavior of Tritium Analytic Code (BOTANIC)

    International Nuclear Information System (INIS)

    Park, Min Young; Kim, Eung Soo

    2014-01-01

    VHTR, one of the Generation IV reactor concepts, has a relatively high operation temperature and is usually suggested as a heat source for many industrial processes, including hydrogen production process. Thus, it is vital to trace tritium behavior in the VHTR system and the potential permeation rate to the industrial process. In other words, tritium is a crucial issue in terms of safety in the fission reactor system. Therefore, it is necessary to understand the behavior of tritium and the development of the tool to enable this is vital.. In this study, a Behavior of Tritium Analytic Code (BOTANIC) an analytic tool which is capable of analyzing tritium behavior is developed using a chemical process code called gPROMS. BOTANIC was then further verified using the analytic solutions and benchmark codes such as Tritium Permeation Analysis Code (TPAC) and COMSOL. In this study, the Behavior of Tritium Analytic Code, BOTANIC, has been developed using a chemical process code called gPROMS. The code has several distinctive features including non-diluted assumption, flexible applications and adoption of distributed permeation model. Due to these features, BOTANIC has the capability to analyze a wide range of tritium level systems and has a higher accuracy as it has the capacity to solve distributed models. BOTANIC was successfully developed and verified using analytical solution and the benchmark code calculation result. The results showed very good agreement with the analytical solutions and the calculation results of TPAC and COMSOL. Future work will be focused on the total system verification

  15. Developments of fuel performance analysis codes in KEPCO NF

    International Nuclear Information System (INIS)

    Han, H. T.; Choi, J. M.; Jung, C. D.; Yoo, J. S.

    2012-01-01

    The KEPCO NF has developed fuel performance analysis and design code named as ROPER, and utility codes of XGCOL and XDNB in order to perform fuel rod design evaluation for Korean nuclear power plants. The ROPER code intends to cover full range of fuel performance evaluation. The XGCOL code is for the clad flattening evaluation and the XDNB code is for the extensive DNB propagation evaluation. In addition to these, the KEPCO NF is now in the developing stage for 3-dimensional fuel performance analysis code, named as OPER3D, using 3-dimensional FEM for the nest generation within the joint project CANDU ENERGY in order to analyze PCMI behavior and fuel performance under load following operation. Of these, the ROPER code is now in the stage of licensing activities by Korean regulatory body and the other two are almost in the final developing stage. After finishing the developing, licensing activities are to be performed. These activities are intending to acquire competitiveness, originality, vendor-free ownership of fuel performance codes in the KEPCO NF

  16. Twelve gordian knots when developing an organizational code of ethics

    NARCIS (Netherlands)

    Kaptein, Muel; Wempe, Johan

    1998-01-01

    Following the example of the many organizations in the United States which have a code of ethics, an increasing interest on the part of companies, trade organizations, (semi-)governmental organizations and professions in the Netherlands to develop codes of ethics can be witnessed. We have been able

  17. Graphical user interface development for the MARS code

    International Nuclear Information System (INIS)

    Jeong, J.-J.; Hwang, M.; Lee, Y.J.; Kim, K.D.; Chung, B.D.

    2003-01-01

    KAERI has developed the best-estimate thermal-hydraulic system code MARS using the RELAP5/MOD3 and COBRA-TF codes. To exploit the excellent features of the two codes, we consolidated the two codes. Then, to improve the readability, maintainability, and portability of the consolidated code, all the subroutines were completely restructured by employing a modular data structure. At present, a major part of the MARS code development program is underway to improve the existing capabilities. The code couplings with three-dimensional neutron kinetics, containment analysis, and transient critical heat flux calculations have also been carried out. At the same time, graphical user interface (GUI) tools have been developed for user friendliness. This paper presents the main features of the MARS GUI. The primary objective of the GUI development was to provide a valuable aid for all levels of MARS users in their output interpretation and interactive controls. Especially, an interactive control function was designed to allow operator actions during simulation so that users can utilize the MARS code like conventional nuclear plant analyzers (NPAs). (author)

  18. Analysis on applicable error-correcting code strength of storage class memory and NAND flash in hybrid storage

    Science.gov (United States)

    Matsui, Chihiro; Kinoshita, Reika; Takeuchi, Ken

    2018-04-01

    A hybrid of storage class memory (SCM) and NAND flash is a promising technology for high performance storage. Error correction is inevitable on SCM and NAND flash because their bit error rate (BER) increases with write/erase (W/E) cycles, data retention, and program/read disturb. In addition, scaling and multi-level cell technologies increase BER. However, error-correcting code (ECC) degrades storage performance because of extra memory reading and encoding/decoding time. Therefore, applicable ECC strength of SCM and NAND flash is evaluated independently by fixing ECC strength of one memory in the hybrid storage. As a result, weak BCH ECC with small correctable bit is recommended for the hybrid storage with large SCM capacity because SCM is accessed frequently. In contrast, strong and long-latency LDPC ECC can be applied to NAND flash in the hybrid storage with large SCM capacity because large-capacity SCM improves the storage performance.

  19. Current State of Military Hybrid Vehicle Development

    Science.gov (United States)

    2011-01-01

    programme were: • EP-50 parallel hybrid light armoured vehicle (LAV-III) and a refuse hauler • advanced hybrid electric drive (AHED) 8 × 8, 20 ton...herein do not necessarily state or reflect those of the US Government or the DoA, and shall not be used for advertising or product endorsement purposes. 2 Unclassified. Distribution Statement A. Approved for public release.

  20. Development of the integrated system reliability analysis code MODULE

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  1. Development of realistic thermal hydraulic system analysis code

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-05-01

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

  2. Development of realistic thermal hydraulic system analysis code

    International Nuclear Information System (INIS)

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

    2002-05-01

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

  3. Foundational development of an advanced nuclear reactor integrated safety code

    International Nuclear Information System (INIS)

    Clarno, Kevin; Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth; Hooper, Russell Warren; Humphries, Larry LaRon

    2010-01-01

    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

  4. Foundational development of an advanced nuclear reactor integrated safety code.

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin (Oak Ridge National Laboratory, Oak Ridge, TN); Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth (Ktech Corporation, Albuquerque, NM); Hooper, Russell Warren; Humphries, Larry LaRon

    2010-02-01

    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  6. Development of the code for filter calculation

    International Nuclear Information System (INIS)

    Gritzay, O.O.; Vakulenko, M.M.

    2012-01-01

    This paper describes a calculation method, which commonly used in the Neutron Physics Department to develop a new neutron filter or to improve the existing neutron filter. This calculation is the first step of the traditional filter development procedure. It allows easy selection of the qualitative and quantitative contents of a composite filter in order to receive the filtered neutron beam with given parameters

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

    International Nuclear Information System (INIS)

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

    2011-03-01

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

  8. Development of Regulatory Audit Core Safety Code : COREDAX

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Chae Yong; Jo, Jong Chull; Roh, Byung Hwan [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Jae Jun; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2005-07-01

    Korea Institute of Nuclear Safety (KINS) has developed a core neutronics simulator, COREDAX code, for verifying core safety of SMART-P reactor, which is technically supported by Korea Advanced Institute of Science and Technology (KAIST). The COREDAX code would be used for regulatory audit calculations of 3- dimendional core neutronics. The COREDAX code solves the steady-state and timedependent multi-group neutron diffusion equation in hexagonal geometry as well as rectangular geometry by analytic function expansion nodal (AFEN) method. AFEN method was developed at KAIST, and it was internationally verified that its accuracy is excellent. The COREDAX code is originally programmed based on the AFEN method. Accuracy of the code on the AFEN method was excellent for the hexagonal 2-dimensional problems, but there was a need for improvement for hexagonal-z 3-dimensional problems. Hence, several solution routines of the AFEN method are improved, and finally the advanced AFEN method is created. COREDAX code is based on the advanced AFEN method . The initial version of COREDAX code is to complete a basic framework, performing eigenvalue calculations and kinetics calculations with thermal-hydraulic feedbacks, for audit calculations of steady-state core design and reactivity-induced accidents of SMART-P reactor. This study describes the COREDAX code for hexagonal geometry.

  9. Development of probabilistic fracture mechanics code PASCAL and user's manual

    Energy Technology Data Exchange (ETDEWEB)

    Shibata, Katsuyuki; Onizawa, Kunio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Li, Yinsheng; Kato, Daisuke [Fuji Research Institute Corporation, Tokyo (Japan)

    2001-03-01

    As a part of the aging and structural integrity research for LWR components, a new PFM (Probabilistic Fracture Mechanics) code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed since FY1996. This code evaluates the failure probability of an aged reactor pressure vessel subjected to transient loading such as PTS (Pressurized Thermal Shock). The development of the code has been aimed to improve the accuracy and reliability of analysis by introducing new analysis methodologies and algorithms considering the recent development in the fracture mechanics methodologies and computer performance. The code has some new functions in optimized sampling and cell dividing procedure in stratified Monte Carlo simulation, elastic-plastic fracture criterion of R6 method, extension analysis models in semi-elliptical crack, evaluation of effect of thermal annealing and etc. In addition, an input data generator of temperature and stress distribution time histories was also prepared in the code. Functions and performance of the code have been confirmed based on the verification analyses and some case studies on the influence parameters. The present phase of the development will be completed in FY2000. Thus this report provides the user's manual and theoretical background of the code. (author)

  10. Development of the CRIPTE Code for Electromagnetic Coupling

    National Research Council Canada - National Science Library

    Parmantier, Jean-Philippe

    2005-01-01

    .... This code was originally developed as part of an experiment performed under the joint US-France international data exchange program on the atmospheric electricity/aircraft interactions, DEA-AF-79-7336...

  11. Hybrid information privacy system: integration of chaotic neural network and RSA coding

    Science.gov (United States)

    Hsu, Ming-Kai; Willey, Jeff; Lee, Ting N.; Szu, Harold H.

    2005-03-01

    Electronic mails are adopted worldwide; most are easily hacked by hackers. In this paper, we purposed a free, fast and convenient hybrid privacy system to protect email communication. The privacy system is implemented by combining private security RSA algorithm with specific chaos neural network encryption process. The receiver can decrypt received email as long as it can reproduce the specified chaos neural network series, so called spatial-temporal keys. The chaotic typing and initial seed value of chaos neural network series, encrypted by the RSA algorithm, can reproduce spatial-temporal keys. The encrypted chaotic typing and initial seed value are hidden in watermark mixed nonlinearly with message media, wrapped with convolution error correction codes for wireless 3rd generation cellular phones. The message media can be an arbitrary image. The pattern noise has to be considered during transmission and it could affect/change the spatial-temporal keys. Since any change/modification on chaotic typing or initial seed value of chaos neural network series is not acceptable, the RSA codec system must be robust and fault-tolerant via wireless channel. The robust and fault-tolerant properties of chaos neural networks (CNN) were proved by a field theory of Associative Memory by Szu in 1997. The 1-D chaos generating nodes from the logistic map having arbitrarily negative slope a = p/q generating the N-shaped sigmoid was given first by Szu in 1992. In this paper, we simulated the robust and fault-tolerance properties of CNN under additive noise and pattern noise. We also implement a private version of RSA coding and chaos encryption process on messages.

  12. Development of code SFINEL (Spent fuel integrity evaluator)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yong Soo; Min, Chin Young; Ohk, Young Kil; Yang, Yong Sik; Kim, Dong Ju; Kim, Nam Ku [Hanyang University, Seoul (Korea)

    1999-01-01

    SFINEL code, an integrated computer program for predicting the spent fuel rod integrity based on burn-up history and major degradation mechanisms, has been developed through this project. This code can sufficiently simulate the power history of a fuel rod during the reactor operation and estimate the degree of deterioration of spent fuel cladding using the recently-developed models on the degradation mechanisms. SFINEL code has been thoroughly benchmarked against the collected in-pile data and operating experiences: deformation and rupture, and cladding oxidation, rod internal pressure creep, then comprehensive whole degradation process. (author). 75 refs., 51 figs., 5 tabs.

  13. Development and application of methods to characterize code uncertainty

    International Nuclear Information System (INIS)

    Wilson, G.E.; Burtt, J.D.; Case, G.S.; Einerson, J.J.; Hanson, R.G.

    1985-01-01

    The United States Nuclear Regulatory Commission sponsors both international and domestic studies to assess its safety analysis codes. The Commission staff intends to use the results of these studies to quantify the uncertainty of the codes with a statistically based analysis method. Development of the methodology is underway. The Idaho National Engineering Laboratory contributions to the early development effort, and testing of two candidate methods are the subjects of this paper

  14. Theoretical Atomic Physics code development IV: LINES, A code for computing atomic line spectra

    International Nuclear Information System (INIS)

    Abdallah, J. Jr.; Clark, R.E.H.

    1988-12-01

    A new computer program, LINES, has been developed for simulating atomic line emission and absorption spectra using the accurate fine structure energy levels and transition strengths calculated by the (CATS) Cowan Atomic Structure code. Population distributions for the ion stages are obtained in LINES by using the Local Thermodynamic Equilibrium (LTE) model. LINES is also useful for displaying the pertinent atomic data generated by CATS. This report describes the use of LINES. Both CATS and LINES are part of the Theoretical Atomic PhysicS (TAPS) code development effort at Los Alamos. 11 refs., 9 figs., 1 tab

  15. Code-first development with Entity Framework

    CERN Document Server

    Barskiy, Sergey

    2015-01-01

    This book is intended for software developers with some prior experience with the Microsoft .NET framework who want to learn how to use Entity Framework. This book will get you up and running quickly, providing many examples that illustrate all the key concepts of Entity Framework.

  16. 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.

  17. Aeroelastic code development activities in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Wright, A.D. [National Renewable Energy Lab., Golden, Colorado (United States)

    1996-09-01

    Designing wind turbines to be fatigue resistant and to have long lifetimes at minimal cost is a major goal of the federal wind program and the wind industry in the United States. To achieve this goal, we must be able to predict critical loads for a wide variety of different wind turbines operating under extreme conditions. The codes used for wind turbine dynamic analysis must be able to analyze a wide range of different wind turbine configurations as well as rapidly predict the loads due to turbulent wind inflow with a minimal set of degrees of freedom. Code development activities in the US have taken a two-pronged approach in order to satisfy both of these criteria: (1) development of a multi-purpose code which can be used to analyze a wide variety of wind turbine configurations without having to develop new equations of motion with each configuration change, and (2) development of specialized codes with minimal sets of specific degrees of freedom for analysis of two- and three-bladed horizontal axis wind turbines and calculation of machine loads due to turbulent inflow. In the first method we have adapted a commercial multi-body dynamics simulation package for wind turbine analysis. In the second approach we are developing specialized codes with limited degrees of freedom, usually specified in the modal domain. This paper will summarize progress to date in the development, validation, and application of these codes. (au) 13 refs.

  18. Collaborative Multi-Layer Network Coding For Hybrid Cellular Cognitive Radio Networks

    KAUST Repository

    Moubayed, Abdallah J.

    2014-05-01

    In this thesis, as an extension to [1], we propose a prioritized multi-layer network coding scheme for collaborative packet recovery in hybrid (interweave and underlay) cellular cognitive radio networks. This scheme allows the uncoordinated collaboration between the collocated primary and cognitive radio base-stations in order to minimize their own as well as each other’s packet recovery overheads, thus by improving their throughput. The proposed scheme ensures that each network’s performance is not degraded by its help to the other network. Moreover, it guarantees that the primary network’s interference threshold is not violated in the same and adjacent cells. Yet, the scheme allows the reduction of the recovery overhead in the collocated primary and cognitive radio networks. The reduction in the cognitive radio network is further amplified due to the perfect detection of spectrum holes which allows the cognitive radio base station to transmit at higher power without fear of violating the interference threshold of the primary network. For the secondary network, simulation results show reductions of 20% and 34% in the packet recovery overhead, compared to the non-collaborative scheme, for low and high probabilities of primary packet arrivals, respectively. For the primary network, this reduction was found to be 12%. Furthermore, with the use of fractional cooperation, the average recovery overhead is further reduced by around 5% for the primary network and around 10% for the secondary network when a high fractional cooperation probability is used.

  19. Parallel deposition, sorting, and reordering methods in the Hybrid Ordered Plasma Simulation (HOPS) code

    International Nuclear Information System (INIS)

    Anderson, D.V.; Shumaker, D.E.

    1993-01-01

    From a computational standpoint, particle simulation calculations for plasmas have not adapted well to the transitions from scalar to vector processing nor from serial to parallel environments. They have suffered from inordinate and excessive accessing of computer memory and have been hobbled by relatively inefficient gather-scatter constructs resulting from the use of indirect indexing. Lastly, the many-to-one mapping characteristic of the deposition phase has made it difficult to perform this in parallel. The authors' code sorts and reorders the particles in a spatial order. This allows them to greatly reduce the memory references, to run in directly indexed vector mode, and to employ domain decomposition to achieve parallelization. In this hybrid simulation the electrons are modeled as a fluid and the field equations solved are obtained from the electron momentum equation together with the pre-Maxwell equations (displacement current neglected). Either zero or finite electron mass can be used in the electron model. The resulting field equations are solved with an iteratively explicit procedure which is thus trivial to parallelize. Likewise, the field interpolations and the particle pushing is simple to parallelize. The deposition, sorting, and reordering phases are less simple and it is for these that the authors present detailed algorithms. They have now successfully tested the parallel version of HOPS in serial mode and it is now being readied for parallel execution on the Cray C-90. They will then port HOPS to a massively parallel computer, in the next year

  20. Collaborative Multi-Layer Network Coding in Hybrid Cellular Cognitive Radio Networks

    KAUST Repository

    Moubayed, Abdallah J.

    2015-05-01

    In this paper, as an extension to [1], we propose a prioritized multi-layer network coding scheme for collaborative packet recovery in hybrid (interweave and underlay) cellular cognitive radio networks. This scheme allows the uncoordinated collaboration between the collocated primary and cognitive radio base-stations in order to minimize their own as well as each other\\'s packet recovery overheads, thus by improving their throughput. The proposed scheme ensures that each network\\'s performance is not degraded by its help to the other network. Moreover, it guarantees that the primary network\\'s interference threshold is not violated in the same and adjacent cells. Yet, the scheme allows the reduction of the recovery overhead in the collocated primary and cognitive radio networks. The reduction in the cognitive radio network is further amplified due to the perfect detection of spectrum holes which allows the cognitive radio base station to transmit at higher power without fear of violating the interference threshold of the primary network. For the secondary network, simulation results show reductions of 20% and 34% in the packet recovery overhead, compared to the non-collaborative scheme, for low and high probabilities of primary packet arrivals, respectively. For the primary network, this reduction was found to be 12%. © 2015 IEEE.

  1. Structural reliability methods: Code development status

    Science.gov (United States)

    Millwater, Harry R.; Thacker, Ben H.; Wu, Y.-T.; Cruse, T. A.

    1991-05-01

    The Probabilistic Structures Analysis Method (PSAM) program integrates state of the art probabilistic algorithms with structural analysis methods in order to quantify the behavior of Space Shuttle Main Engine structures subject to uncertain loadings, boundary conditions, material parameters, and geometric conditions. An advanced, efficient probabilistic structural analysis software program, NESSUS (Numerical Evaluation of Stochastic Structures Under Stress) was developed as a deliverable. NESSUS contains a number of integrated software components to perform probabilistic analysis of complex structures. A nonlinear finite element module NESSUS/FEM is used to model the structure and obtain structural sensitivities. Some of the capabilities of NESSUS/FEM are shown. A Fast Probability Integration module NESSUS/FPI estimates the probability given the structural sensitivities. A driver module, PFEM, couples the FEM and FPI. NESSUS, version 5.0, addresses component reliability, resistance, and risk.

  2. Development of computer code in PNC, 8

    International Nuclear Information System (INIS)

    Ohhira, Mitsuru

    1990-01-01

    Private buildings applied base isolation system, are on the practical stage now. So, under Construction and Maintenance Management Office, we are doing an application study of base isolation system to nuclear fuel facilities. On the process of this study, we have developed Dynamic Analysis Program-Base Isolation System (DAP-BS) which is able to run a 32-bit personal computer. Using this program, we can analyze a 3-dimensional structure, and evaluate the various properties of base isolation parts that are divided into maximum 16 blocks. And from the results of some simulation analyses, we thought that DAP-BS had good reliability and marketability. So, we put DAP-BS on the market. (author)

  3. Development and validation of a nodal code for core calculation

    International Nuclear Information System (INIS)

    Nowakowski, Pedro Mariano

    2004-01-01

    The code RHENO solves the multigroup three-dimensional diffusion equation using a nodal method of polynomial expansion.A comparative study has been made between this code and present internationals nodal diffusion codes, resulting that the RHENO is up to date.The RHENO has been integrated to a calculation line and has been extend to make burnup calculations.Two methods for pin power reconstruction were developed: modulation and imbedded. The modulation method has been implemented in a program, while the implementation of the imbedded method will be concluded shortly.The validation carried out (that includes experimental data of a MPR) show very good results and calculation efficiency

  4. Development and validation of a fuel performance analysis code

    International Nuclear Information System (INIS)

    Majalee, Aaditya V.; Chaturvedi, S.

    2015-01-01

    CAD has been developing a computer code 'FRAVIZ' for calculation of steady-state thermomechanical behaviour of nuclear reactor fuel rods. It contains four major modules viz., Thermal module, Fission Gas Release module, Material Properties module and Mechanical module. All these four modules are coupled to each other and feedback from each module is fed back to others to get a self-consistent evolution in time. The computer code has been checked against two FUMEX benchmarks. Modelling fuel performance in Advance Heavy Water Reactor would require additional inputs related to the fuel and some modification in the code.(author)

  5. Development and validation of sodium fire analysis code ASSCOPS

    International Nuclear Information System (INIS)

    Ohno, Shuji

    2001-01-01

    A version 2.1 of the ASSCOPS sodium fire analysis code was developed to evaluate the thermal consequences of a sodium leak and consequent fire in LMFBRs. This report describes the computational models and the validation studies using the code. The ASSCOPS calculates sodium droplet and pool fire, and consequential heat/mass transfer behavior. Analyses of sodium pool or spray fire experiments confirmed that this code and parameters used in the validation studies gave valid results on the thermal consequences of sodium leaks and fires. (author)

  6. Development of safety analysis codes for light water reactor

    International Nuclear Information System (INIS)

    Akimoto, Masayuki

    1985-01-01

    An overview is presented of currently used major codes for the prediction of thermohydraulic transients in nuclear power plants. The overview centers on the two-phase fluid dynamics of the coolant system and the assessment of the codes. Some of two-phase phenomena such as phase separation are not still predicted with engineering accuracy. MINCS-PIPE are briefly introduced. The MINCS-PIPE code is to assess constitutive relations and to aid development of various experimental correlations for 1V1T model to 2V2T model. (author)

  7. SCDAP/RELAP5/MOD3 code development

    International Nuclear Information System (INIS)

    Allison, C.M.; Siefken, J.L.; Coryell, E.W.

    1992-01-01

    The SCOAP/RELAP5/MOD3 computer code is designed to describe the overall reactor coolant system (RCS) thermal-hydraulic response, core damage progression, and fission product release and transport during severe accidents. The code is being developed at the Idaho National Engineering Laboratory (INEL) under the primary sponsorship of the Office of Nuclear Regulatory Research of the US Nuclear Regulatory Commission (NRC). Code development activities are currently focused on three main areas - (a) code usability, (b) early phase melt progression model improvements, and (c) advanced reactor thermal-hydraulic model extensions. This paper describes the first two activities. A companion paper describes the advanced reactor model improvements being performed under RELAP5/MOD3 funding

  8. 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.)

  9. SCDAP/RELAP5 code development and assessment

    International Nuclear Information System (INIS)

    Allison, C.M.; Hohorst, J.K.

    1996-01-01

    The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The current version of the code is SCDAP/RELAP5/MOD3.1e. Although MOD3.1e contains a number of significant improvements since the initial version of MOD3.1 was released, new models to treat the behavior of the fuel and cladding during reflood have had the most dramatic impact on the code's calculations. This paper provides a brief description of the new reflood models, presents highlights of the assessment of the current version of MOD3.1, and discusses future SCDAP/RELAP5/MOD3.2 model development activities

  10. Development of in-vessel source term analysis code, tracer

    International Nuclear Information System (INIS)

    Miyagi, K.; Miyahara, S.

    1996-01-01

    Analyses of radionuclide transport in fuel failure accidents (generally referred to source terms) are considered to be important especially in the severe accident evaluation. The TRACER code has been developed to realistically predict the time dependent behavior of FPs and aerosols within the primary cooling system for wide range of fuel failure events. This paper presents the model description, results of validation study, the recent model advancement status of the code, and results of check out calculations under reactor conditions. (author)

  11. Computer codes developed in FRG to analyse hypothetical meltdown accidents

    International Nuclear Information System (INIS)

    Hassmann, K.; Hosemann, J.P.; Koerber, H.; Reineke, H.

    1978-01-01

    It is the purpose of this paper to give the status of all significant computer codes developed in the core melt-down project which is incorporated in the light water reactor safety research program of the Federal Ministry of Research and Technology. For standard pressurized water reactors, results of some computer codes will be presented, describing the course and the duration of the hypothetical core meltdown accident. (author)

  12. Develop a Hybrid Coordinate Ocean Model with Data Assimilation Capabilities

    National Research Council Canada - National Science Library

    Thacker, W. C

    2003-01-01

    .... The objectives of the research are as follows: (1) to develop a methodology for assimilating temperature and salinity profiles from XBT, CTD, and ARGO float data that accommodates the peculiarities of HYCOM's hybrid vertical coordinates, allowing...

  13. Agronomic potentials of quality protein maize hybrids developed in ...

    African Journals Online (AJOL)

    Agronomic potentials of quality protein maize hybrids developed in Ghana. ... Ghana Journal of Agricultural Science ... and Wheat Improvement (CIMMYT) were evaluated on research stations and in farmers\\' fields in Ghana from 1995 to 1996.

  14. A Hybrid Scheme Based on Pipelining and Multitasking in Mobile Application Processors for Advanced Video Coding

    Directory of Open Access Journals (Sweden)

    Muhammad Asif

    2015-01-01

    Full Text Available One of the key requirements for mobile devices is to provide high-performance computing at lower power consumption. The processors used in these devices provide specific hardware resources to handle computationally intensive video processing and interactive graphical applications. Moreover, processors designed for low-power applications may introduce limitations on the availability and usage of resources, which present additional challenges to the system designers. Owing to the specific design of the JZ47x series of mobile application processors, a hybrid software-hardware implementation scheme for H.264/AVC encoder is proposed in this work. The proposed scheme distributes the encoding tasks among hardware and software modules. A series of optimization techniques are developed to speed up the memory access and data transferring among memories. Moreover, an efficient data reusage design is proposed for the deblock filter video processing unit to reduce the memory accesses. Furthermore, fine grained macroblock (MB level parallelism is effectively exploited and a pipelined approach is proposed for efficient utilization of hardware processing cores. Finally, based on parallelism in the proposed design, encoding tasks are distributed between two processing cores. Experiments show that the hybrid encoder is 12 times faster than a highly optimized sequential encoder due to proposed techniques.

  15. The Development of the World Anti-Doping Code.

    Science.gov (United States)

    Young, Richard

    2017-01-01

    This chapter addresses both the development and substance of the World Anti-Doping Code, which came into effect in 2003, as well as the subsequent Code amendments, which came into effect in 2009 and 2015. Through an extensive process of stakeholder input and collaboration, the World Anti-Doping Code has transformed the hodgepodge of inconsistent and competing pre-2003 anti-doping rules into a harmonized and effective approach to anti-doping. The Code, as amended, is now widely recognized worldwide as the gold standard in anti-doping. The World Anti-Doping Code originally went into effect on January 1, 2004. The first amendments to the Code went into effect on January 1, 2009, and the second amendments on January 1, 2015. The Code and the related international standards are the product of a long and collaborative process designed to make the fight against doping more effective through the adoption and implementation of worldwide harmonized rules and best practices. © 2017 S. Karger AG, Basel.

  16. Cooperation of experts' opinion, experiment and computer code development

    International Nuclear Information System (INIS)

    Wolfert, K.; Hicken, E.

    The connection between code development, code assessment and confidence in the analysis of transients will be discussed. In this manner, the major sources of errors in the codes and errors in applications of the codes will be shown. Standard problem results emphasize that, in order to have confidence in licensing statements, the codes must be physically realistic and the code user must be qualified and experienced. We will discuss why there is disagreement between the licensing authority and vendor concerning assessment of the fullfillment of safety goal requirements. The answer to the question lies in the different confidence levels of the assessment of transient analysis. It is expected that a decrease in the disagreement will result from an increased confidence level. Strong efforts will be made to increase this confidence level through improvements in the codes, experiments and related organizational strcutures. Because of the low probability for loss-of-coolant-accidents in the nuclear industry, assessment must rely on analytical techniques and experimental investigations. (orig./HP) [de

  17. 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

  18. Developing a Blended Type Course of Introduction to Hybrid Vehicles

    Directory of Open Access Journals (Sweden)

    Na Zhu

    2016-02-01

    Full Text Available An innovative course of introduction to hybrid vehicles is developed for both associate and bachelor degree programs for engineering technology with automotive/mechanical concentration. The hybrid vehicle course content includes several topics, such as the rational of pure electric vehicle and hybrid vehicle, hybrid vehicle propulsion systems, fundamentals of motor/generator systems, fundamentals of battery and energy management system, and introduction to various configurations of hybrid vehicle systems available in market and under development. Hybrid vehicle technology is a new area and developed rapidly in the field of automotive and mechanical engineering. Students need not only the fundamentals and concepts from college, but also the ability to keep up with the latest technology after their graduation. Therefore, a blended course type is employed to help students have a better understanding of the fundamentals of hybrid vehicle and developing their self-studying ability. Topics in the course have three steps of learning. Firstly, on-ground lecture is given in class, where the instructor explains basic knowledge, such as principles, equations, and design rules.  In this way, the students will have enough background knowledge and be able to conduct further self-reading and research work. Secondly, students are required to go to university’s desire to learn (D2L online system and finish the online part of the topic. In the D2L system, students will find a quiz and its supporting materials. Thirdly, students come back to the on-ground lecture and discuss the quiz in groups with instructor. After the discussion, the instructor gives students a conclusion of the topic and moves forward to the next topic. A computer simulation class is also given to help student better understand the operation strategies of the hybrid vehicle systems and have a trial of design of hybrid vehicle.

  19. Hybrid PAPR reduction scheme with Huffman coding and DFT-spread technique for direct-detection optical OFDM systems

    Science.gov (United States)

    Peng, Miao; Chen, Ming; Zhou, Hui; Wan, Qiuzhen; Jiang, LeYong; Yang, Lin; Zheng, Zhiwei; Chen, Lin

    2018-01-01

    High peak-to-average power ratio (PAPR) of the transmit signal is a major drawback in optical orthogonal frequency division multiplexing (OOFDM) system. In this paper, we propose and experimentally demonstrate a novel hybrid scheme, combined the Huffman coding and Discrete Fourier Transmission-Spread (DFT-spread), in order to reduce high PAPR in a 16-QAM short-reach intensity-modulated and direct-detection OOFDM (IMDD-OOFDM) system. The experimental results demonstrated that the hybrid scheme can reduce the PAPR by about 1.5, 2, 3 and 6 dB, and achieve 1.5, 1, 2.5 and 3 dB receiver sensitivity improvement compared to clipping, DFT-spread and Huffman coding and original OFDM signals, respectively, at an error vector magnitude (EVM) of -10 dB after transmission over 20 km standard single-mode fiber (SSMF). Furthermore, the throughput gain can be of the order of 30% by using the hybrid scheme compared with the cases of without applying the Huffman coding.

  20. Development and Meiosis of Three Interspecific Hybrids with Cultivated Barley (Hordeum vulgare L.)

    DEFF Research Database (Denmark)

    Von Bothmer, R.; Flink, J.; Linde-Laursen, Ib

    1986-01-01

    The development and meiosis of three interspecific hybrids between cultivated barley (Hordeum vulgare L.) and H. secalinum Schreb., H. tetraploidum Covas, and H. parodii Covas, respectively, were studied. All three hybrid combinations developed very slowly vegetatively. Meiosis of the hybrids...

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

    International Nuclear Information System (INIS)

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

    2015-07-01

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

  2. Development of a new EMP code at LANL

    Science.gov (United States)

    Colman, J. J.; Roussel-Dupré, R. A.; Symbalisty, E. M.; Triplett, L. A.; Travis, B. J.

    2006-05-01

    A new code for modeling the generation of an electromagnetic pulse (EMP) by a nuclear explosion in the atmosphere is being developed. The source of the EMP is the Compton current produced by the prompt radiation (γ-rays, X-rays, and neutrons) of the detonation. As a first step in building a multi- dimensional EMP code we have written three kinetic codes, Plume, Swarm, and Rad. Plume models the transport of energetic electrons in air. The Plume code solves the relativistic Fokker-Planck equation over a specified energy range that can include ~ 3 keV to 50 MeV and computes the resulting electron distribution function at each cell in a two dimensional spatial grid. The energetic electrons are allowed to transport, scatter, and experience Coulombic drag. Swarm models the transport of lower energy electrons in air, spanning 0.005 eV to 30 keV. The swarm code performs a full 2-D solution to the Boltzmann equation for electrons in the presence of an applied electric field. Over this energy range the relevant processes to be tracked are elastic scattering, three body attachment, two body attachment, rotational excitation, vibrational excitation, electronic excitation, and ionization. All of these occur due to collisions between the electrons and neutral bodies in air. The Rad code solves the full radiation transfer equation in the energy range of 1 keV to 100 MeV. It includes effects of photo-absorption, Compton scattering, and pair-production. All of these codes employ a spherical coordinate system in momentum space and a cylindrical coordinate system in configuration space. The "z" axis of the momentum and configuration spaces is assumed to be parallel and we are currently also assuming complete spatial symmetry around the "z" axis. Benchmarking for each of these codes will be discussed as well as the way forward towards an integrated modern EMP code.

  3. 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.

  4. 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

  5. Multi-dimensional Code Development for Safety Analysis of LMR

    International Nuclear Information System (INIS)

    Ha, K. S.; Jeong, H. Y.; Kwon, Y. M.; Lee, Y. B.

    2006-08-01

    A liquid metal reactor loaded a metallic fuel has the inherent safety mechanism due to the several negative reactivity feedback. Although this feature demonstrated through experiments in the EBR-II, any of the computer programs until now did not exactly analyze it because of the complexity of the reactivity feedback mechanism. A multi-dimensional detail program was developed through the International Nuclear Energy Research Initiative(INERI) from 2003 to 2005. This report includes the numerical coupling the multi-dimensional program and SSC-K code which is used to the safety analysis of liquid metal reactors in KAERI. The coupled code has been proved by comparing the analysis results using the code with the results using SAS-SASSYS code of ANL for the UTOP, ULOF, and ULOHS applied to the safety analysis for KALIMER-150

  6. Development of 1D Liner Compression Code for IDL

    Science.gov (United States)

    Shimazu, Akihisa; Slough, John; Pancotti, Anthony

    2015-11-01

    A 1D liner compression code is developed to model liner implosion dynamics in the Inductively Driven Liner Experiment (IDL) where FRC plasmoid is compressed via inductively-driven metal liners. The driver circuit, magnetic field, joule heating, and liner dynamics calculations are performed at each time step in sequence to couple these effects in the code. To obtain more realistic magnetic field results for a given drive coil geometry, 2D and 3D effects are incorporated into the 1D field calculation through use of correction factor table lookup approach. Commercial low-frequency electromagnetic fields solver, ANSYS Maxwell 3D, is used to solve the magnetic field profile for static liner condition at various liner radius in order to derive correction factors for the 1D field calculation in the code. The liner dynamics results from the code is verified to be in good agreement with the results from commercial explicit dynamics solver, ANSYS Explicit Dynamics, and previous liner experiment. The developed code is used to optimize the capacitor bank and driver coil design for better energy transfer and coupling. FRC gain calculations are also performed using the liner compression data from the code for the conceptual design of the reactor sized system for fusion energy gains.

  7. Development of REFLA/TRAC code for engineering work station

    International Nuclear Information System (INIS)

    Ohnuki, Akira; Akimoto, Hajime; Murao, Yoshio

    1994-03-01

    The REFLA/TRAC code is a best-estimate code which is expected to check reactor safety analysis codes for light water reactors (LWRs) and to perform accident analyses for LWRs and also for an advanced LWR. Therefore, a high predictive capability is required and the assessment of each physical model becomes important because the models govern the predictive capability. In the case of the assessment of three-dimensional models in REFLA/TRAC code, a conventional large computer is being used and it is difficult to perform the assessment efficiently because the turnaround time for the calculation and the analysis is long. Then, a REFLA/TRAC code which can run on an engineering work station (EWS) was developed. Calculational speed of the current EWS is the same order as that of large computers and the EWS has an excellent function for multidimensional graphical drawings. Besides, the plotting processors for X-Y drawing and for two-dimensional graphical drawing were developed in order to perform efficient analyses for three-dimensional calculations. In future, we can expect that the assessment of three-dimensional models becomes more efficient by introducing an EWS with higher calculational speed and with improved graphical drawings. In this report, each outline for the following three programs is described: (1) EWS version of REFLA/TRAC code, (2) Plot processor for X-Y drawing and (3) Plot processor for two-dimensional graphical drawing. (author)

  8. 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

  9. Development of a national code of practice for structural masonry ...

    African Journals Online (AJOL)

    The problems and constraints faced by most developing countries, particularly Ghana, in developing codes of practice for structural masonry are highlighted. The steps that must be undertaken through the coordinated efforts of the National Standards Boards, Research Institutions, Universities and Professional Bodies in the ...

  10. Development of FBR integrity system code. Basic concept

    International Nuclear Information System (INIS)

    Asayama, Tai

    2001-05-01

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

  11. Development of throughflow calculation code for axial flow compressors

    International Nuclear Information System (INIS)

    Kim, Ji Hwan; Kim, Hyeun Min; No, Hee Cheon

    2005-01-01

    The power conversion systems of the current HTGRs are based on closed Brayton cycle and major concern is thermodynamic performance of the axial flow helium gas turbines. Particularly, the helium compressor has some unique design challenges compared to the air-breathing compressor such as high hub-to-tip ratios throughout the machine and a large number of stages due to the physical property of the helium and thermodynamic cycle. Therefore, it is necessary to develop a design and analysis code for helium compressor that can estimate the design point and off-design performance accurately. KAIST nuclear system laboratory has developed a compressor design and analysis code by means of throughflow calculation and several loss models. This paper presents the outline of the development of a throughflow calculation code and its verification results

  12. Hybrid mesh finite volume CFD code for studying heat transfer in a forward-facing step

    Energy Technology Data Exchange (ETDEWEB)

    Jayakumar, J S; Kumar, Inder [Bhabha Atomic Research Center, Mumbai (India); Eswaran, V, E-mail: jsjayan@gmail.com, E-mail: inderk@barc.gov.in, E-mail: eswar@iitk.ac.in [Indian Institute of Technology, Kanpur (India)

    2010-12-15

    Computational fluid dynamics (CFD) methods employ two types of grid: structured and unstructured. Developing the solver and data structures for a finite-volume solver is easier than for unstructured grids. But real-life problems are too complicated to be fitted flexibly by structured grids. Therefore, unstructured grids are widely used for solving real-life problems. However, using only one type of unstructured element consumes a lot of computational time because the number of elements cannot be controlled. Hence, a hybrid grid that contains mixed elements, such as the use of hexahedral elements along with tetrahedral and pyramidal elements, gives the user control over the number of elements in the domain, and thus only the domain that requires a finer grid is meshed finer and not the entire domain. This work aims to develop such a finite-volume hybrid grid solver capable of handling turbulence flows and conjugate heat transfer. It has been extended to solving flow involving separation and subsequent reattachment occurring due to sudden expansion or contraction. A significant effect of mixing high- and low-enthalpy fluid occurs in the reattached regions of these devices. This makes the study of the backward-facing and forward-facing step with heat transfer an important field of research. The problem of the forward-facing step with conjugate heat transfer was taken up and solved for turbulence flow using a two-equation model of k-{omega}. The variation in the flow profile and heat transfer behavior has been studied with the variation in Re and solid to fluid thermal conductivity ratios. The results for the variation in local Nusselt number, interface temperature and skin friction factor are presented.

  13. Hybrid mesh finite volume CFD code for studying heat transfer in a forward-facing step

    Science.gov (United States)

    Jayakumar, J. S.; Kumar, Inder; Eswaran, V.

    2010-12-01

    Computational fluid dynamics (CFD) methods employ two types of grid: structured and unstructured. Developing the solver and data structures for a finite-volume solver is easier than for unstructured grids. But real-life problems are too complicated to be fitted flexibly by structured grids. Therefore, unstructured grids are widely used for solving real-life problems. However, using only one type of unstructured element consumes a lot of computational time because the number of elements cannot be controlled. Hence, a hybrid grid that contains mixed elements, such as the use of hexahedral elements along with tetrahedral and pyramidal elements, gives the user control over the number of elements in the domain, and thus only the domain that requires a finer grid is meshed finer and not the entire domain. This work aims to develop such a finite-volume hybrid grid solver capable of handling turbulence flows and conjugate heat transfer. It has been extended to solving flow involving separation and subsequent reattachment occurring due to sudden expansion or contraction. A significant effect of mixing high- and low-enthalpy fluid occurs in the reattached regions of these devices. This makes the study of the backward-facing and forward-facing step with heat transfer an important field of research. The problem of the forward-facing step with conjugate heat transfer was taken up and solved for turbulence flow using a two-equation model of k-ω. The variation in the flow profile and heat transfer behavior has been studied with the variation in Re and solid to fluid thermal conductivity ratios. The results for the variation in local Nusselt number, interface temperature and skin friction factor are presented.

  14. Hybrid mesh finite volume CFD code for studying heat transfer in a forward-facing step

    International Nuclear Information System (INIS)

    Jayakumar, J S; Kumar, Inder; Eswaran, V

    2010-01-01

    Computational fluid dynamics (CFD) methods employ two types of grid: structured and unstructured. Developing the solver and data structures for a finite-volume solver is easier than for unstructured grids. But real-life problems are too complicated to be fitted flexibly by structured grids. Therefore, unstructured grids are widely used for solving real-life problems. However, using only one type of unstructured element consumes a lot of computational time because the number of elements cannot be controlled. Hence, a hybrid grid that contains mixed elements, such as the use of hexahedral elements along with tetrahedral and pyramidal elements, gives the user control over the number of elements in the domain, and thus only the domain that requires a finer grid is meshed finer and not the entire domain. This work aims to develop such a finite-volume hybrid grid solver capable of handling turbulence flows and conjugate heat transfer. It has been extended to solving flow involving separation and subsequent reattachment occurring due to sudden expansion or contraction. A significant effect of mixing high- and low-enthalpy fluid occurs in the reattached regions of these devices. This makes the study of the backward-facing and forward-facing step with heat transfer an important field of research. The problem of the forward-facing step with conjugate heat transfer was taken up and solved for turbulence flow using a two-equation model of k-ω. The variation in the flow profile and heat transfer behavior has been studied with the variation in Re and solid to fluid thermal conductivity ratios. The results for the variation in local Nusselt number, interface temperature and skin friction factor are presented.

  15. Development and assessment of the COBRA/RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jae Jun; Ha, Kwi Seok; Sim, Seok Ku

    1997-04-01

    The COBRA/RELAP5 code, a merged version of the COBRA-TF and RELAP5/MOD3.2 codes, has been developed to combine the realistic three-dimensional reactor vessel model of COBRA-TF with RELAP5/MOD3, thus to produce an advanced system analysis code with a multidimensional thermal-hydraulic module. This report provides the integration scheme of the two codes and the results of developmental assessments. These includes single channel tests, manometric flow oscillation problem, THTF Test 105, and LOFT L2-3 large-break loss-of-coolant experiment. From the single channel tests the integration scheme and its implementation were proven to be valid. Other simulation results showed good agreement with the experiments. The computational speed was also satisfactory. So it is confirmed that COBRA/RELAP5 can be a promising tool for analysis of complicated, multidimensional two-phase flow transients. The area of further improvements in the code integration are also identified. This report also serves as a user`s manual for the COBRA/RELAP5 code. (author). 6 tabs., 20 figs., 20 refs.

  16. On transform coding tools under development for VP10

    Science.gov (United States)

    Parker, Sarah; Chen, Yue; Han, Jingning; Liu, Zoe; Mukherjee, Debargha; Su, Hui; Wang, Yongzhe; Bankoski, Jim; Li, Shunyao

    2016-09-01

    Google started the WebM Project in 2010 to develop open source, royaltyfree video codecs designed specifically for media on the Web. The second generation codec released by the WebM project, VP9, is currently served by YouTube, and enjoys billions of views per day. Realizing the need for even greater compression efficiency to cope with the growing demand for video on the web, the WebM team embarked on an ambitious project to develop a next edition codec, VP10, that achieves at least a generational improvement in coding efficiency over VP9. Starting from VP9, a set of new experimental coding tools have already been added to VP10 to achieve decent coding gains. Subsequently, Google joined a consortium of major tech companies called the Alliance for Open Media to jointly develop a new codec AV1. As a result, the VP10 effort is largely expected to merge with AV1. In this paper, we focus primarily on new tools in VP10 that improve coding of the prediction residue using transform coding techniques. Specifically, we describe tools that increase the flexibility of available transforms, allowing the codec to handle a more diverse range or residue structures. Results are presented on a standard test set.

  17. ICARE/CATHARE and ASTEC code development trends

    International Nuclear Information System (INIS)

    Chatelard, P.; Dorsselaere, J.-P. van

    2000-01-01

    Regarding the computer code development for simulation of LWR severe accidents, IPSN developed a two-tier approach based on detailed codes such as ICARE/CATHARE and simplified models to be assembled in the ASTEC integral code. The ICARE/CATHARE code results from the coupling between the ICARE2 code modelling the core degradation phenomena and the thermalhydraulics code CATHARE2. It allows to calculate PWR and VVER severe accident sequences in the whole RCS. The modelling of the early degradation phase can be considered as rather complete in the ICARE/CATHARE V1 mod1 version (to be released by mid-2000) whereas some models are still missing for the late phase. The main future developments (ICARE/CATHARE V2) will concern the multi-dimensional thermalhydraulics, the quenching of partially damaged cores (mechanical and chemical effects), the debris bed two-phase thermalhydraulics (including reflooding) and the corium behaviour in the lower head. The main other physical improvements should concern the behaviour of boron carbide control rods, the processes governing the core loss of geometry (transition phase) and the oxidation of relocated melts. The ASTEC (Accident Source Term Evaluation Code) integral code, commonly developed by IPSN and GRS, aims to predict an entire LWR (PWR, VVER and BWR) severe accident sequence from the initiating event through to FP release out of the containment, for source term, PSA level 2, or accident management studies. The version ASTEC VO.3 to be released by mid-2000 can be considered now as robust and fast-running enough (between 2 and 12 hours for a one day accident) and allows to perform, with a containment multi-compartment configuration, any scenario accident study accounting for the main safety systems and operator procedures (spray, recombiner, etc.). The next version ASTEC V1, to be released beginning of 2002, will include the frontend simulation and improve modelling of in-vessel core degradation. A large validation activity will

  18. Development of particle and heavy ion transport code system

    International Nuclear Information System (INIS)

    Niita, Koji

    2004-01-01

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

  19. Recent developments in seismic analysis in the code Aster

    International Nuclear Information System (INIS)

    Guihot, P.; Devesa, G.; Dumond, A.; Panet, M.; Waeckel, F.

    1996-01-01

    Progress in the field of seismic qualification and design methods made these last few years allows physical phenomena actually in play to be better considered, while cutting down the conservatism associated with some simplified design methods. So following the change in methods and developing the most advantageous ones among them contributes to the process of the seismic margins assessment and the preparation of new design tools for future series. In this paper, the main developments and improvements in methods which have been made these last two years in the Code Aster, in order to improve seismic calculation methods and seismic margin assessment are presented. The first development relates to making the MISS3D soil structure interaction code available, thanks to an interface made with the Code Aster. The second relates to the possibility of making modal basis time calculations on multi-supported structures by considering local non linearities like impact, friction or squeeze fluid forces. Recent developments in random dynamics and postprocessing devoted to earthquake designs are then mentioned. Three applications of these developments are then ut forward. The first application relates to a test case for soil structure interaction design using MISS3D-Aster coupling. The second is a test case for a multi-supported structure. The last application, more for manufacturing, refers to seismic qualification of Main Live Steam stop valves. First results of the independent validation of the Code Aster seismic design functionalities, which provide and improve the quality of software, are also recalled. (authors)

  20. Development of high performance hybrid rocket fuels

    Science.gov (United States)

    Zaseck, Christopher R.

    In this document I discuss paraffin fuel combustion and investigate the effects of additives on paraffin entrainment and regression. In general, hybrid rockets offer an economical and safe alternative to standard liquid and solid rockets. However, slow polymeric fuel regression and low combustion efficiency have limited the commercial use of hybrid rockets. Paraffin is a fast burning fuel that has received significant attention in the 2000's and 2010's as a replacement for standard fuels. Paraffin regresses three to four times faster than polymeric fuels due to the entrainment of a surface melt layer. However, further regression rate enhancement over the base paraffin fuel is necessary for widespread hybrid rocket adoption. I use a small scale opposed flow burner to investigate the effect of additives on the combustion of paraffin. Standard additives such as aluminum combust above the flame zone where sufficient oxidizer levels are present. As a result no heat is generated below the flame itself. In small scale opposed burner experiments the effect of limited heat feedback is apparent. Aluminum in particular does not improve the regression of paraffin in the opposed burner. The lack of heat feedback from additive combustion limits the applicability of the opposed burner. In turn, the results obtained in the opposed burner with metal additive loaded hybrid fuels do not match results from hybrid rocket experiments. In addition, nano-scale aluminum increases melt layer viscosity and greatly slows the regression of paraffin in the opposed flow burner. However, the reactive additives improve the regression rate of paraffin in the opposed burner where standard metals do not. At 5 wt.% mechanically activated titanium and carbon (Ti-C) improves the regression rate of paraffin by 47% in the opposed burner. The mechanically activated Ti C likely reacts in or near the melt layer and provides heat feedback below the flame region that results in faster opposed burner regression

  1. Methods for the development of large computer codes under LTSS

    International Nuclear Information System (INIS)

    Sicilian, J.M.

    1977-06-01

    TRAC is a large computer code being developed by Group Q-6 for the analysis of the transient thermal hydraulic behavior of light-water nuclear reactors. A system designed to assist the development of TRAC is described. The system consists of a central HYDRA dataset, R6LIB, containing files used in the development of TRAC, and a file maintenance program, HORSE, which facilitates the use of this dataset

  2. Development of a domestically-made system code

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

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

  3. Development Of A Navier-Stokes Computer Code

    Science.gov (United States)

    Yoon, Seokkwan; Kwak, Dochan

    1993-01-01

    Report discusses aspects of development of CENS3D computer code, solving three-dimensional Navier-Stokes equations of compressible, viscous, unsteady flow. Implements implicit finite-difference or finite-volume numerical-integration scheme, called "lower-upper symmetric-Gauss-Seidel" (LU-SGS), offering potential for very low computer time per iteration and for fast convergence.

  4. Application of software engineering to development of reactor safety codes

    International Nuclear Information System (INIS)

    Wilburn, N.P.; Niccoli, L.G.

    1981-01-01

    Software Engineering, which is a systematic methodology by which a large scale software development project is partitioned into manageable pieces, has been applied to the development of LMFBR safety codes. The techniques have been applied extensively in the business and aerospace communities and have provided an answer to the drastically increasing cost of developing and maintaining software. The five phases of software engineering (Survey, Analysis, Design, Implementation, and Testing) were applied in turn to development of these codes, along with Walkthroughs (peer review) at each stage. The application of these techniques has resulted in SUPERIOR SOFTWARE which is well documented, thoroughly tested, easy to modify, easier to use and maintain. The development projects have resulted in lower overall cost. (orig.) [de

  5. Development of a code of practice for deep geothermal wells

    International Nuclear Information System (INIS)

    Leaver, J.D.; Bolton, R.S.; Dench, N.D.; Fooks, L.

    1990-01-01

    Recent and on-going changes to the structure of the New Zealand geothermal industry has shifted responsibility for the development of geothermal resources from central government to private enterprise. The need for a code of practice for deep geothermal wells was identified by the Geothermal Inspectorate of the Ministry of Commerce to maintain adequate standards of health and safety and to assist with industry deregulation. This paper reports that the Code contains details of methods, procedures, formulae and design data necessary to attain those standards, and includes information which drilling engineers having experience only in the oil industry could not be expected to be familiar with

  6. Theoretical atomic physics code development I: CATS: Cowan Atomic Structure Code

    International Nuclear Information System (INIS)

    Abdallah, J. Jr.; Clark, R.E.H.; Cowan, R.D.

    1988-12-01

    An adaptation of R.D. Cowan's Atomic Structure program, CATS, has been developed as part of the Theoretical Atomic Physics (TAPS) code development effort at Los Alamos. CATS has been designed to be easy to run and to produce data files that can interface with other programs easily. The CATS produced data files currently include wave functions, energy levels, oscillator strengths, plane-wave-Born electron-ion collision strengths, photoionization cross sections, and a variety of other quantities. This paper describes the use of CATS. 10 refs

  7. BBU code development for high-power microwave generators

    International Nuclear Information System (INIS)

    Houck, T.L.; Westenskow, G.A.; Yu, S.S.

    1992-01-01

    We are developing a two-dimensional, time-dependent computer code for the simulation of transverse instabilities in support of relativistic klystron-two beam accelerator research at LLNL. The code addresses transient effects as well as both cumulative and regenerative beam breakup modes. Although designed specifically for the transport of high current (kA) beams through traveling-wave structures, it is applicable to devices consisting of multiple combinations of standing-wave, traveling-wave, and induction accelerator structures. In this paper we compare code simulations to analytical solutions for the case where there is no rf coupling between cavities, to theoretical scaling parameters for coupled cavity structures, and to experimental data involving beam breakup in the two traveling-wave output structure of our microwave generator. (Author) 4 figs., tab., 5 refs

  8. Development of dynamic simulation code for fuel cycle fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aoki, Isao; Seki, Yasushi [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki (Japan); Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan

    1999-02-01

    A dynamic simulation code for fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during 2 days pulse operation cycles. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the fuel burn and the function of exhaust, purification, and supply. The processing constants of subsystem for steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using this code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  12. Development of simplified decommissioning cost estimation code for nuclear facilities

    International Nuclear Information System (INIS)

    Tachibana, Mitsuo; Shiraishi, Kunio; Ishigami, Tsutomu

    2010-01-01

    The simplified decommissioning cost estimation code for nuclear facilities (DECOST code) was developed in consideration of features and structures of nuclear facilities and similarity of dismantling methods. The DECOST code could calculate 8 evaluation items of decommissioning cost. Actual dismantling in the Japan Atomic Energy Agency (JAEA) was evaluated; unit conversion factors used to calculate the manpower of dismantling activities were evaluated. Consequently, unit conversion factors of general components could be classified into three kinds. Weights of components and structures of the facility were necessary for calculation of manpower. Methods for evaluating weights of components and structures of the facility were studied. Consequently, the weight of components in the facility was proportional to the weight of structures of the facility. The weight of structures of the facility was proportional to the total area of floors in the facility. Decommissioning costs of 7 nuclear facilities in the JAEA were calculated by using the DECOST code. To verify the calculated results, the calculated manpower was compared with the manpower gained from actual dismantling. Consequently, the calculated manpower and actual manpower were almost equal. The outline of the DECOST code, evaluation results of unit conversion factors, the evaluation method of the weights of components and structures of the facility are described in this report. (author)

  13. Development of a subchannel analysis code MATRA (Ver. α)

    International Nuclear Information System (INIS)

    Yoo, Y. J.; Hwang, D. H.

    1998-04-01

    A subchannel analysis code MATRA-α, an interim version of MATRA, has been developed to be run on an IBM PC or HP WS based on the existing CDC CYBER mainframe version of COBRA-IV-I. This MATRA code is a thermal-hydraulic analysis code based on the subchannel approach for calculating the enthalpy and flow distribution in fuel assemblies and reactor cores for both steady-state and transient conditions. MATRA-α has been provided with an improved structure, various functions, and models to give the more convenient user environment and to increase the code accuracy, various functions, and models to give the more convenient user environment and to increase the code accuracy. Among them, the pressure drop model has been improved to be applied to non-square-lattice rod arrays, and the lateral transport models between adjacent subchannels have been improved to increase the accuracy in predicting two-phase flow phenomena. Also included in this report are the detailed instructions for input data preparation and for auxiliary pre-processors to serve as a guide to those who want to use MATRA-α. In addition, we compared the predictions of MATRA-α with the experimental data on the flow and enthalpy distribution in three sample rod-bundle cases to evaluate the performance of MATRA-α. All the results revealed that the prediction of MATRA-α were better than those of COBRA-IV-I. (author). 16 refs., 1 tab., 13 figs

  14. 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)

  15. Development of parallel Fokker-Planck code ALLAp

    International Nuclear Information System (INIS)

    Batishcheva, A.A.; Sigmar, D.J.; Koniges, A.E.

    1996-01-01

    We report on our ongoing development of the 3D Fokker-Planck code ALLA for a highly collisional scrape-off-layer (SOL) plasma. A SOL with strong gradients of density and temperature in the spatial dimension is modeled. Our method is based on a 3-D adaptive grid (in space, magnitude of the velocity, and cosine of the pitch angle) and a second order conservative scheme. Note that the grid size is typically 100 x 257 x 65 nodes. It was shown in our previous work that only these capabilities make it possible to benchmark a 3D code against a spatially-dependent self-similar solution of a kinetic equation with the Landau collision term. In the present work we show results of a more precise benchmarking against the exact solutions of the kinetic equation using a new parallel code ALLAp with an improved method of parallelization and a modified boundary condition at the plasma edge. We also report first results from the code parallelization using Message Passing Interface for a Massively Parallel CRI T3D platform. We evaluate the ALLAp code performance versus the number of T3D processors used and compare its efficiency against a Work/Data Sharing parallelization scheme and a workstation version

  16. Development of fast and accurate Monte Carlo code MVP

    International Nuclear Information System (INIS)

    Mori, Takamasa

    2001-01-01

    The development work of fast and accurate Monte Carlo code MVP has started at JAERI in late 80s. From the beginning, the code was designed to utilize vector supercomputers and achieved higher computation speed by a factor of 10 or more compared with conventional codes. In 1994, the first version of MVP was released together with cross section libraries based on JENDL-3.1 and JENDL-3.2. In 1996, minor revision was made by adding several functions such as treatments of ENDF-B6 file 6 data, time dependent problem, and so on. Since 1996, several works have been carried out for the next version of MVP. The main works are (1) the development of continuous energy Monte Carlo burn-up calculation code MVP-BURN, (2) the development of a system to generate cross section libraries at arbitrary temperature, and (3) the study on error estimations and their biases in Monte Carlo eigenvalue calculations. This paper summarizes the main features of MVP, results of recent studies and future plans for MVP. (author)

  17. The history and development of nonlinear stellar pulsation codes

    International Nuclear Information System (INIS)

    Davis, C.G.

    1987-01-01

    This review is limited to the history and development of nonlinear stellar pulsation codes and methods. The narrative includes examples of practical interest in the application of these numerical methods to problems in stellar pulsation such as Cepheid mass discrepancy, the delineation of the RR Lyrae instability strip, and the question of the development of double-mode pulsation as observed in Cepheids, RR Lyrae and other variable stars. 15 refs

  18. Development of the Multi-Phase/Multi-Dimensional Code BUBBLEX

    International Nuclear Information System (INIS)

    Lee, Sang Yong; Kim, Shin Whan; Kim, Eun Kee

    2005-01-01

    A test version of the two-fluid program has been developed by extending the PISO algorithm. Unlike the conventional industry two-fluid codes, such as, RELAP5 and TRAC, this scheme does not need to develop a pressure matrix. Instead, it adopts the iterative procedure to implement the implicitness of the pressure. In this paper, a brief introduction to the numerical scheme will be presented. Then, its application to bubble column simulation will be described. Some concluding remarks will be followed

  19. Application of software to development of reactor-safety codes

    International Nuclear Information System (INIS)

    Wilburn, N.P.; Niccoli, L.G.

    1980-09-01

    Over the past two-and-a-half decades, the application of new techniques has reduced hardware cost for digital computer systems and increased computational speed by several orders of magnitude. A corresponding cost reduction in business and scientific software development has not occurred. The same situation is seen for software developed to model the thermohydraulic behavior of nuclear systems under hypothetical accident situations. For all cases this is particularly noted when costs over the total software life cycle are considered. A solution to this dilemma for reactor safety code systems has been demonstrated by applying the software engineering techniques which have been developed over the course of the last few years in the aerospace and business communities. These techniques have been applied recently with a great deal of success in four major projects at the Hanford Engineering Development Laboratory (HEDL): 1) a rewrite of a major safety code (MELT); 2) development of a new code system (CONACS) for description of the response of LMFBR containment to hypothetical accidents, and 3) development of two new modules for reactor safety analysis

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

    Science.gov (United States)

    Savage, B.; Snoke, J. A.

    2017-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-03-15

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

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

    International Nuclear Information System (INIS)

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

    2007-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-05-15

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

  4. Simulation of lower hybrid current drive in enhanced reversed shear plasmas in the tokamak fusion test reactor using the lower hybrid simulation code

    International Nuclear Information System (INIS)

    Kaita, R.; Bernabei, S.; Budny, R.

    1996-01-01

    The Enhanced Reversed Shear (ERS) mode has already shown great potential for improving the performance of the Tokamak Fusion Test Reactor (TFTR) and other devices. Sustaining the ERS, however, remains an outstanding problem. Lower hybrid (LH) current drive is a possible method for modifying the current profile and controlling its time evolution. To predict its effectiveness in TFTR, the Lower Hybrid Simulation Code (LSC) model is used in the TRANSP code and the Tokamak Simulation Code (TSC). Among the results from the simulations are the following. (1) Single-pass absorption is expected in TFTR ERS plasmas. The simulations show that the LH current follows isotherms of the electron temperature. The ability to control the location of the minimum in the q profile (q min ) has been demonstrated by varying the phase velocity of the launched LH waves and observing the change in the damping location. (2) LH current drive can been used to sustain the q min location. The tendency of qmin to drift inward, as the inductive current diffuses during the formation phase of the reversed shear discharge, is prevented by the LH current driven at a fixed radial location. If this results in an expanded plasma volume with improved confinement as high power neutral beam injection is applied, the high bootstrap currents induced during this phase can then maintain the larger qmin radius. (3) There should be no LH wave damping on energetic beam particles. The values of perpendicular index of refraction in the calculations never exceed about 20, while ions at TFR injection energies are resonant with waves having values closer to 100. Other issues being addressed in the study include the LH current drive efficiency in the presence of high bootstrap currents, and the effect of fast electron diffusion on LH current localization

  5. Village power hybrid systems development in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Flowers, L.; Green, J. [National Renewable Energy Lab., Golden, CO (United States); Bergey, M. [Bergey Windpower Co., Norman, OK (United States); Lilley, A. [Westinghouse Electric Corp., Pittsburgh, PA (United States); Mott, L. [Northern Power Systems, Moretown, VT (United States)

    1994-11-01

    The energy demand in developing countries is growing at a rate seven times that of the OECD countries, even though there are still 2 billion people living in developing countries without electricity. Many developing countries have social and economic development programs aimed at stemming the massive migration from the rural communities to the overcrowded, environmentally problematic, unemployment-bound urban centers. To address the issue of providing social, educational, health, and economic benefits to the rural communities of the developing world, a number of government and nongovernment agencies are sponsoring pilot programs to install and evaluate renewable energy systems as alternatives to line extension, diesels, kerosene, and batteries. The use of renewables in remote villages has yielded mixed results over the last 20 years. However, recently, photovoltaics, small wind turbines, and microhydro system shave gained increasing recognition as reliable, cost-effective alternatives to grid extension and diesel gensets for village-electricity applications. At the same time, hybrid systems based on combinations of PV/wind/batteries/diesel gensets have proven reliable and economic for remote international telecommunications markets. With the growing emphasis on environmentally and economically sustainable development of international rural communities, the US hybrid industry is responding with the development and demonstration of hybrid systems and architectures that will directly compete with conventional alternatives for village electrification. Assisting the US industry in this development, the National Renewable Energy Laboratory (NREL) has embarked on a program of collaborative technology development and technical assistance in the area of hybrid systems for village power. Following a brief review of village-power hybrid systems application and design issues, this paper presents the present industry development activities of three US suppliers and the NREL.

  6. Python-Assisted MODFLOW Application and Code Development

    Science.gov (United States)

    Langevin, C.

    2013-12-01

    The U.S. Geological Survey (USGS) has a long history of developing and maintaining free, open-source software for hydrological investigations. The MODFLOW program is one of the most popular hydrologic simulation programs released by the USGS, and it is considered to be the most widely used groundwater flow simulation code. MODFLOW was written using a modular design and a procedural FORTRAN style, which resulted in code that could be understood, modified, and enhanced by many hydrologists. The code is fast, and because it uses standard FORTRAN it can be run on most operating systems. Most MODFLOW users rely on proprietary graphical user interfaces for constructing models and viewing model results. Some recent efforts, however, have focused on construction of MODFLOW models using open-source Python scripts. Customizable Python packages, such as FloPy (https://code.google.com/p/flopy), can be used to generate input files, read simulation results, and visualize results in two and three dimensions. Automating this sequence of steps leads to models that can be reproduced directly from original data and rediscretized in space and time. Python is also being used in the development and testing of new MODFLOW functionality. New packages and numerical formulations can be quickly prototyped and tested first with Python programs before implementation in MODFLOW. This is made possible by the flexible object-oriented design capabilities available in Python, the ability to call FORTRAN code from Python, and the ease with which linear systems of equations can be solved using SciPy, for example. Once new features are added to MODFLOW, Python can then be used to automate comprehensive regression testing and ensure reliability and accuracy of new versions prior to release.

  7. Development and verifications of fast reactor fuel design code ''Ceptar''

    International Nuclear Information System (INIS)

    Ozawa, T.; Nakazawa, H.; Abe, T.

    2001-01-01

    The annular fuel is very beneficial for fast reactors, because it is available for both high power and high burn-up. Concerning the irradiation behavior of the annular fuel, most of annular pellets irradiated up to high burn-up showed shrinkage of the central hole due to deformation and restructuring of the pellets. It is needed to predict precisely the shrinkage of the central hole during irradiation, because it has a great influence on power-to-melt. In this paper, outline of CEPTAR code (Calculation code to Evaluate fuel pin stability for annular fuel design) developed to meet this need is presented. In this code, the radial profile of fuel density can be computed by using the void migration model, and law of conservation of mass defines the inner diameter. For the mechanical analysis, the fuel and cladding deformation caused by the thermal expansion, swelling and creep is computed by the stress-strain analysis using the approximation of plane-strain. In addition, CEPTAR can also take into account the effect of Joint-Oxide-Gain (JOG) which is observed in fuel-cladding gap of high burn-up fuel. JOG has an effect to decrease the fuel swelling and to improve the gap conductance due to deposition of solid fission product. Based on post-irradiation data on PFR annular fuel, we developed an empirical model for JOG. For code verifications, the thermal and mechanical data obtained from various irradiation tests and post-irradiation examinations were compared with the predictions of this code. In this study, INTA (instrumented test assembly) test in JOYO, PTM (power-to-melt) test in JOYO, EBR-II, FFTF and MTR in Harwell laboratory, and post-irradiation examinations on a number of PFR fuels, were used as verification data. (author)

  8. Status of development and verification of the CTFD code FLUBOX

    International Nuclear Information System (INIS)

    Graf, U.; Paradimitriou, P.

    2004-01-01

    The Computational Two-Fluid Dynamics (CTFD) code FLUBOX is developed at GRS for the multidimensional simulation of two-phase flows. FLUBOX will also be used as a multidimensional module for the German system code ATHLET. The Benchmark test cases of the European ASTAR project were used to verify the ability of the code FLUBOX to calculate typical two-phase flow phenomena and conditions: void and pressure wave propagation, phase transitions, countercurrent flows, sharp interface movements, compressible (vapour) and nearly incompressible (water) conditions, thermal and mechanical non-equilibrium, stiff source terms due to mass and heat transfer between the phases. Realistic simulations of two-phase require beside the pure conservation equations additional transport equations for the interfacial area, turbulent energy and dissipation. A transport equation for the interfacial area density covering the whole two-phase flow range is in development. First validation calculations are presented in the paper. Turbulent shear stress for two-phase flows will be modelled by the development of transport equations for the turbulent kinetic energy and the turbulent dissipation rate. The development of the transport equations is mainly based on first principles on bubbles or drops and is largely free from empiricism. (author)

  9. Development of the criticality accident analysis code, AGNES

    International Nuclear Information System (INIS)

    Nakajima, Ken

    1989-01-01

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

  10. SCDAP/RELAP5/MOD3 code development and assessment

    International Nuclear Information System (INIS)

    Allison, C.M.; Heath, C.H.; Siefken, L.J.; Hohorst, J.K.

    1991-01-01

    The SCDAP/RELAP5/MOD3 computer code is designed to describe the overall reactor coolant system (RCS) thermal-hydraulic response, core damage progression, and fission product release and transport during severe accidents. The code is being developed at the Idaho National Engineering Laboratory (INEL) under the primary sponsorship of the Office of Nuclear Regulatory Research of the Nuclear Regulatory Commission (NRC). SCDAP/RELAP5/MOD3, created in January, 1991, is the result of merging RELAP5/MOD3 with SCDAP and TRAP-MELT models from SCDAP/RELAP5/MOD2.5. The RELAP5 models calculate the overall RCS thermal-hydraulics, control system interactions, reactor kinetics, and the transport of noncondensible gases, fission products, and aerosols. The SCDAP models calculate the damage progression in the core structures, the formation, heatup, and melting of debris, and the creep rupture failure of the lower head and other RCS structures. The TRAP-MELT models calculate the deposition of fission products upon aerosols or structural surfaces; the formation, growth, or deposition of aerosols; and the evaporation of species from surfaces. The systematic assessment of modeling uncertainties in SCDAP/RELAP5 code is currently underway. This assessment includes (a) the evaluation of code-to-data comparisons using stand-alone SCDAP and SCDAP/RELAP5/MOD3, (b) the estimation of modeling and experimental uncertainties, and (c) the determination of the influence of those uncertainties on predicted severe accident behavior

  11. Recent Developments in the Code RITRACKS (Relativistic Ion Tracks)

    Science.gov (United States)

    Plante, Ianik; Ponomarev, Artem L.; Blattnig, Steve R.

    2018-01-01

    The code RITRACKS (Relativistic Ion Tracks) was developed to simulate detailed stochastic radiation track structures of ions of different types and energies. Many new capabilities were added to the code during the recent years. Several options were added to specify the times at which the tracks appear in the irradiated volume, allowing the simulation of dose-rate effects. The code has been used to simulate energy deposition in several targets: spherical, ellipsoidal and cylindrical. More recently, density changes as well as a spherical shell were implemented for spherical targets, in order to simulate energy deposition in walled tissue equivalent proportional counters. RITRACKS is used as a part of the new program BDSTracks (Biological Damage by Stochastic Tracks) to simulate several types of chromosome aberrations in various irradiation conditions. The simulation of damage to various DNA structures (linear and chromatin fiber) by direct and indirect effects has been improved and is ongoing. Many improvements were also made to the graphic user interface (GUI), including the addition of several labels allowing changes of units. A new GUI has been added to display the electron ejection vectors. The parallel calculation capabilities, notably the pre- and post-simulation processing on Windows and Linux machines have been reviewed to make them more portable between different systems. The calculation part is currently maintained in an Atlassian Stash® repository for code tracking and possibly future collaboration.

  12. Development of the biosphere code BIOMOD: final report

    International Nuclear Information System (INIS)

    Kane, P.

    1983-05-01

    Final report to DoE on the development of the biosphere code BIOMOD. The work carried out under the contract is itemised. Reference is made to the six documents issued along with the final report. These consist of two technical notes issued as interim consultative documents, a user's guide and a programmer's guide to BIOMOD, a database description, program test document and a technical note entitled ''BIOMOD - preliminary findings''. (author)

  13. Scientific codes developed and used at GRS. Nuclear simulation chain

    Energy Technology Data Exchange (ETDEWEB)

    Schaffrath, Andreas; Sonnenkalb, Martin; Sievers, Juergen; Luther, Wolfgang; Velkov, Kiril [Gesellschaft fuer Anlagen und Reaktorsicherheit (GRS) gGmbH, Garching/Muenchen (Germany). Forschungszentrum

    2016-05-15

    Over 60 technical experts of the reactor safety research division of the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH are developing and validating reliable methods and computer codes - summarized under the term nuclear simulation chain - for the safety-related assessment for all types of nuclear power plants (NPP) and other nuclear facilities considering the current state of science and technology. This nuclear simulation chain has to be able to simulate and assess all relevant physical processes and phenomena for all operating states and (severe) accidents. In the present contribution, the nuclear simulation chain developed and applied by GRS as well as selected examples of its application are presented. The latter demonstrate impressively the width of its scope and its performance. The GRS codes can be passed on request to other (national as well as international) organizations. This contributes to a worldwide increase of the nuclear safety standards. The code transfer is especially important for developing and emerging countries lacking the financial means and/or the necessary know-how for this purpose. At the end of this contribution, the respective course of action is described.

  14. Sex-specific asymmetry in eye development in interspecific hybrids ...

    Indian Academy of Sciences (India)

    Home; Journals; Journal of Genetics; Volume 94; Issue 3. Sex-specific asymmetry in eye development in interspecific hybrids in the Drosophila bipectinata species complex. Bashisth N. Singh Parul Banerjee. Research Note Volume 94 Issue 3 September 2015 pp 493-495 ...

  15. Development of bio-hybrid material based on Salmonella ...

    African Journals Online (AJOL)

    Teodoro

    2016-07-13

    Jul 13, 2016 ... Full Length Research Paper. Development of bio-hybrid material based on. Salmonella Typhimurium and layered double hydroxides. Slah Hidouri .... the LDH with co-precipitation synthesis method was successfully done according the study given by Hidouri et al. (2011), Abdelkader et al. (2011), Hidouri et ...

  16. Development of a hybrid system of artificial neural networks and ...

    African Journals Online (AJOL)

    Development of a hybrid system of artificial neural networks and artificial bee colony algorithm for prediction and modeling of customer choice in the market. ... attempted to present a new method for the modeling and prediction of customer choice in the market using the combination of artificial intelligence and data mining.

  17. Mobile, hybrid Compton/coded aperture imaging for detection, identification and localization of gamma-ray sources at stand-off distances

    Science.gov (United States)

    Tornga, Shawn R.

    The Stand-off Radiation Detection System (SORDS) program is an Advanced Technology Demonstration (ATD) project through the Department of Homeland Security's Domestic Nuclear Detection Office (DNDO) with the goal of detection, identification and localization of weak radiological sources in the presence of large dynamic backgrounds. The Raytheon-SORDS Tri-Modal Imager (TMI) is a mobile truck-based, hybrid gamma-ray imaging system able to quickly detect, identify and localize, radiation sources at standoff distances through improved sensitivity while minimizing the false alarm rate. Reconstruction of gamma-ray sources is performed using a combination of two imaging modalities; coded aperture and Compton scatter imaging. The TMI consists of 35 sodium iodide (NaI) crystals 5x5x2 in3 each, arranged in a random coded aperture mask array (CA), followed by 30 position sensitive NaI bars each 24x2.5x3 in3 called the detection array (DA). The CA array acts as both a coded aperture mask and scattering detector for Compton events. The large-area DA array acts as a collection detector for both Compton scattered events and coded aperture events. In this thesis, developed coded aperture, Compton and hybrid imaging algorithms will be described along with their performance. It will be shown that multiple imaging modalities can be fused to improve detection sensitivity over a broader energy range than either alone. Since the TMI is a moving system, peripheral data, such as a Global Positioning System (GPS) and Inertial Navigation System (INS) must also be incorporated. A method of adapting static imaging algorithms to a moving platform has been developed. Also, algorithms were developed in parallel with detector hardware, through the use of extensive simulations performed with the Geometry and Tracking Toolkit v4 (GEANT4). Simulations have been well validated against measured data. Results of image reconstruction algorithms at various speeds and distances will be presented as well as

  18. Development status of the lattice physics code in COSINE project

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y.; Yu, H.; Li, S.; Liu, Z.; Yan, Y. [State Nuclear Power Software Development Center, SNPTC, National Energy Key Laboratory of Nuclear Power Software NEKLS, North Third Ring Road, Beijing 100029 (China)

    2013-07-01

    LATC is an essential part of COSINE code package, which stands for Core and System Integrated Engine for design and analysis. LATC performs 2D multi-group assembly transport calculation and generates few group constants and the required cross-section data for CORE, the core simulator code. LATC is designed to have the capability of modeling the API 000 series assemblies. The development is a continuously improved process. Currently, LATC uses well-proven technology to achieve the key functions. In the next stage, more advanced methods and modules will be implemented. At present, WIMS and WIMS improved format library could be read in LATC code. For resonance calculation, equivalent relation with rational approximations is utilized. For transport calculation, two options are available. One choice is collision probability method in cell homogenization while discrete coordinate method in assembly homogenization, the other is method of characteristics in assembly homogenization directly. For depletion calculation, an improved linear rate 'constant power' depletion method has been developed. (authors)

  19. Development status of the lattice physics code in COSINE project

    International Nuclear Information System (INIS)

    Chen, Y.; Yu, H.; Li, S.; Liu, Z.; Yan, Y.

    2013-01-01

    LATC is an essential part of COSINE code package, which stands for Core and System Integrated Engine for design and analysis. LATC performs 2D multi-group assembly transport calculation and generates few group constants and the required cross-section data for CORE, the core simulator code. LATC is designed to have the capability of modeling the API 000 series assemblies. The development is a continuously improved process. Currently, LATC uses well-proven technology to achieve the key functions. In the next stage, more advanced methods and modules will be implemented. At present, WIMS and WIMS improved format library could be read in LATC code. For resonance calculation, equivalent relation with rational approximations is utilized. For transport calculation, two options are available. One choice is collision probability method in cell homogenization while discrete coordinate method in assembly homogenization, the other is method of characteristics in assembly homogenization directly. For depletion calculation, an improved linear rate 'constant power' depletion method has been developed. (authors)

  20. Development and preliminary validation of flux map processing code MAPLE

    International Nuclear Information System (INIS)

    Li Wenhuai; Zhang Xiangju; Dang Zhen; Chen Ming'an; Lu Haoliang; Li Jinggang; Wu Yuanbao

    2013-01-01

    The self-reliant flux map processing code MAPLE was developed by China General Nuclear Power Corporation (CGN). Weight coefficient method (WCM), polynomial expand method (PEM) and thin plane spline (TPS) method were applied to fit the deviation between measured and predicted detector signal results for two-dimensional radial plane, to interpolate or extrapolate the non-instrumented location deviation. Comparison of results in the test cases shows that the TPS method can better capture the information of curved fitting lines than the other methods. The measured flux map data of the Lingao Nuclear Power Plant were processed using MAPLE as validation test cases, combined with SMART code. Validation results show that the calculation results of MAPLE are reasonable and satisfied. (authors)

  1. Process development and tooling design for intrinsic hybrid composites

    Science.gov (United States)

    Riemer, M.; Müller, R.; Drossel, W. G.; Landgrebe, D.

    2017-09-01

    Hybrid parts, which combine the advantages of different material classes, are moving into the focus of lightweight applications. This development is amplified by their high potential for usage in the field of crash relevant structures. By the current state of the art, hybrid parts are mainly made in separate, subsequent forming and joining processes. By using the concept of an intrinsic hybrid, the shaping of the part and the joining of the different materials are performed in a single process step for shortening the overall processing time and thereby the manufacturing costs. The investigated hybrid part is made from continuous fibre reinforced plastic (FRP), in which a metallic reinforcement structure is integrated. The connection between these layered components is realized by a combination of adhesive bonding and a geometrical form fit. The form fit elements are intrinsically generated during the forming process. This contribution regards the development of the forming process and the design of the forming tool for the single step production of a hybrid part. To this end a forming tool, which combines the thermo-forming and the metal forming process, is developed. The main challenge by designing the tool is the temperature management of the tool elements for the variothermal forming process. The process parameters are determined in basic tests and finite element (FE) simulation studies. On the basis of these investigations a control concept for the steering of the motion axes and the tool temperature is developed. Forming tests are carried out with the developed tool and the manufactured parts are analysed by computer assisted tomography (CT) scans.

  2. The possibility of developing hybrid PV/T solar system

    Science.gov (United States)

    Dobrnjac, M.; Zivkovic, P.; Babic, V.

    2017-05-01

    An alternative and cost-effective solution to developing integrated PV system is to use hybrid photovoltaic/thermal (PV/T) solar system. The temperature of PV modules increases due to the absorbed solar radiation that is not converted into electricity, causing a decrease in their efficiency. In hybrid PV/T solar systems the reduction of PV module temperature can be combined with a useful fluid heating. In this paper we present the possibility of developing a new hybrid PV/T solar system. Hybrid PV/T system can provide electrical and thermal energy, thus achieving a higher energy conversion rate of the absorbed solar radiation. We developed PV/T prototype consisted of commercial PV module and thermal panel with our original solution of aluminium absorber with special geometric shapes. The main advantages of our combined PV/T system are: removing of heat from the PV panel; extending the lifetime of photovoltaic cells; excess of the removing heat from PV part is used to heat the fluid in the thermal part of the panel; the possibility of using on the roof and facade constructions because less weight.

  3. Development of Educational SharePoint portal for coding students

    OpenAIRE

    Colomer Castelló, Gerard

    2016-01-01

    The project will explain what is SharePoint, why is used and who does use it. It will also expose some alternatives, compare them to SharePoint and expose the pros and cons. The main objective of this project will be developing an educational SharePoint portal. Using this portal, coding students will be able to share their solutions to different exercises, vote the best solutions and comment them. This portal will be developed using only software and servers obtained legally without any cost....

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

    Science.gov (United States)

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

    2004-01-01

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

  5. Development of the SCHAMBETA code for scoping analysis of HCDA

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Soo Dong; Hahn, D. H

    2000-06-01

    A computer code, SCHAMBETA(Scoping Code for HCDA Analysis using Modified Bethe-Tait Method), is developed to investigate the core disassembly process following a meltdown accident in the framework of a mofified Bethe-Tait method as part of the scoping analysis work to demonstrate the inherent safety of conceptual designs of Korea Advanced Liquid Metal Reactor(KALIMER), A 150 Mwe pool-type sodium cooled prototype fast reactor that uses U-Pu-Zr metallic fuel. The methodologies adopted in the code ared particularly useful to perform various parametric studies for better understanding of core disassembly process of liquid metal fast reactors as well as to estimate upper-limit values of the energy release resulting from a power excursion. In the SCHAMBETA code, the core kinetics and hydraulic behavior of the KALIMER is followed over the period of the super-prompt critical power excursion induced by the ramp reactivity insertion, starting at the time that the sodium-voided core reaches the melting temperature of the metallic fuels. For this purpose, the equations of state of pressure-energy density relationship are derived for the saturated-vapor as well as the solid liquid of metallic uranium fuel, and implemenmted into the formulations of the disassembly reactivity. Mathematical formulations are then developed, in the framework of Modified Bethe-Tait method, in a form relevant to utilize the improved equations of state as well as to consider Doppler effects, for scoping analysis of the super-prompt-critical power excursions driven by a specified rate of reactivity insertion.

  6. Report on FY15 alloy 617 code rules development

    Energy Technology Data Exchange (ETDEWEB)

    Sham, Sam [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jetter, Robert I [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hollinger, Greg [Becht Engineering Co., Inc., Liberty Corner, NJ (United States); Pease, Derrick [Becht Engineering Co., Inc., Liberty Corner, NJ (United States); Carter, Peter [Stress Engineering Services, Inc., Houston, TX (United States); Pu, Chao [Univ. of Tennessee, Knoxville, TN (United States); Wang, Yanli [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    Due to its strength at very high temperatures, up to 950°C (1742°F), Alloy 617 is the reference construction material for structural components that operate at or near the outlet temperature of the very high temperature gas-cooled reactors. However, the current rules in the ASME Section III, Division 5 Subsection HB, Subpart B for the evaluation of strain limits and creep-fatigue damage using simplified methods based on elastic analysis have been deemed inappropriate for Alloy 617 at temperatures above 650°C (1200°F) (Corum and Brass, Proceedings of ASME 1991 Pressure Vessels and Piping Conference, PVP-Vol. 215, p.147, ASME, NY, 1991). The rationale for this exclusion is that at higher temperatures it is not feasible to decouple plasticity and creep, which is the basis for the current simplified rules. This temperature, 650°C (1200°F), is well below the temperature range of interest for this material for the high temperature gas-cooled reactors and the very high temperature gas-cooled reactors. The only current alternative is, thus, a full inelastic analysis requiring sophisticated material models that have not yet been formulated and verified. To address these issues, proposed code rules have been developed which are based on the use of elastic-perfectly plastic (EPP) analysis methods applicable to very high temperatures. The proposed rules for strain limits and creep-fatigue evaluation were initially documented in the technical literature (Carter, Jetter and Sham, Proceedings of ASME 2012 Pressure Vessels and Piping Conference, papers PVP 2012 28082 and PVP 2012 28083, ASME, NY, 2012), and have been recently revised to incorporate comments and simplify their application. Background documents have been developed for these two code cases to support the ASME Code committee approval process. These background documents for the EPP strain limits and creep-fatigue code cases are documented in this report.

  7. Development of GUI systems for the MIDAS code

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  8. AREVA Developments for an Efficient and Reliable use of Monte Carlo codes for Radiation Transport Applications

    Science.gov (United States)

    Chapoutier, Nicolas; Mollier, François; Nolin, Guillaume; Culioli, Matthieu; Mace, Jean-Reynald

    2017-09-01

    In the context of the rising of Monte Carlo transport calculations for any kind of application, AREVA recently improved its suite of engineering tools in order to produce efficient Monte Carlo workflow. Monte Carlo codes, such as MCNP or TRIPOLI, are recognized as reference codes to deal with a large range of radiation transport problems. However the inherent drawbacks of theses codes - laboring input file creation and long computation time - contrast with the maturity of the treatment of the physical phenomena. The goals of the recent AREVA developments were to reach similar efficiency as other mature engineering sciences such as finite elements analyses (e.g. structural or fluid dynamics). Among the main objectives, the creation of a graphical user interface offering CAD tools for geometry creation and other graphical features dedicated to the radiation field (source definition, tally definition) has been reached. The computations times are drastically reduced compared to few years ago thanks to the use of massive parallel runs, and above all, the implementation of hybrid variance reduction technics. From now engineering teams are capable to deliver much more prompt support to any nuclear projects dealing with reactors or fuel cycle facilities from conceptual phase to decommissioning.

  9. AREVA Developments for an Efficient and Reliable use of Monte Carlo codes for Radiation Transport Applications

    Directory of Open Access Journals (Sweden)

    Chapoutier Nicolas

    2017-01-01

    Full Text Available In the context of the rising of Monte Carlo transport calculations for any kind of application, AREVA recently improved its suite of engineering tools in order to produce efficient Monte Carlo workflow. Monte Carlo codes, such as MCNP or TRIPOLI, are recognized as reference codes to deal with a large range of radiation transport problems. However the inherent drawbacks of theses codes - laboring input file creation and long computation time - contrast with the maturity of the treatment of the physical phenomena. The goals of the recent AREVA developments were to reach similar efficiency as other mature engineering sciences such as finite elements analyses (e.g. structural or fluid dynamics. Among the main objectives, the creation of a graphical user interface offering CAD tools for geometry creation and other graphical features dedicated to the radiation field (source definition, tally definition has been reached. The computations times are drastically reduced compared to few years ago thanks to the use of massive parallel runs, and above all, the implementation of hybrid variance reduction technics. From now engineering teams are capable to deliver much more prompt support to any nuclear projects dealing with reactors or fuel cycle facilities from conceptual phase to decommissioning.

  10. The role of the PIRT process in identifying code improvements and executing code development

    International Nuclear Information System (INIS)

    Wilson, G.E.; Boyack, B.E.

    1997-01-01

    In September 1988, the USNRC issued a revised ECCS rule for light water reactors that allows, as an option, the use of best estimate (BE) plus uncertainty methods in safety analysis. The key feature of this licensing option relates to quantification of the uncertainty in the determination that an NPP has a low probability of violating the safety criteria specified in 10 CFR 50. To support the 1988 licensing revision, the USNRC and its contractors developed the CSAU evaluation methodology to demonstrate the feasibility of the BE plus uncertainty approach. The PIRT process, Step 3 in the CSAU methodology, was originally formulated to support the BE plus uncertainty licensing option as executed in the CSAU approach to safety analysis. Subsequent work has shown the PIRT process to be a much more powerful tool than conceived in its original form. Through further development and application, the PIRT process has shown itself to be a robust means to establish safety analysis computer code phenomenological requirements in their order of importance to such analyses. Used early in research directed toward these objectives, PIRT results also provide the technical basis and cost effective organization for new experimental programs needed to improve the safety analysis codes for new applications. The primary purpose of this paper is to describe the generic PIRT process, including typical and common illustrations from prior applications. The secondary objective is to provide guidance to future applications of the process to help them focus, in a graded approach, on systems, components, processes and phenomena that have been common in several prior applications

  11. The role of the PIRT process in identifying code improvements and executing code development

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, G.E. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Boyack, B.E. [Los Alamos National Lab., NM (United States)

    1997-07-01

    In September 1988, the USNRC issued a revised ECCS rule for light water reactors that allows, as an option, the use of best estimate (BE) plus uncertainty methods in safety analysis. The key feature of this licensing option relates to quantification of the uncertainty in the determination that an NPP has a {open_quotes}low{close_quotes} probability of violating the safety criteria specified in 10 CFR 50. To support the 1988 licensing revision, the USNRC and its contractors developed the CSAU evaluation methodology to demonstrate the feasibility of the BE plus uncertainty approach. The PIRT process, Step 3 in the CSAU methodology, was originally formulated to support the BE plus uncertainty licensing option as executed in the CSAU approach to safety analysis. Subsequent work has shown the PIRT process to be a much more powerful tool than conceived in its original form. Through further development and application, the PIRT process has shown itself to be a robust means to establish safety analysis computer code phenomenological requirements in their order of importance to such analyses. Used early in research directed toward these objectives, PIRT results also provide the technical basis and cost effective organization for new experimental programs needed to improve the safety analysis codes for new applications. The primary purpose of this paper is to describe the generic PIRT process, including typical and common illustrations from prior applications. The secondary objective is to provide guidance to future applications of the process to help them focus, in a graded approach, on systems, components, processes and phenomena that have been common in several prior applications.

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  13. Development of INCTAC code for analyzing criticality accident phenomena

    International Nuclear Information System (INIS)

    Mitake, Susumu; Hayashi, Yamato; Sakurai, Shungo

    2003-01-01

    Aiming at understanding nuclear transients and thermal- and hydraulic-phenomena of the criticality accident, a code named INCTAC has been newly developed at the Institute of Nuclear Safety. The code is applicable to the analysis of criticality accident transients of aqueous homogenous fuel solution system. Neutronic transient model is composed of equations for the kinetics and for the spatial distributions, which are deduced from the time dependent multi-group transport equations with the quasi steady state assumption. Thermal-hydraulic transient model is composed of a complete set of the mass, momentum and energy equations together with the two-phase flow assumptions. Validation tests of INCTAC were made using the data obtained at TRACY, a transient experiment criticality facility of JAERI. The calculated results with INCTAC showed a very good agreement with the experiment data, except a slight discrepancy of the time when the peak of reactor power was attained. But, the discrepancy was resolved with the use of an adequate model for movement and transfer of the void in the fuel solution mostly generated by radiolysis. With a simulation model for the transport of radioactive materials through ventilation systems to the environment, INCTAC will be used as an overall safety evaluation code of the criticality accident. (author)

  14. Hybrid stage I palliation for hypo-plastic left heart condition without a hybrid suite: Suggestions for developing nations

    Directory of Open Access Journals (Sweden)

    S. Anuradha

    2012-07-01

    Full Text Available Cardiac hybrid procedures are performed in modern, spacious, and highly equipped hybrid suites in developed countries. Organizing such expensive suites in countries with an emerging economy is difficult from both a financial and logistics point of view. We share our experience of safely performing a Hybrid stage I palliation procedure for Aortic atresia with ventricular septal defect on a 2-month-old infant weighing 3.35 kg using minimal resources in a conventional catheterization laboratory.

  15. Development of tools for automatic generation of PLC code

    CERN Document Server

    Koutli, Maria; Rochez, Jacques

    This Master thesis was performed at CERN and more specifically in the EN-ICE-PLC section. The Thesis describes the integration of two PLC platforms, that are based on CODESYS development tool, to the CERN defined industrial framework, UNICOS. CODESYS is a development tool for PLC programming, based on IEC 61131-3 standard, and is adopted by many PLC manufacturers. The two PLC development environments are, the SoMachine from Schneider and the TwinCAT from Beckhoff. The two CODESYS compatible PLCs, should be controlled by the SCADA system of Siemens, WinCC OA. The framework includes a library of Function Blocks (objects) for the PLC programs and a software for automatic generation of the PLC code based on this library, called UAB. The integration aimed to give a solution that is shared by both PLC platforms and was based on the PLCOpen XML scheme. The developed tools were demonstrated by creating a control application for both PLC environments and testing of the behavior of the code of the library.

  16. Development Of The Computer Code For Comparative Neutron Activation Analysis

    International Nuclear Information System (INIS)

    Purwadi, Mohammad Dhandhang

    2001-01-01

    The qualitative and quantitative chemical analysis with Neutron Activation Analysis (NAA) is an importance utilization of a nuclear research reactor, and this should be accelerated and promoted in application and its development to raise the utilization of the reactor. The application of Comparative NAA technique in GA Siwabessy Multi Purpose Reactor (RSG-GAS) needs special (not commercially available yet) soft wares for analyzing the spectrum of multiple elements in the analysis at once. The application carried out using a single spectrum software analyzer, and comparing each result manually. This method really degrades the quality of the analysis significantly. To solve the problem, a computer code was designed and developed for comparative NAA. Spectrum analysis in the code is carried out using a non-linear fitting method. Before the spectrum analyzed, it was passed to the numerical filter which improves the signal to noise ratio to do the deconvolution operation. The software was developed using the G language and named as PASAN-K The testing result of the developed software was benchmark with the IAEA spectrum and well operated with less than 10 % deviation

  17. Near term hybrid passenger vehicle development program, phase 1

    Science.gov (United States)

    1980-01-01

    Missions for hybrid vehicles that promise to yield high petroleum impact were identified and a preliminary design, was developed that satisfies the mission requirements and performance specifications. Technologies that are critical to successful vehicle design, development and fabrication were determined. Trade-off studies to maximize fuel savings were used to develop initial design specifications of the near term hybrid vehicle. Various designs were "driven" through detailed computer simulations which calculate the petroleum consumption in standard driving cycles, the petroleum and electricity consumptions over the specified missions, and the vehicle's life cycle costs over a 10 year vehicle lifetime. Particular attention was given to the selection of the electric motor, heat engine, drivetrain, battery pack and control system. The preliminary design reflects a modified current compact car powered by a currently available turbocharged diesel engine and a 24 kW (peak) compound dc electric motor.

  18. Hybrid Underwater Vehicle: ARV Design and Development

    Directory of Open Access Journals (Sweden)

    Zhigang DENG

    2014-02-01

    Full Text Available The development of SMU-I, a new autonomous & remotely-operated vehicle (ARV is described. Since it has both the characteristics of autonomous underwater vehicle (AUV and remote operated underwater vehicle (ROV, it is able to achieve precision fix station operation and manual timely intervention. In the paper the initial design of basic components, such as vehicle, propulsion, batteries etc. and the control design of motion are introduced and analyzed. ROV’s conventional cable is replaced by a fiber optic cable, which makes it available for high-bandwidth real-time video, data telemetry and high-quality teleoperation. Furthermore, with the aid of the manual real-time remote operation and ranging sonar, it also resolves the AUV’s conflicting issue, which can absolutely adapt the actual complex sea environment and satisfy the unknown mission need. The whole battery system is designed as two-battery banks, whose voltages and temperatures are monitored through CAN (controller area network bus to avoid battery fire and explosion. A fuzzy-PID controller is designed for its motion control, including depth control and direction control. The controller synthesizes the advantage of fuzzy control and PID control, utilizes the fuzzy rules to on-line tune the parameters of PID controller, and achieves a better control effect. Experiment results demonstrate to show the effectiveness of the test-bed.

  19. Development of health effect assessment software using MACCS2 code

    International Nuclear Information System (INIS)

    Hwang, Seok-Won; Park, Jong-Woon; Kang, Kyung Min; Jae, Moosung

    2008-01-01

    The extended regulatory interests in severe accidents management and enhanced safety regulatory requirements raise a need of more accurate analysis of the effect to the public health by users with diverse disciplines. This facilitates this work to develop web-based radiation health effect assessment software, RASUM, by using the MACCS2 code and HTML language to provide diverse users (regulators, operators, and public) with easy understanding, modeling, calculating, analyzing, documenting and reporting of the radiation health effect under hypothetical severe accidents. The engine of the web-based RASUM uses the MACCS2 as a base code developed by NRC and is composed of five modules such as development module, PSA training module, output module, input data module (source term, population distribution, meteorological data, etc.), and MACCS2 run module. For verification and demonstration of the RASUM, the offsite consequence analysis using the RASUM frame is performed for such as early fatality risk, organ does, and whole body does for two selected scenarios. Moreover, CCDF results from the RASUM for KSNP and CANDU type reactors are presented and compared. (author)

  20. Perspectives in the development of hybrid bifunctional antitumour agents.

    Science.gov (United States)

    Musso, Loana; Dallavalle, Sabrina; Zunino, Franco

    2015-08-15

    In spite of the development of a large number of novel target-specific antitumour agents, the single-agent therapy is in general not able to provide an effective durable control of the malignant process. The limited efficacy of the available agents (both conventional cytotoxic and novel target-specific) reflects not only the expression of defence mechanisms, but also the complexity of tumour cell alterations and the redundancy of survival pathways, thus resulting in tumour cell ability to survive under stress conditions. A well-established strategy to improve the efficacy of antitumour therapy is the rational design of drug combinations aimed at achieving synergistic effects and overcoming drug resistance. An alternative strategy could be the use of agents designed to inhibit simultaneously multiple cellular targets relevant to tumour growth/survival. Among these novel agents are hybrid bifunctional drugs, i.e. compounds resulting by conjugation of different drugs or containing the pharmocophores of different drugs. This strategy has been pursued using various conventional or target-specific agents (with DNA damaging agents and histone deacetylase inhibitors as the most exploited compounds). A critical overview of the most representative compounds is provided with emphasis on the HDAC inhibitor-based hybrid agents. In spite of some promising results, the actual pharmacological advantages of the hybrid agents remain to be defined. This commentary summarizes the recent advances in this field and highlights the pharmacological basis for a rational design of hybrid bifunctional agents. Copyright © 2015. Published by Elsevier Inc.

  1. Development and assessment of ASTEC code for severe accident simulation

    International Nuclear Information System (INIS)

    Van Dorsselaere, J.P.; Pignet, S.; Seropian, C.; Montanelli, T.; Giordano, P.; Jacq, F.; Schwinges, B.

    2005-01-01

    Full text of publication follows: The ASTEC integral code, jointly developed by IRSN and GRS since several years for evaluation of source term during a severe accident (SA) in a Light Water Reactor, will play a central role in the SARNET network of excellence of the 6. Framework Programme (FwP) of the European Commission which started in spring 2004. It should become the reference European SA integral code in the next years. The version V1.1, released in June 2004, allows to model most of the main physical phenomena (except steam explosion) near or at the state of the art. In order to allow to study a great number of scenarios, a compromise must be found between precision of results and calculation time: one day of accident time usually takes less than one day of real time to be simulated on a PC computer. Important efforts are being made on validation by covering more than 30 reference experiments, often International Standard Problems from OECD (CORA, LOFT, PACTEL, BETA, VANAM, ACE-RTF, Phebus.FPT1...). The code is also used for the detailed interpretation of all the integral Phebus.FP experiments. Eighteen European partners performed a first independent evaluation of the code capabilities in 2000-03 within the frame of the EVITA 5. FwP project on one hand by comparison to experiments and on another hand by benchmarking with MAAP4 and MELCOR integral codes on plant applications on PWR and VVER. Their main conclusions were the needs of improvement of code robustness (especially the 2 new modules CESAR and DIVA simulating respectively circuit thermal hydraulics and core degradation) and of post-processing tools. Some improvements have already been achieved in the latest version V 1.1 on these two aspects. A new module MEDICIS devoted to Molten Core Concrete Interaction (MCCI) is implemented in this version, with a tight coupling to the containment thermal hydraulics module CPA. The paper presents a detailed analysis of a TMLB sequence on a French 900 MWe PWR, from

  2. Development of a safety analysis code for molten salt reactors

    International Nuclear Information System (INIS)

    Zhang Dalin; Qiu Suizheng; Su Guanghui

    2009-01-01

    The molten salt reactor (MSR) well suited to fulfill the criteria defined by the Generation IV International Forum (GIF) is presently revisited all around the world because of different attractive features of current renewed relevance. The MSRs are characterized by using the fluid-fuel, so that their technologies are fundamentally different from those used in the conventional solid-fuel reactors. In this work, in particular, the attention is focused on the safety characteristic analysis of the MSRs, in which a point kinetic model considering the flow effects of the fuel salt is established for the MSRs and calculated by developing a microcomputer code coupling with a simplified heat transfer model in the core. The founded models and developed code are applied to analyze the safety characteristics of the molten salt actinide recycler and transmuter system (MOSART) by simulating three types of basic transient conditions including the unprotected loss of flow, unprotected overcooling accident and unprotected transient overpower. Some reasonable results are obtained for the MOSART, which show that the MOSART conceptual design is an inherently stable reactor design. The present study provides some valuable information for the research and design of the new generation MSRs.

  3. Qualifying codes under software quality assurance: Two examples as guidelines for codes that are existing or under development

    Energy Technology Data Exchange (ETDEWEB)

    Mangold, D.

    1993-05-01

    Software quality assurance is an area of concem for DOE, EPA, and other agencies due to the poor quality of software and its documentation they have received in the past. This report briefly summarizes the software development concepts and terminology increasingly employed by these agencies and provides a workable approach to scientific programming under the new requirements. Following this is a practical description of how to qualify a simulation code, based on a software QA plan that has been reviewed and officially accepted by DOE/OCRWM. Two codes have recently been baselined and qualified, so that they can be officially used for QA Level 1 work under the DOE/OCRWM QA requirements. One of them was baselined and qualified within one week. The first of the codes was the multi-phase multi-component flow code TOUGH version 1, an already existing code, and the other was a geochemistry transport code STATEQ that was under development The way to accomplish qualification for both types of codes is summarized in an easy-to-follow step-by step fashion to illustrate how to baseline and qualify such codes through a relatively painless procedure.

  4. Qualifying codes under software quality assurance: Two examples as guidelines for codes that are existing or under development

    International Nuclear Information System (INIS)

    Mangold, D.

    1993-05-01

    Software quality assurance is an area of concern for DOE, EPA, and other agencies due to the poor quality of software and its documentation they have received in the past. This report briefly summarizes the software development concepts and terminology increasingly employed by these agencies and provides a workable approach to scientific programming under the new requirements. Following this is a practical description of how to qualify a simulation code, based on a software QA plan that has been reviewed and officially accepted by DOE/OCRWM. Two codes have recently been baselined and qualified, so that they can be officially used for QA Level 1 work under the DOE/OCRWM QA requirements. One of them was baselined and qualified within one week. The first of the codes was the multi-phase multi-component flow code TOUGH version 1, an already existing code, and the other was a geochemistry transport code STATEQ that was under development The way to accomplish qualification for both types of codes is summarized in an easy-to-follow step-by step fashion to illustrate how to baseline and qualify such codes through a relatively painless procedure

  5. A hybrid WDM/OCDMA ring with a dynamic add/drop function based on Fourier code for local area networks.

    Science.gov (United States)

    Choi, Yong-Kyu; Hosoya, Kenta; Lee, Chung Ghiu; Hanawa, Masanori; Park, Chang-Soo

    2011-03-28

    We propose and experimentally demonstrate a hybrid WDM/OCDMA ring with a dynamic add/drop function based on Fourier code for local area networks. Dynamic function is implemented by mechanically tuning the Fourier encoder/decoder for optical code division multiple access (OCDMA) encoding/decoding. Wavelength division multiplexing (WDM) is utilized for node assignment and 4-chip Fourier code recovers the matched signal from the codes. For an optical source well adapted to WDM channels and its short optical pulse generation, reflective semiconductor optical amplifiers (RSOAs) are used with a fiber Bragg grating (FBG) and gain-switched. To demonstrate we experimentally investigated a two-node hybrid WDM/OCDMA ring with a 4-chip Fourier encoder/decoder fabricated by cascading four FBGs with the bit error rate (BER) of <10(-9) for the node span of 10.64 km at 1.25 Gb/s.

  6. The development of hybrid vehicles for the Indian market

    Energy Technology Data Exchange (ETDEWEB)

    Wadhera, R.; Srinivas, A. [Mahindra and Mahindra Ltd., Mumbai (India); Ellinger, R.; Sattler, M. [AVL List GmbH, Graz (Austria)

    2012-11-01

    The passenger vehicle market in India continues to witness a steady growth rate and expected to do so in the future. While the growth rate is primarily driven by the conventional Internal Combustion based engines, Indian automotive industry is exploring various technological options to comply imminent regulations, remain competitive, reduce emissions and ownership cost. Based on the road trials conducted in various parts of India and subsequent analysis of the traffic characteristics, Hybrid Electric Vehicle promises to be appropriate technology for the near future so as to improve fuel efficiency, performance and reduced emissions. Mahindra and Mahindra has established a clear roadmap for development of HEV. Hybrid Electric Vehicles (HEV's) are seen as a major technology route in order to fulfill future automotive demands with regard to fuel efficiency, emissions, performance and drivability. Mahindra and Mahindra has therefore set-up a clear strategy and roadmap for the development and implementation of HEV's in their product portfolio. The development of hybrid electrical vehicle is also confronted by numerous technical challenges such as Thermal Management, Safety Concept, durability requirements and Noise vibration and Harshness. Also the cost-effectiveness needs to be ensured so as to have a bright market penetration and a viable business. This lecture covers the passenger vehicle sales statistics in India, prevailing traffic characteristics in various types of roads, environmental boundaries and challenges encountered for Indian Market. (orig.)

  7. Developing improved MD codes for understanding processive cellulases

    International Nuclear Information System (INIS)

    Crowley, M F; Nimlos, M R; Himmel, M E; Uberbacher, E C; Iii, C L Brooks; Walker, R C

    2008-01-01

    The mechanism of action of cellulose-degrading enzymes is illuminated through a multidisciplinary collaboration that uses molecular dynamics (MD) simulations and expands the capabilities of MD codes to allow simulations of enzymes and substrates on petascale computational facilities. There is a class of glycoside hydrolase enzymes called cellulases that are thought to decrystallize and processively depolymerize cellulose using biochemical processes that are largely not understood. Understanding the mechanisms involved and improving the efficiency of this hydrolysis process through computational models and protein engineering presents a compelling grand challenge. A detailed understanding of cellulose structure, dynamics and enzyme function at the molecular level is required to direct protein engineers to the right modifications or to understand if natural thermodynamic or kinetic limits are in play. Much can be learned about processivity by conducting carefully designed molecular dynamics (MD) simulations of the binding and catalytic domains of cellulases with various substrate configurations, solvation models and thermodynamic protocols. Most of these numerical experiments, however, will require significant modification of existing code and algorithms in order to efficiently use current (terascale) and future (petascale) hardware to the degree of parallelism necessary to simulate a system of the size proposed here. This work will develop MD codes that can efficiently use terascale and petascale systems, not just for simple classical MD simulations, but also for more advanced methods, including umbrella sampling with complex restraints and reaction coordinates, transition path sampling, steered molecular dynamics, and quantum mechanical/molecular mechanical simulations of systems the size of cellulose degrading enzymes acting on cellulose

  8. Development of Parallel Code for the Alaska Tsunami Forecast Model

    Science.gov (United States)

    Bahng, B.; Knight, W. R.; Whitmore, P.

    2014-12-01

    The Alaska Tsunami Forecast Model (ATFM) is a numerical model used to forecast propagation and inundation of tsunamis generated by earthquakes and other means in both the Pacific and Atlantic Oceans. At the U.S. National Tsunami Warning Center (NTWC), the model is mainly used in a pre-computed fashion. That is, results for hundreds of hypothetical events are computed before alerts, and are accessed and calibrated with observations during tsunamis to immediately produce forecasts. ATFM uses the non-linear, depth-averaged, shallow-water equations of motion with multiply nested grids in two-way communications between domains of each parent-child pair as waves get closer to coastal waters. Even with the pre-computation the task becomes non-trivial as sub-grid resolution gets finer. Currently, the finest resolution Digital Elevation Models (DEM) used by ATFM are 1/3 arc-seconds. With a serial code, large or multiple areas of very high resolution can produce run-times that are unrealistic even in a pre-computed approach. One way to increase the model performance is code parallelization used in conjunction with a multi-processor computing environment. NTWC developers have undertaken an ATFM code-parallelization effort to streamline the creation of the pre-computed database of results with the long term aim of tsunami forecasts from source to high resolution shoreline grids in real time. Parallelization will also permit timely regeneration of the forecast model database with new DEMs; and, will make possible future inclusion of new physics such as the non-hydrostatic treatment of tsunami propagation. The purpose of our presentation is to elaborate on the parallelization approach and to show the compute speed increase on various multi-processor systems.

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  10. The FELIX program of experiments and code development

    International Nuclear Information System (INIS)

    Turner, L.R.

    1983-01-01

    An experimental program and test bed called FELIX (Fusion Electromagnetic Induction Experiment) which is under construction at Argonne National Laboratory is described. The facility includes the following facilities; (a) a sizable constant field, analogous to a tokamak toroidal field or the confining field of a mirror reactor, (b) a pulsed field with a sizable rate of change, analogous to a pulsed poloidal field or to the changing field of a plasma disruption, perpendicular to the constant field, and (c) a sufficiently large volume to assure that large, complex test pieces can be tested, and that the forces, torques, currents, and field distortions which are developed are large enough to be measured accurately. The development of the necessary computer codes and the experimental program are examined. (U.K.)

  11. Neutron flux investigation on certain alternative fluids in a hybrid system by using MCNPX Monte Carlo transport code

    Energy Technology Data Exchange (ETDEWEB)

    Guenay, Mehtap [Inoenue Univ., Malatya (Turkey). Physics Dept.

    2014-04-15

    In this study, the molten salt-heavy metal mixtures 93-85 % Li{sub 20}Sn{sub 80} + 5 % SFG-PuO{sub 2} and 2-10 % UO{sub 2}, 93-85 % Li{sub 20}Sn{sub 80} + 5 % SFG-PuO{sub 2} and 2-10 % NpO{sub 2}, 93-85 % Li{sub 20}Sn{sub 80} + 5 % SFG-PuO{sub 2} and 2-10 % UCO were used as fluids. The fluids were used in the liquid first wall, blanket and shield zones of the designed hybrid reactor system. Four centimeter thick 9Cr2WVTa ferritic steel was used as the structural material. In this study, the effect of mixture components on the neutron flux was investigated in a designed fusion-fission hybrid reactor system. The neutron flux was investigated according to the mixture components, radial flux distribution and energy spectrum in the designed system. Three-dimensional analyses were performed using the most recent MCNPX-2.7.0 Monte Carlo radiation transport code and the ENDF/B-VII.0 nuclear data library. (orig.)

  12. Development and application of the BOA code in Spain

    International Nuclear Information System (INIS)

    Tortuero Lopez, C.; Doncel Gutierrez, N.; Culebras, F.

    2012-01-01

    The BOA code allows to quantitatively establish the level of risk of Axial Offset Anomaly and increased deposition of crud on the basis of specific conditions in each case. For this reason, the code is parameterized according to the individual characteristics of each plant. This paper summarizes the results obtained in the implementation of the code, as well as its future perspective.

  13. Time development of cascades by the binary collision approximation code

    International Nuclear Information System (INIS)

    Fukumura, A.; Ishino, S.; Sekimura, N.

    1991-01-01

    To link a molecular dynamic calculation to binary collision approximation codes to explore high energy cascade damage, time between consecutive collisions is introduced into the binary collision MARLOWE code. Calculated results for gold by the modified code show formation of sub-cascades and their spatial and time overlapping, which can affect formation of defect clusters. (orig.)

  14. ESE a 2D compressible multiphase flow code developed for MFCI analysis - code validation

    International Nuclear Information System (INIS)

    Leskovar, M.; Mavko, B.

    1998-01-01

    ESE (Evaluation of Steam Explosions) is a general second order accurate two-dimensional compressible multiphase flow computer code. It has been developed to model the interaction of molten core debris with water during the first premixing stage of a steam explosion. A steam explosion is a physical event, which may occur during a severe reactor accident following core meltdown when the molten fuel comes into contact with the coolant water. Since the exchanges of mass, momentum and energy are regime dependent, different exchange laws have been incorporated in ESE for the major flow regimes. With ESE a number of premixing experiments performed at the Oxford University and at the QUEOS facility at Forschungszentrum Karlsruhe has been simulated. In these premixing experiments different jets of spheres were injected in a water poll. The ESE validation plan was carefully chosen, starting from very simple, well-defined problems, and gradually working up to more complicated ones. The results of ESE simulations, which were compared to experimental data and also to first order accurate calculations, are presented in form graphs. Most of the ESE results agree qualitatively as quantitatively reasonably well with experimental data and in general better than the results obtained with the first order accurate calculation.(author)

  15. Present status of transport code development based on Monte Carlo method

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki

    1985-01-01

    The present status of development in Monte Carlo code is briefly reviewed. The main items are the followings; Application fields, Methods used in Monte Carlo code (geometry spectification, nuclear data, estimator and variance reduction technique) and unfinished works, Typical Monte Carlo codes and Merits of continuous energy Monte Carlo code. (author)

  16. Development of the pneumatic service robot with a hybrid type

    International Nuclear Information System (INIS)

    Choi, Cheol U; Choi, Hyeun Seok; Han, Chang Soo

    2001-01-01

    In this paper, the pneumatic service robot with a hybrid type is developed. A pneumatic has the advantage of good compliance, high payload-to-weight and payload-to-volume ratios, high speed and force capabilities. Using pneumatic actuators which have low stiffness, the service robot can guarantee safety. By suggesting a new serial-parallel hybrid type for the service robot which separates into positioning motion and orienting motion, we can achieve large workspace and high strength-to-moving-weight ratio at the same time. A sliding mode controller can be designed for tracking the desired output using the Lyapunov stability theory and structural properties of pneumatic servo systems. Through many experiments of circular trajectory, the pneumatic service robot is evaluated and verified

  17. Further development of the computer code ATHLET-CD

    International Nuclear Information System (INIS)

    Weber, Sebastian; Austregesilo, Henrique; Bals, Christine; Band, Sebastian; Hollands, Thorsten; Koellein, Carsten; Lovasz, Liviusz; Pandazis, Peter; Schubert, Johann-Dietrich; Sonnenkalb, Martin

    2016-10-01

    In the framework of the reactor safety research program sponsored by the German Federal Ministry for Economic Affairs and Energy (BMWi), the computer code system ATHLET/ATHLET-CD has been further developed as an analysis tool for the simulation of accidents in nuclear power plants with pressurized and boiling water reactors as well as for the evaluation of accident management procedures. The main objective was to provide a mechanistic analysis tool for best estimate calculations of transients, accidents, and severe accidents with core degradation in light water reactors. With the continued development, the capability of the code system has been largely improved, allowing best estimate calculations of design and beyond design base accidents, and the simulation of advanced core degradation with enhanced model extent in a reasonable calculation time. ATHLET comprises inter alia a 6-equation model, models for the simulation of non-condensable gases and tracking of boron concentration, as well as additional component and process models for the complete system simulation. Among numerous model improvements, the code application has been extended to super critical pressures. The mechanistic description of the dynamic development of flow regimes on the basis of a transport equation for the interface area has been further developed. This ATHLET version is completely integrated in ATHLET-CD. ATHLET-CD further comprises dedicated models for the simulation of fuel and control assembly degradation for both pressurized and boiling water reactors, debris bed with melting in the core region, as well as fission product and aerosol release and transport in the cooling system, inclusive of decay of nuclide inventories and of chemical reactions in the gas phase. The continued development also concerned the modelling of absorber material release, of melting, melt relocation and freezing, and the interaction with the wall of the reactor pressure vessel. The following models were newly

  18. The effect on radiation damage of structural material in a hybrid system by using a Monte Carlo radiation transport code

    International Nuclear Information System (INIS)

    Günay, Mehtap; Şarer, Başar; Kasap, Hızır

    2014-01-01

    Highlights: • The effects of some fluids on gas production rates in structural material were investigated. • The MCNPX-2.7.0 Monte Carlo code was used for three-dimensional calculations. • It was found that biggest contribution to gas production rates comes from Fe isotope of the. • The desirable values for 5% SFG-PuO 2 with respect to radiation damage were specified. - Abstract: In this study, the molten salt-heavy metal mixtures 99–95% Li20Sn80-1-5% SFG-Pu, 99–95% Li20Sn80-1-5% SFG-PuF4, 99-95% Li20Sn80-1-5% SFG-PuO2 were used as fluids. The fluids were used in the liquid first-wall, blanket and shield zones of the designed hybrid reactor system. 9Cr2WVTa ferritic steel with the width of 4 cm was used as the structural material. The parameters of radiation damage are proton, deuterium, tritium, He-3 and He-4 gas production rates. In this study, the effects of the selected fluid on the radiation damage, in terms of individual as well as total isotopes in the structural material, were investigated for 30 full power years (FPYs). Three-dimensional analyses were performed using the most recent version of the MCNPX-2.7.0 Monte Carlo radiation transport code and the ENDF/B-VII.0 nuclear data library

  19. Methodology, status and plans for development and assessment of Cathare code

    Energy Technology Data Exchange (ETDEWEB)

    Bestion, D.; Barre, F.; Faydide, B. [CEA - Grenoble (France)

    1997-07-01

    This paper presents the methodology, status and plans for the development, assessment and uncertainty evaluation of the Cathare code. Cathare is a thermalhydraulic code developed by CEA (DRN), IPSN, EDF and FRAMATOME for PWR safety analysis. First, the status of the code development and assessment is presented. The general strategy used for the development and the assessment of the code is presented. Analytical experiments with separate effect tests, and component tests are used for the development and the validation of closure laws. Successive Revisions of constitutive laws are implemented in successive Versions of the code and assessed. System tests or integral tests are used to validate the general consistency of the Revision. Each delivery of a code Version + Revision is fully assessed and documented. A methodology is being developed to determine the uncertainty on all constitutive laws of the code using calculations of many analytical tests and applying the Discrete Adjoint Sensitivity Method (DASM). At last, the plans for the future developments of the code are presented. They concern the optimization of the code performance through parallel computing - the code will be used for real time full scope plant simulators - the coupling with many other codes (neutronic codes, severe accident codes), the application of the code for containment thermalhydraulics. Also, physical improvements are required in the field of low pressure transients and in the modeling for the 3-D model.

  20. What do Codes of Conduct do? Hybrid Constitutionalization and Militarization in Military Markets

    DEFF Research Database (Denmark)

    Leander, Anna

    2012-01-01

    that makes it possible to hold firms accountable and a militarization of politics. It does so by showing that the codes create first-, second- and third-order rules but also processes of misrecognition through distraction, distinction and diffusion that empower military professionals. It draws on a study...... performative but are so in multiple ways. It can provide no easy way to dissolve the specific dilemma this multiple jurisgenerativity poses in the context of military markets specifically. But logically flowing from the argument is a suggestion that encouraging and empowering a broader, non...

  1. Development of a multispectral autoradiography using a coded aperture

    Science.gov (United States)

    Noto, Daisuke; Takeda, Tohoru; Wu, Jin; Lwin, Thet T.; Yu, Quanwen; Zeniya, Tsutomu; Yuasa, Tetsuya; Hiranaka, Yukio; Itai, Yuji; Akatsuka, Takao

    2000-11-01

    Autoradiography is a useful imaging technique to understand biological functions using tracers including radio isotopes (RI's). However, it is not easy to describe the distribution of different kinds of tracers simultaneously by conventional autoradiography using X-ray film or Imaging plate. Each tracer describes each corresponding biological function. Therefore, if we can simultaneously estimate distribution of different kinds of tracer materials, the multispectral autoradiography must be a quite powerful tool to better understand physiological mechanisms of organs. So we are developing a system using a solid state detector (SSD) with high energy- resolution. Here, we introduce an imaging technique with a coded aperture to get spatial and spectral information more efficiently. In this paper, the imaging principle is described, and its validity and fundamental property are discussed by both simulation and phantom experiments with RI's such as 201Tl, 99mTc, 67Ga, and 123I.

  2. Recent development of three-dimensional piping code SHAPS

    International Nuclear Information System (INIS)

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

    1985-01-01

    This paper describes the recent development of the three-dimensional, structural, and hydrodynamic analysis piping code SHAPS. Several new features have been incorporated into the program, including (1) an elbow hydrodynamic model for analyzing the effect of global motion on the pressure-wave propagation, (2) a component hydrodynamic model for treating fluid motion in the vicinity of rigid obstacles and baffle plates, (3) the addition of the implicit time integration scheme in the structural-dynamic analysis, (4) the option of an implicit-implicit fluid-structural linking scheme, and (5) provisions for two constitutive equations for materials under various loading conditions. Sample problems are given to illustrate these features. Their results are discussed in detail. 7 refs., 8 figs

  3. Analyses to support development of risk-informed separation distances for hydrogen codes and standards.

    Energy Technology Data Exchange (ETDEWEB)

    LaChance, Jeffrey L.; Houf, William G. (Sandia National Laboratories, Livermore, CA); Fluer, Inc., Paso Robels, CA; Fluer, Larry (Fluer, Inc., Paso Robels, CA); Middleton, Bobby

    2009-03-01

    The development of a set of safety codes and standards for hydrogen facilities is necessary to ensure they are designed and operated safely. To help ensure that a hydrogen facility meets an acceptable level of risk, code and standard development organizations are tilizing risk-informed concepts in developing hydrogen codes and standards.

  4. The newly developed Toyota plug-in hybrid system

    Energy Technology Data Exchange (ETDEWEB)

    Takaoka, Toshifumi; Ichinose, Hiroki [Toyota Motor Corporation (Japan)

    2010-07-01

    Toyota has been introducing several hybrid vehicles (HV) as a countermeasure to the automobile's concerns, like CO2 reduction, energy security, and emission reduction in urban areas. A next step towards an even more effective solution for these concerns is a plug-in hybrid vehicle (PHV). This vehicle combines the advantages of electric vehicles (EV), which use clean electric energy, and HV, with it's high environmental potential and user- friendliness comparable to conventional vehicles, such as a long cruising range. This paper describes a newly developed plug-in hybrid system and its vehicle performance. This system uses a Li-ion battery with high energy density and has an affordable EV range without sacrificing cabin space. The vehicle achieves a CO2 emission of 59g/km and meets the most stringent emission regulations in the world. The new PHV is a forerunner of the large-scale mass production PHV two years later. PHVs have the potential to become popular as a realistic solution towards sustainable mobility by renewable electricity usage in the future. (orig.)

  5. On the development of an intrinsic hybrid composite

    International Nuclear Information System (INIS)

    Kießling, R; Ihlemann, J; Riemer, M; Drossel, W-G; Scharf, I; Lampke, T; Sharafiev, S; Pouya, M; Wagner, M F-X

    2016-01-01

    Hybrid parts, which combine low weight with high strength, are moving into the focus of the automotive industry, due to their high potential for usage in the field of crash-relevant structures. In this contribution, the development of an intrinsic hybrid composite is presented, with a focus on the manufacturing process, complex simulations of the material behaviour and material testing. The hybrid composite is made up of a continuous fibre- reinforced plastic (FRP), in which a metallic insert is integrated. The mechanical behaviour of the individual components is characterised. For material modelling, an approach is pointed out that enables modelling at large strains by directly connected rheological elements. The connection between the FRP and the metallic insert is realised by a combination of form fit and adhesive bonds. On the one hand, adhesive bonds are generated within a sol gel process. On the other hand, local form elements of the metallic insert are pressed into the FRP. We show how these form elements are generated during the macroscopic forming process. In addition, the applied sol gel process is explained. Finally, we consider design concepts for a specimen type for high strain testing of the resulting interfaces. (paper)

  6. Development of a solar-hydrogen hybrid energy system

    International Nuclear Information System (INIS)

    Sebastian, P.J.; Gamboa, S.A.; Vejar, Set; Campos, J.

    2009-01-01

    Full text: The details of the development of a PV-hydrogen hybrid energy system is presented. An arrangement of photovoltaic modules (125 W/module) was established to provide 9 kW installed power in a three-phase configuration at 127 Vrms/phase. A 5 kW fuel cell system (hydrogen/oxygen) operate as a dynamic backup of the photovoltaic system. The autonomous operation of the hybrid power system implies the production of hydrogen by electrolysis. The hydrogen is produced by water electrolysis using an electrolyzer of 1 kW power. The electrical energy used to produce hydrogen is supplied from solar panels by using 1kW of photovoltaic modules. The photovoltaic modules are installed in a sun-tracker arrangement for increasing the energy conversion efficiency. The hydrogen is stored in solar to electric commercial metal hydride based containers and supplied to the fuel cell. The hybrid system is monitored by internet and some dynamic characteristics such as demanding power, energy and power factor could be analyzed independently from the system. Some energy saving recommendations has been implemented as a pilot program at CIE-UNAM to improve the efficient use of clean energy in normal operating conditions in offices and laboratories. (author)

  7. 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.)

  8. 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.)

  9. Development of general-purpose particle and heavy ion transport monte carlo code

    International Nuclear Information System (INIS)

    Iwase, Hiroshi; Nakamura, Takashi; Niita, Koji

    2002-01-01

    The high-energy particle transport code NMTC/JAM, which has been developed at JAERI, was improved for the high-energy heavy ion transport calculation by incorporating the JQMD code, the SPAR code and the Shen formula. The new NMTC/JAM named PHITS (Particle and Heavy-Ion Transport code System) is the first general-purpose heavy ion transport Monte Carlo code over the incident energies from several MeV/nucleon to several GeV/nucleon. (author)

  10. Learning Concepts, Language, and Literacy in Hybrid Linguistic Codes: The Multilingual Maze of Urban Grade 1 Classrooms in South Africa

    Science.gov (United States)

    Henning, Elizabeth

    2012-01-01

    From the field of developmental psycholinguistics and from conceptual development theory there is evidence that excessive linguistic "code-switching" in early school education may pose some hazards for the learning of young multilingual children. In this article the author addresses the issue, invoking post-Piagetian and neo-Vygotskian…

  11. Development of the code package KASKAD for calculations of WWERs

    International Nuclear Information System (INIS)

    Bolobov, P.A.; Lazarenko, A.P.; Tomilov, M.Ju.

    2008-01-01

    The new version of software package for neutron calculation of WWER cores KASKAD 2007 consists of some calculating and service modules, which are integrated in the common framework. The package is based on the old version, which was expanded with some new functions and the new calculating modules, such as: -the BIPR-2007 code is the new one which performs calculation of power distribution in three-dimensional geometry for 2-group neutron diffusion calculation. This code is based on the BIPR-8KN model, provides all possibilities of BIPR-7A code and uses the same input data; -the PERMAK-2007 code is pin-by-pin few-group multilayer and 3-D code for neutron diffusion calculation; -graphical user interface for input data preparation of the TVS-M code. The report also includes some calculation results obtained with modified version of the KASKAD 2007 package. (Authors)

  12. Development of multi-group spectral code TVS-M

    International Nuclear Information System (INIS)

    Lazarenko, A. P.; Pryanichnikov, A. V.; Kalugin, M. A.; Gurevich, M. I.

    2011-01-01

    This paper is dedicated to the latest version of TVS-M code - TVS-M 2007, which allows the neutron flux distribution inside fuel assemblies to be calculated without using the diffusion approximation. The new spatial calculation module PERST introduced in TBS-M code is based on the first collisions probability method and allows the scattering anisotropy to be accounted for. This paper presents some preliminary results calculated with the use of the new version of TVS-M code. (Authors)

  13. Development of hybrid artificial intelligent based handover decision algorithm

    Directory of Open Access Journals (Sweden)

    A.M. Aibinu

    2017-04-01

    Full Text Available The possibility of seamless handover remains a mirage despite the plethora of existing handover algorithms. The underlying factor responsible for this has been traced to the Handover decision module in the Handover process. Hence, in this paper, the development of novel hybrid artificial intelligent handover decision algorithm has been developed. The developed model is made up of hybrid of Artificial Neural Network (ANN based prediction model and Fuzzy Logic. On accessing the network, the Received Signal Strength (RSS was acquired over a period of time to form a time series data. The data was then fed to the newly proposed k-step ahead ANN-based RSS prediction system for estimation of prediction model coefficients. The synaptic weights and adaptive coefficients of the trained ANN was then used to compute the k-step ahead ANN based RSS prediction model coefficients. The predicted RSS value was later codified as Fuzzy sets and in conjunction with other measured network parameters were fed into the Fuzzy logic controller in order to finalize handover decision process. The performance of the newly developed k-step ahead ANN based RSS prediction algorithm was evaluated using simulated and real data acquired from available mobile communication networks. Results obtained in both cases shows that the proposed algorithm is capable of predicting ahead the RSS value to about ±0.0002 dB. Also, the cascaded effect of the complete handover decision module was also evaluated. Results obtained show that the newly proposed hybrid approach was able to reduce ping-pong effect associated with other handover techniques.

  14. Development of an Auto-Validation Program for MARS Code Assessments

    International Nuclear Information System (INIS)

    Lee, Young Jin; Chung, Bub Dong

    2006-01-01

    MARS (Multi-dimensional Analysis of Reactor Safety) code is a best-estimate thermal hydraulic system analysis code developed at KAERI. It is important for a thermal hydraulic computer code to be assessed against theoretical and experimental data to verify and validate the performance and the integrity of the structure, models and correlations of the code. The code assessment efforts for complex thermal hydraulics code such as MARS code can be tedious, time-consuming and require large amount of human intervention in data transfer to see the results in graphic forms. Code developers produce many versions of a code during development and each version need to be verified for integrity. Thus, for MARS code developers, it is desirable to have an automatic way of carrying out the code assessment calculations. In the present work, an Auto-Validation program that carries out the code assessment efforts has been developed. The program uses the user supplied configuration file (with '.vv' extension) which contain commands to read input file, to execute the user selected MARS program, and to generate result graphs. The program can be useful if a same set of code assessments is repeated with different versions of the code. The program is written with the Delphi program language. The program runs under the Microsoft Windows environment

  15. Developing HYDMN code to include the transient of MNSR

    International Nuclear Information System (INIS)

    Al-Barhoum, M.

    2000-11-01

    A description of the programs added to HYDMN code (a code for thermal-hydraulic steady state of MNSR) to include the transient of the same MNSR is presented. The code asks the initial conditions for the power (in k W) and the cold initial core inlet temperature (in degrees centigrade). A time-dependent study of the coolant inlet and outlet temperature, its speed, pool and tank temperatures is done for MNSR in general and for the Syrian MNSR in particular. The study solves the differential equations taken from reference (1) by using some numerical methods found in reference (3). The code becomes this way independent of any external information source. (Author)

  16. Code development and analyses within the area of transmutation and safety

    International Nuclear Information System (INIS)

    Maschek, W.

    2002-01-01

    A strong code development is going on to meet various demands resulting from the development of dedicated reactors for transmutation and incineration. Code development is concerned with safety codes and general codes needed for assessing scenarios and transmutation strategies. Analyses concentrate on various ADS systems with solid and liquid molten salt fuels. Analyses deal with ADS Demo Plant (5th FP EU) and transmuters with advanced fuels

  17. Development status of Severe Accident Analysis Code SAMPSON

    International Nuclear Information System (INIS)

    Iwashita, Tsuyoshi; Ujita, Hiroshi

    2000-01-01

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

  18. Hybrid and Parallel Domain-Decomposition Methods Development to Enable Monte Carlo for Reactor Analyses

    International Nuclear Information System (INIS)

    Wagner, John C.; Mosher, Scott W.; Evans, Thomas M.; Peplow, Douglas E.; Turner, John A.

    2010-01-01

    This paper describes code and methods development at the Oak Ridge National Laboratory focused on enabling high-fidelity, large-scale reactor analyses with Monte Carlo (MC). Current state-of-the-art tools and methods used to perform real commercial reactor analyses have several undesirable features, the most significant of which is the non-rigorous spatial decomposition scheme. Monte Carlo methods, which allow detailed and accurate modeling of the full geometry and are considered the gold standard for radiation transport solutions, are playing an ever-increasing role in correcting and/or verifying the deterministic, multi-level spatial decomposition methodology in current practice. However, the prohibitive computational requirements associated with obtaining fully converged, system-wide solutions restrict the role of MC to benchmarking deterministic results at a limited number of state-points for a limited number of relevant quantities. The goal of this research is to change this paradigm by enabling direct use of MC for full-core reactor analyses. The most significant of the many technical challenges that must be overcome are the slow, non-uniform convergence of system-wide MC estimates and the memory requirements associated with detailed solutions throughout a reactor (problems involving hundreds of millions of different material and tally regions due to fuel irradiation, temperature distributions, and the needs associated with multi-physics code coupling). To address these challenges, our research has focused on the development and implementation of (1) a novel hybrid deterministic/MC method for determining high-precision fluxes throughout the problem space in k-eigenvalue problems and (2) an efficient MC domain-decomposition (DD) algorithm that partitions the problem phase space onto multiple processors for massively parallel systems, with statistical uncertainty estimation. The hybrid method development is based on an extension of the FW-CADIS method, which

  19. Hybrid and parallel domain-decomposition methods development to enable Monte Carlo for reactor analyses

    International Nuclear Information System (INIS)

    Wagner, J.C.; Mosher, S.W.; Evans, T.M.; Peplow, D.E.; Turner, J.A.

    2010-01-01

    This paper describes code and methods development at the Oak Ridge National Laboratory focused on enabling high-fidelity, large-scale reactor analyses with Monte Carlo (MC). Current state-of-the-art tools and methods used to perform 'real' commercial reactor analyses have several undesirable features, the most significant of which is the non-rigorous spatial decomposition scheme. Monte Carlo methods, which allow detailed and accurate modeling of the full geometry and are considered the 'gold standard' for radiation transport solutions, are playing an ever-increasing role in correcting and/or verifying the deterministic, multi-level spatial decomposition methodology in current practice. However, the prohibitive computational requirements associated with obtaining fully converged, system-wide solutions restrict the role of MC to benchmarking deterministic results at a limited number of state-points for a limited number of relevant quantities. The goal of this research is to change this paradigm by enabling direct use of MC for full-core reactor analyses. The most significant of the many technical challenges that must be overcome are the slow, non-uniform convergence of system-wide MC estimates and the memory requirements associated with detailed solutions throughout a reactor (problems involving hundreds of millions of different material and tally regions due to fuel irradiation, temperature distributions, and the needs associated with multi-physics code coupling). To address these challenges, our research has focused on the development and implementation of (1) a novel hybrid deterministic/MC method for determining high-precision fluxes throughout the problem space in k-eigenvalue problems and (2) an efficient MC domain-decomposition (DD) algorithm that partitions the problem phase space onto multiple processors for massively parallel systems, with statistical uncertainty estimation. The hybrid method development is based on an extension of the FW-CADIS method

  20. Development of three-dimensional neoclassical transport simulation code with high performance Fortran on a vector-parallel computer

    International Nuclear Information System (INIS)

    Satake, Shinsuke; Okamoto, Masao; Nakajima, Noriyoshi; Takamaru, Hisanori

    2005-11-01

    A neoclassical transport simulation code (FORTEC-3D) applicable to three-dimensional configurations has been developed using High Performance Fortran (HPF). Adoption of computing techniques for parallelization and a hybrid simulation model to the δf Monte-Carlo method transport simulation, including non-local transport effects in three-dimensional configurations, makes it possible to simulate the dynamism of global, non-local transport phenomena with a self-consistent radial electric field within a reasonable computation time. In this paper, development of the transport code using HPF is reported. Optimization techniques in order to achieve both high vectorization and parallelization efficiency, adoption of a parallel random number generator, and also benchmark results, are shown. (author)

  1. Development of seismic tomography software for hybrid supercomputers

    Science.gov (United States)

    Nikitin, Alexandr; Serdyukov, Alexandr; Duchkov, Anton

    2015-04-01

    Seismic tomography is a technique used for computing velocity model of geologic structure from first arrival travel times of seismic waves. The technique is used in processing of regional and global seismic data, in seismic exploration for prospecting and exploration of mineral and hydrocarbon deposits, and in seismic engineering for monitoring the condition of engineering structures and the surrounding host medium. As a consequence of development of seismic monitoring systems and increasing volume of seismic data, there is a growing need for new, more effective computational algorithms for use in seismic tomography applications with improved performance, accuracy and resolution. To achieve this goal, it is necessary to use modern high performance computing systems, such as supercomputers with hybrid architecture that use not only CPUs, but also accelerators and co-processors for computation. The goal of this research is the development of parallel seismic tomography algorithms and software package for such systems, to be used in processing of large volumes of seismic data (hundreds of gigabytes and more). These algorithms and software package will be optimized for the most common computing devices used in modern hybrid supercomputers, such as Intel Xeon CPUs, NVIDIA Tesla accelerators and Intel Xeon Phi co-processors. In this work, the following general scheme of seismic tomography is utilized. Using the eikonal equation solver, arrival times of seismic waves are computed based on assumed velocity model of geologic structure being analyzed. In order to solve the linearized inverse problem, tomographic matrix is computed that connects model adjustments with travel time residuals, and the resulting system of linear equations is regularized and solved to adjust the model. The effectiveness of parallel implementations of existing algorithms on target architectures is considered. During the first stage of this work, algorithms were developed for execution on

  2. PURE OR HYBRID? THE DEVELOPMENT OF MIXED DICTIONARY GENRES

    Directory of Open Access Journals (Sweden)

    Reinhard R. K. Hartmann

    2005-11-01

    Full Text Available This paper explores 'hybrid' genres of dictionaries and other reference works. Against the tradition of general dictionaries becoming ever more specialised, there has also been a growing trend of mixing two or more 'pure' dictionary types for achieving specific purposes, e.g. the combination of alphabetic and thematic dictionary,general dictionary and technical glossary, dictionary and thesaurus, dictionary and encyclopedia, monolingual and bilingual dictionary, etc. Examples of these various sub-types are discussed (admitting that dictionary research has neglected their study,with the aim of determining overall trends and implications, particularly with regard to the possibility of their further development with the means of information technology.

  3. Optimization and Validation of the Developed Uranium Isotopic Analysis Code

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

    γ-ray spectroscopy is a representative non-destructive assay for nuclear material, and less time-consuming and less expensive than the destructive analysis method. The destructive technique is more precise than NDA technique, however, there is some correction algorithm which can improve the performance of γ-spectroscopy. For this reason, an analysis code for uranium isotopic analysis is developed by Applied Nuclear Physics Group in Seoul National University. Overlapped γ- and x-ray peaks in the 89-101 keV X{sub α}-region are fitted with Gaussian and Lorentzian distribution peak functions, tail and background functions. In this study, optimizations for the full-energy peak efficiency calibration and fitting parameters of peak tail and background are performed, and validated with 24 hour acquisition of CRM uranium samples. The optimization of peak tail and background parameters are performed with the validation by using CRM uranium samples. The analysis performance is improved in HEU samples, but more optimization of fitting parameters is required in LEU sample analysis. In the future, the optimization research about the fitting parameters with various type of uranium samples will be performed. {sup 234}U isotopic analysis algorithms and correction algorithms (coincidence effect, self-attenuation effect) will be developed.

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

    International Nuclear Information System (INIS)

    Shi, Chengbin; Cheng, Maosong; Liu, Guimin

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-15

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

  6. Development on hybrid evaluated nuclear data library HENDL1.0/MG/MC

    International Nuclear Information System (INIS)

    Xu Dezheng; Gao Chunjing; Zheng Shanliang; Liu Haibo; Zhu Xiaoxiang; Li Jingjing; Wu Yican

    2004-01-01

    A Hybrid Evaluated Nuclear Data Library (HENDL) named as HENDL1.0 has been developed by Fusion Design Study (FDS) team of Institute of Plasma Physics, Academia Sinica (ASIPP) to take into account the requirements in design and research relevant to fusion, fission and fusion-fission sub-critical hybrid reactor. HENDLI1.0 contains one basic evaluated sub-library naming HENDL1.0/E and to processed working sub-libraries naming HENDL1.0/MG and HENDL1.0/MC, respectively. Through carefully comparing, distinguishing and choosing, HENDL1.0/E integrated basic evaluated neutron data files of 213 nuclides from the several main data libraries for evaluated neutron reaction cross sections including ENDF/B-VI (USA), JEF-2.2 (OECD/NEA, Europe), JENDL-3.2 (Japan), CENDL-2 (China), BROND-2 (Russia) and FENDL-2 (IAEA/NDS, ITER program). Based on this, 175-group neutron and 42-group photon neutron-photon coupled multi-group working library HENDL1.0/MG used for discrete ordinate Sn method transport calculation (such as ANISN code) and a compact ENDF form (ACE), continuous energy structure (pointwise) neutron cross section library HENDL1.0/MC for Monte Carlo method transport simulation (as MCMP code) can be attainable with the current group constants processing system NJOY and transport cross section preparation code TRANSX referring to the Vitamin-J energy group structure. In addition, two special bases i.e. transmutation (burnup) library BURNUP. DAT and response function library RESPONSE.DAT, have been also made for fuel cycle calculation and reactivity analyses of nuclear reactor. The relevant sample testing, benchmark checking and primary confirmation are also carried out to assess the validity of multi-purpose data library HENDL1.0. (authors)

  7. Trends in EFL Technology and Educational Coding: A Case Study of an Evaluation Application Developed on LiveCode

    Science.gov (United States)

    Uehara, Suwako; Noriega, Edgar Josafat Martinez

    2016-01-01

    The availability of user-friendly coding software is increasing, yet teachers might hesitate to use this technology to develop for educational needs. This paper discusses studies related to technology for educational uses and introduces an evaluation application being developed. Through questionnaires by student users and open-ended discussion by…

  8. Progress in ICRH and lower hybrid launcher development

    International Nuclear Information System (INIS)

    Kaye, A.S.

    1993-01-01

    Radio frequency methods of heating and non-inductive current drive have become well established and are likely to be part of any next-step Tokamak programme. The present state of development of antennae for ion cyclotron heating and recent developments to enhance the effectiveness of fast wave current drive systems are reviewed. The performance achieved by present systems enables the provision of an ICRH system for next step devices within the existing technology limits. The main Lower Hybrid current drive systems are also reviewed. Present operating limits suggest that the design power density at the grill in large multijunction launchers must be somewhat reduced due to peaking of the electric field. The resulting launcher for a next step machine based on present technology is a large and highly complex device. Development of recent proposals such as the rod array or the hyperguide, in parallel with necessary improvements in the current drive efficiency, would make Lower Hybrid a more attractive method of non-inductive current drive for next step machines. (Author)

  9. Hybridization-based reconstruction of small non-coding RNA transcripts from deep sequencing data.

    Science.gov (United States)

    Ragan, Chikako; Mowry, Bryan J; Bauer, Denis C

    2012-09-01

    Recent advances in RNA sequencing technology (RNA-Seq) enables comprehensive profiling of RNAs by producing millions of short sequence reads from size-fractionated RNA libraries. Although conventional tools for detecting and distinguishing non-coding RNAs (ncRNAs) from reference-genome data can be applied to sequence data, ncRNA detection can be improved by harnessing the full information content provided by this new technology. Here we present NorahDesk, the first unbiased and universally applicable method for small ncRNAs detection from RNA-Seq data. NorahDesk utilizes the coverage-distribution of small RNA sequence data as well as thermodynamic assessments of secondary structure to reliably predict and annotate ncRNA classes. Using publicly available mouse sequence data from brain, skeletal muscle, testis and ovary, we evaluated our method with an emphasis on the performance for microRNAs (miRNAs) and piwi-interacting small RNA (piRNA). We compared our method with Dario and mirDeep2 and found that NorahDesk produces longer transcripts with higher read coverage. This feature makes it the first method particularly suitable for the prediction of both known and novel piRNAs.

  10. Ordered particles versus ordered pointers in the hybrid ordered plasma simulation (HOPS) code

    International Nuclear Information System (INIS)

    Anderson, D.V.; Shumaker, D.E.

    1993-01-01

    From a computational standpoint, particle simulation calculations for plasmas have not adapted well to the transitions from scalar to vector processing nor from serial to parallel environments. They have suffered from inordinate and excessive accessing of computer memory and have been hobbled by relatively inefficient gather-scatter constructs resulting from the use of indirect indexing. Lastly, the many-to-one mapping characteristic of the deposition phase has made it difficult to perform this in parallel. The authors' code sorts and reorders the particles in a spatial order. This allows them to greatly reduce the memory references, to run in directly indexed vector mode, and to employ domain decomposition to achieve parallelization. The field model solves pre-maxwell equations by interatively implicit methods. The OSOP (Ordered Storage Ordered Processing) version of HOPS keeps the particle tables ordered by rebuilding them after each particle pushing phase. Alternatively, the RSOP (Random Storage Ordered Processing) version keeps a table of pointers ordered by rebuilding them. Although OSOP is somewhat faster than RSOP in tests on vector-parallel machines, it is not clear this advantage will carry over to massively parallel computers

  11. Coronal mass ejection hits mercury: A.I.K.E.F. hybrid-code results compared to MESSENGER data

    Science.gov (United States)

    Exner, W.; Heyner, D.; Liuzzo, L.; Motschmann, U.; Shiota, D.; Kusano, K.; Shibayama, T.

    2018-04-01

    Mercury is the closest orbiting planet around the sun and is therefore embedded in an intensive and highly varying solar wind. In-situ data from the MESSENGER spacecraft of the plasma environment near Mercury indicates that a coronal mass ejection (CME) passed the planet on 23 November 2011 over the span of the 12 h MESSENGER orbit. Slavin et al. (2014) derived the upstream parameters of the solar wind at the time of that orbit, and were able to explain the observed MESSENGER data in the cusp and magnetopause segments of MESSENGER's trajectory. These upstream parameters will be used for our first simulation run. We use the hybrid code A.I.K.E.F. which treats ions as individual particles and electrons as a mass-less fluid, to conduct hybrid simulations of Mercury's magnetospheric response to the impact of the CME on ion gyro time scales. Results from the simulation are in agreement with magnetic field measurements from the inner day-side magnetosphere and the bow-shock region. However, at the planet's nightside, Mercury's plasma environment seemed to be governed by different solar wind conditions, in conclusion, Mercury's interaction with the CME is not sufficiently describable by only one set of upstream parameters. Therefore, to simulate the magnetospheric response while MESSENGER was located in the tail region, we use parameters obtained from the MHD solar wind simulation code SUSANOO (Shiota et al. (2014)) for our second simulation run. The parameters of the SUSANOO model achieve a good agreement of the data concerning the plasma tail crossing and the night-side approach to Mercury. However, the polar and closest approach are hardly described by both upstream parameters, namely, neither upstream dataset is able to reproduce the MESSENGER crossing of Mercury's magnetospheric cusp. We conclude that the respective CME was too variable on the timescale of the MESSENGER orbit to be described by only two sets of upstream conditions. Our results suggest locally strong

  12. Development of a PET/Cerenkov-light hybrid imaging system

    International Nuclear Information System (INIS)

    Yamamoto, Seiichi; Hamamura, Fuka; Kato, Katsuhiko; Ogata, Yoshimune; Watabe, Tadashi; Ikeda, Hayato; Kanai, Yasukazu; Hatazawa, Jun; Watabe, Hiroshi

    2014-01-01

    Purpose: Cerenkov-light imaging is a new molecular imaging technology that detects visible photons from high-speed electrons using a high sensitivity optical camera. However, the merit of Cerenkov-light imaging remains unclear. If a PET/Cerenkov-light hybrid imaging system were developed, the merit of Cerenkov-light imaging would be clarified by directly comparing these two imaging modalities. Methods: The authors developed and tested a PET/Cerenkov-light hybrid imaging system that consists of a dual-head PET system, a reflection mirror located above the subject, and a high sensitivity charge coupled device (CCD) camera. The authors installed these systems inside a black box for imaging the Cerenkov-light. The dual-head PET system employed a 1.2 × 1.2 × 10 mm 3 GSO arranged in a 33 × 33 matrix that was optically coupled to a position sensitive photomultiplier tube to form a GSO block detector. The authors arranged two GSO block detectors 10 cm apart and positioned the subject between them. The Cerenkov-light above the subject is reflected by the mirror and changes its direction to the side of the PET system and is imaged by the high sensitivity CCD camera. Results: The dual-head PET system had a spatial resolution of ∼1.2 mm FWHM and sensitivity of ∼0.31% at the center of the FOV. The Cerenkov-light imaging system's spatial resolution was ∼275μm for a 22 Na point source. Using the combined PET/Cerenkov-light hybrid imaging system, the authors successfully obtained fused images from simultaneously acquired images. The image distributions are sometimes different due to the light transmission and absorption in the body of the subject in the Cerenkov-light images. In simultaneous imaging of rat, the authors found that 18 F-FDG accumulation was observed mainly in the Harderian gland on the PET image, while the distribution of Cerenkov-light was observed in the eyes. Conclusions: The authors conclude that their developed PET/Cerenkov-light hybrid imaging

  13. Reactor Systems Technology Division code development and configuration/quality control procedures

    International Nuclear Information System (INIS)

    Johnson, E.C.

    1985-06-01

    Procedures are prescribed for executing a code development task and implementing the resulting coding in an official version of a computer code. The responsibilities of the project manager, development staff members, and the Code Configuration/Quality Control Group are defined. Examples of forms, logs, computer job control language, and suggested outlines for reports associated with software production and implementation are included in Appendix A. 1 raf., 2 figs

  14. Development of ASME Code Section 11 visual examination requirements

    International Nuclear Information System (INIS)

    Cook, J.F.

    1990-01-01

    Section XI of the American Society for Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) defines three types of nondestructive examinations, visual, surface, and volumetric. Visual examination is important since it is the primary examination method for many safety-related components and systems and is also used as a backup examination for the components and systems which receive surface or volumetric examinations. Recent activity in the Section XI Code organization to improve the rules for visual examinations is reviewed and the technical basis for the new rules, which cover illumination, vision acuity, and performance demonstration, is explained

  15. Development of a hybrid chemical/mechanical heat pump

    Science.gov (United States)

    Grzyll, Lawrence R.; Silvestri, John J.; Scaringe, Robert P.

    1991-01-01

    The authors present the current development status of a hybrid chemical/mechanical heat pump for low-lift applications. The heat pump provides electronics cooling by evaporating a pure refrigerant from an absorbent/refrigerant mixture in a generator/cold plate. The current development focused on evaluation of absorbent/refrigerant pairs, corrosion testing, pump and compressor design, and electronic cold plate design. Two cycle configurations were considered. The first configuration utilized a standard mechanical compressor and pump. The second cycle configuration investigated pumps and compressors with non-moving parts. An innovative generator/cold plate design is also presented. The development to date shows that this cycle has about the same performance as standard vapor compression heat pumps with standard refrigerants but may have some performance and reliability advantages over vapor compression heat pumps.

  16. The role of the uncertainty in code development

    Energy Technology Data Exchange (ETDEWEB)

    Barre, F. [CEA-Grenoble (France)

    1997-07-01

    From a general point of view, all the results of a calculation should be given with their uncertainty. It is of most importance in nuclear safety where sizing of the safety systems, therefore protection of the population and the environment essentially depends on the calculation results. Until these last years, the safety analysis was performed with conservative tools. Two types of critics can be made. Firstly, conservative margins can be too large and it may be possible to reduce the cost of the plant or its operation with a best estimate approach. Secondly, some of the conservative hypotheses may not really conservative in the full range of physical events which can occur during an accident. Simpson gives an interesting example: in some cases, the majoration of the residual power during a small break LOCA can lead to an overprediction of the swell level and thus of an overprediction of the core cooling, which is opposite to a conservative prediction. A last question is: does the accumulation of conservative hypotheses for a problem always give a conservative result? The two phase flow physics, mainly dealing with situation of mechanical and thermal non-equilibrium, is too much complicated to answer these questions with a simple engineer judgement. The objective of this paper is to make a review of the quantification of the uncertainties which can be made during code development and validation.

  17. The role of the uncertainty in code development

    International Nuclear Information System (INIS)

    Barre, F.

    1997-01-01

    From a general point of view, all the results of a calculation should be given with their uncertainty. It is of most importance in nuclear safety where sizing of the safety systems, therefore protection of the population and the environment essentially depends on the calculation results. Until these last years, the safety analysis was performed with conservative tools. Two types of critics can be made. Firstly, conservative margins can be too large and it may be possible to reduce the cost of the plant or its operation with a best estimate approach. Secondly, some of the conservative hypotheses may not really conservative in the full range of physical events which can occur during an accident. Simpson gives an interesting example: in some cases, the majoration of the residual power during a small break LOCA can lead to an overprediction of the swell level and thus of an overprediction of the core cooling, which is opposite to a conservative prediction. A last question is: does the accumulation of conservative hypotheses for a problem always give a conservative result? The two phase flow physics, mainly dealing with situation of mechanical and thermal non-equilibrium, is too much complicated to answer these questions with a simple engineer judgement. The objective of this paper is to make a review of the quantification of the uncertainties which can be made during code development and validation

  18. Development of a hybrid pneumatic-power vehicle

    International Nuclear Information System (INIS)

    Huang, K. David; Tzeng, S.-C.

    2005-01-01

    Many complex technologies have been developed and applied to improve the energy efficiency and exhaust emission of an engine under different driving conditions. The overall thermal efficiency of an internal-combustion engine, however, can be maintained at only about 20-30%, with aggravated problems in the design and development, such as overall difficulty, excessive time consumption or excessively high cost. For electric cars, there is still no major technological breakthrough for the rapid recharging of a large capacity battery and detection of remaining power in it. Although all currently available hybrid-power engines are able to lower the amount of exhaust emissions and the fuel consumption of the engine, they are still unable to achieve a stable and optimal running condition immediately after ignition; hence the engine's thermal-efficiency remains low. To solve the aforementioned problems, an innovative concept - a hybrid pneumatic power-system (HPPS), which stores 'flow work' instead of storing electrochemical energy of the battery - is introduced. This innovative power system not only ensures that the internal-combustion ensures optimally but also recycles the exhaust flow to propel the vehicle. The optimization of the internal-combustion and recycling of the exhaust energy can increase the vehicle's efficiency from an original 15% to 33%, an overall increase of 18%

  19. Process algebras for hybrid systems : comparison and development

    NARCIS (Netherlands)

    Khadim, U.

    2008-01-01

    Our research is about formal speci¯cation and analysis of hybrid systems. The formalism used is process algebra. Hybrid systems are systems that exhibit both discrete and continuous behaviour. An example of a hybrid system is a digital controller controlling a physical device such as present in

  20. Development of computing code system for level 3 PSA

    International Nuclear Information System (INIS)

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

    1997-07-01

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

  1. Development of computing code system for level 3 PSA

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  2. PCCS model development for SBWR using the CONTAIN code

    International Nuclear Information System (INIS)

    Tills, J.; Murata, K.K.; Washington, K.E.

    1994-01-01

    The General Electric Simplified Boiling Water Reactor (SBWR) employs a passive containment cooling system (PCCS) to maintain long-term containment gas pressure and temperature below design limits during accidents. This system consists of a steam supply line that connects the upper portion of the drywell with a vertical shell-and-tube single pass heat exchanger located in an open water pool outside of the containment safety envelope. The heat exchanger tube outlet is connected to a vent line that is submerged below the suppression pool surface but above the main suppression pool horizontal vents. Steam generated in the post-shutdown period flows into the heat exchanger tubes as the result of suction and/or a low pressure differential between the drywell and suppression chamber. Operation of the PCCS is complicated by the presence of noncondensables in the flow stream. Build-up of noncondensables in the exchanger and vent line for the periods when the vent is not cleared causes a reduction in the exchanger heat removal capacity. As flow to the exchanger is reduced due to the noncondensable gas build-up, the drywell pressure increases until the vent line is cleared and the noncondensables are purged into the suppression chamber, restoring the heat removal capability of the PCCS. This paper reports on progress made in modeling SBWR containment loads using the CONTAIN code. As a central part of this effort, a PCCS model development effort has recently been undertaken to implement an appropriate model in CONTAIN. The CONTAIN PCCS modeling approach is discussed and validated. A full SBWR containment input deck has also been developed for CONTAIN. The plant response to a postulated design basis accident (DBA) has been calculated with the CONTAIN PCCS model and plant deck, and the preliminary results are discussed

  3. On the performance of hybrid-ARQ with incremental redundancy and with code combining over relay channels

    KAUST Repository

    Chelli, Ali

    2013-08-01

    In this paper, we consider a relay network consisting of a source, a relay, and a destination. The source transmits a message to the destination using hybrid automatic repeat request (HARQ). The relay overhears the transmitted messages over the different HARQ rounds and tries to decode the data packet. In case of successful decoding at the relay, both the relay and the source cooperate to transmit the message to the destination. The channel realizations are independent for different HARQ rounds. We assume that the transmitter has no channel state information (CSI). Under such conditions, power and rate adaptation are not possible. To overcome this problem, HARQ allows the implicit adaptation of the transmission rate to the channel conditions by the use of feedback. There are two major HARQ techniques, namely HARQ with incremental redundancy (IR) and HARQ with code combining (CC). We investigate the performance of HARQ-IR and HARQ-CC over a relay channel from an information theoretic perspective. Analytical expressions are derived for the information outage probability, the average number of transmissions, and the average transmission rate. We illustrate through our investigation the benefit of relaying. We also compare the performance of HARQ-IR and HARQ-CC and show that HARQ-IR outperforms HARQ-CC. © 2013 IEEE.

  4. Recent developments in seismic analysis in the code Aster; Les developpements recents en analyse sismique dans le code aster

    Energy Technology Data Exchange (ETDEWEB)

    Guihot, P.; Devesa, G.; Dumond, A.; Panet, M.; Waeckel, F.

    1996-12-31

    Progress in the field of seismic qualification and design methods made these last few years allows physical phenomena actually in play to be better considered, while cutting down the conservatism associated with some simplified design methods. So following the change in methods and developing the most advantageous ones among them contributes to the process of the seismic margins assessment and the preparation of new design tools for future series. In this paper, the main developments and improvements in methods which have been made these last two years in the Code Aster, in order to improve seismic calculation methods and seismic margin assessment are presented. The first development relates to making the MISS3D soil structure interaction code available, thanks to an interface made with the Code Aster. The second relates to the possibility of making modal basis time calculations on multi-supported structures by considering local non linearities like impact, friction or squeeze fluid forces. Recent developments in random dynamics and postprocessing devoted to earthquake designs are then mentioned. Three applications of these developments are then ut forward. The first application relates to a test case for soil structure interaction design using MISS3D-Aster coupling. The second is a test case for a multi-supported structure. The last application, more for manufacturing, refers to seismic qualification of Main Live Steam stop valves. First results of the independent validation of the Code Aster seismic design functionalities, which provide and improve the quality of software, are also recalled. (authors). 11 refs.

  5. Development and application of a deterministic-realistic hybrid methodology for LOCA licensing analysis

    International Nuclear Information System (INIS)

    Liang, Thomas K.S.; Chou, Ling-Yao; Zhang, Zhongwei; Hsueh, Hsiang-Yu; Lee, Min

    2011-01-01

    Highlights: → A new LOCA licensing methodology (DRHM, deterministic-realistic hybrid methodology) was developed. → DRHM involves conservative Appendix K physical models and statistical treatment of plant status uncertainties. → DRHM can generate 50-100 K PCT margin as compared to a traditional Appendix K methodology. - Abstract: It is well recognized that a realistic LOCA analysis with uncertainty quantification can generate greater safety margin as compared with classical conservative LOCA analysis using Appendix K evaluation models. The associated margin can be more than 200 K. To quantify uncertainty in BELOCA analysis, generally there are two kinds of uncertainties required to be identified and quantified, which involve model uncertainties and plant status uncertainties. Particularly, it will take huge effort to systematically quantify individual model uncertainty of a best estimate LOCA code, such as RELAP5 and TRAC. Instead of applying a full ranged BELOCA methodology to cover both model and plant status uncertainties, a deterministic-realistic hybrid methodology (DRHM) was developed to support LOCA licensing analysis. Regarding the DRHM methodology, Appendix K deterministic evaluation models are adopted to ensure model conservatism, while CSAU methodology is applied to quantify the effect of plant status uncertainty on PCT calculation. Generally, DRHM methodology can generate about 80-100 K margin on PCT as compared to Appendix K bounding state LOCA analysis.

  6. Development of a multi-mode hybrid electric bus

    Energy Technology Data Exchange (ETDEWEB)

    Shemmans, M.J. [Overland Custom Coach, Thorndale, ON (Canada); Bland, C. [BET Services Inc., Mississauga, ON (Canada)

    2004-04-01

    This paper describes the development of an energy efficient, low floor, 28 foot hybrid electric bus for use as an airport shuttle bus or other specialized transit operations. A multi-mode concept was also adopted to include the capability of operating in battery-only drive, engine-only drive or a range of hybrid electric drive modes. The electric drivetrain was powered by a battery pack or a combination of a battery pack and an internal combustion engine-powered electric generator. The participating companies in this project include Overland Custom Coach, BET Services Inc., Siemens and Transport Canada. The technical feasibility study was described with reference to duty cycles, performance issues, vehicle weight, mechanical drive issues, brakes, suspension, powertrain cooling, heating, ventilation, electrical system, batteries and control system. The commercial feasibility was also described in terms of capital and operating costs. Results of the prototype tests validate the possibilities of zero or reduced emission transit in real world applications. 25 tabs., 32 figs.

  7. Development and characterisation of hybrid polysaccharide membranes for dehydration processes.

    Science.gov (United States)

    Meireles, Inês T; Huertas, Rosa M; Torres, Cristiana A V; Coelhoso, Isabel M; Crespo, João G

    2018-07-01

    The purpose of this work is the development and characterisation of new hybrid polysaccharide (FucoPol) membranes. These membranes were prepared by incorporation of a SiO 2 network homogeneously dispersed by using a sol-gel method with GPTMS as a crosslinker silica precursor. They were further crosslinked with CaCl 2 for reinforcement of mechanical properties and improvement of their permeation performance. They were characterised in terms of their structural, mechanical and thermal properties. They presented a dense and homogeneous structure, resistant to deformation, with a Tg of 43 °C and a thermal decomposition between 240 and 251 °C. The hybrid FucoPol membranes were tested for ethanol dehydration by pervaporation and also for nitrogen dehydration. They exhibited high water selectivity values, similar to PERVAP ® 4101, however they lost their stability when exposed to solutions of 10.0 wt.% water in ethanol. In contrast, these membranes were stable when applied in N 2 dehydration, leading to reproducible performance and very high water selectivities. Copyright © 2018 Elsevier Ltd. All rights reserved.

  8. Developing and modifying behavioral coding schemes in pediatric psychology: a practical guide.

    Science.gov (United States)

    Chorney, Jill MacLaren; McMurtry, C Meghan; Chambers, Christine T; Bakeman, Roger

    2015-01-01

    To provide a concise and practical guide to the development, modification, and use of behavioral coding schemes for observational data in pediatric psychology. This article provides a review of relevant literature and experience in developing and refining behavioral coding schemes. A step-by-step guide to developing and/or modifying behavioral coding schemes is provided. Major steps include refining a research question, developing or refining the coding manual, piloting and refining the coding manual, and implementing the coding scheme. Major tasks within each step are discussed, and pediatric psychology examples are provided throughout. Behavioral coding can be a complex and time-intensive process, but the approach is invaluable in allowing researchers to address clinically relevant research questions in ways that would not otherwise be possible. © The Author 2014. Published by Oxford University Press on behalf of the Society of Pediatric Psychology. All rights reserved. For permissions, please e-mail: journals.permissions@oup.com.

  9. Development of severe accident analysis code - Development of a finite element code for lower head failure analysis

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hoon; Lee, Choong Ho; Choi, Tae Hoon; Kim, Hyun Sup; Kim, Se Ho; Kang, Woo Jong; Seo, Chong Kwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-08-01

    The study concerns the development of analysis models and computer codes for lower head failure analysis when a severe accident occurs in a nuclear reactor system. Although the lower head failure modes consists of several failure modes, the study this year was focused on the global rupture with the collapse pressure and mode by limit analysis and elastic deformation. The behavior of molten core causes elevation of temperature in the reactor vessel wall and deterioration of load-carrying capacity of a reactor vessel. The behavior of molten core and the heat transfer modes were, therefore, postulated in several types and the temperature distributions according to the assumed heat flux modes were calculated. The collapse pressure of a nuclear reactor lower head decreases rapidly with elevation of temperature as time passes. The calculation shows the safety of a nuclear reactor is enhanced with the lager collapse pressure when the hot spot is located far from the pole. 42 refs., 2 tabs., 31 figs. (author)

  10. Development of Teaching Materials for a Physical Chemistry Experiment Using the QR Code

    OpenAIRE

    吉村, 忠与志

    2008-01-01

    The development of teaching materials with the QR code was attempted in an educational environment using a mobile telephone. The QR code is not sufficiently utilized in education, and the current study is one of the first in the field. The QR code is encrypted. However, the QR code can be deciphered by mobile telephones, thus enabling the expression of text in a small space.Contents of "Physical Chemistry Experiment" which are available on the Internet are briefly summarized and simplified. T...

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

    International Nuclear Information System (INIS)

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

    1985-03-01

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

  12. Development of Hybrid Product Breakdown Structure for NASA Ground Systems

    Science.gov (United States)

    Monaghan, Mark W.; Henry, Robert J.

    2013-01-01

    The Product Breakdown Structure is traditionally a method of identification of the products of a project in a tree structure. It is a tool used to assess, plan, document, and display the equipment requirements for a project. It is part of a product based planning technique, and attempts to break down all components of a project in as much detail as possible, so that nothing is overlooked. The PBS for ground systems at the Kennedy Space Center is being developed to encompass the traditional requirements including the alignment of facility, systems, and components to the organizational hierarchy. The Ground Operations Product Breakdown Structure is a hybrid in nature in that some aspects of a work breakdown structure will be incorporated and merged with the Architecture Concept of Operations, Master Subsystem List, customer interface, and assigned management responsibility. The Ground Operations Product Breakdown Structure needs to be able to identify the flexibility of support differing customers (internal and external) usage of ground support equipment within the Kennedy Space Center launch and processing complex. The development of the Product Breakdown Structure is an iterative activity Initially documenting the organization hierarchy structure and relationships. The Product Breakdown Structure identifies the linkage between the customer program requirements, allocation of system resources, development of design goals, and identification logistics products. As the Product Breakdown Structure progresses the incorporation of the results of requirement planning for the customer occurs identifying facility needs and systems. The mature Product Breakdown Structure is baselined with a hierarchical drawing, the Product Breakdown Structure database, and an associated document identifying the verification of the data through the life cycle of the program/product line. This paper will document, demonstrate, and identify key aspects of the life cycle of a Hybrid Product

  13. SWAAM-code development and verification and application to steam generator designs

    International Nuclear Information System (INIS)

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

    1990-01-01

    This paper describes the family of SWAAM codes which were 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 the resulting thermal and hydraulic effects on the steam generator and the intermediate heat transport system (IHTS). The paper discusses the theoretical foundations and numerical treatments on which the codes are based, followed by a description of code capabilities and limitations, verification of the codes and applications to steam generator and IHTS designs. 25 refs., 14 figs

  14. The development of the code package PERMAK--3D//SC--1

    International Nuclear Information System (INIS)

    Bolobov, P. A.; Oleksuk, D. A.

    2011-01-01

    Code package PERMAK-3D//SC-1 was developed for performing pin-by-pin coupled neutronic and thermal hydraulic calculation of the core fragment of seven fuel assemblies and was designed on the basis of 3D multigroup pin-by-pin code PERMAK-3D and 3D (subchannel) thermal hydraulic code SC-1 The code package predicts axial and radial pin-by-pin power distribution and coolant parameters in stimulated region (enthalpies,, velocities,, void fractions,, boiling and DNBR margins).. The report describes some new steps in code package development. Some PERMAK-3D//SC-1 outcomes of WWER calculations are presented in the report. (Authors)

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  16. Large-Signal Code TESLA: Current Status and Recent Development

    National Research Council Canada - National Science Library

    Chernyavskiy, Igor A; Vlasov, Alexander N; Cooke, Simon J; Abe, David K; Levush, Baruch; Antonsen, Jr., Thomas M; Nguyen, Khanh T

    2008-01-01

    .... One such tool is the large-signal code TESLA, which was successfully applied for the modeling of single-beam and multiple-beam klystron devices at the Naval Research Laboratory and which is now used by number of U.S. companies...

  17. Development of Scientific Simulation 3D Full Wave ICRF Code for Stellarators and Heating/CD Scenarios Development

    Energy Technology Data Exchange (ETDEWEB)

    Vdovin V.L.

    2005-08-15

    In this report we describe theory and 3D full wave code description for the wave excitation, propagation and absorption in 3-dimensional (3D) stellarator equilibrium high beta plasma in ion cyclotron frequency range (ICRF). This theory forms a basis for a 3D code creation, urgently needed for the ICRF heating scenarios development for the operated LHD, constructed W7-X, NCSX and projected CSX3 stellarators, as well for re evaluation of ICRF scenarios in operated tokamaks and in the ITER . The theory solves the 3D Maxwell-Vlasov antenna-plasma-conducting shell boundary value problem in the non-orthogonal flux coordinates ({Psi}, {theta}, {var_phi}), {Psi} being magnetic flux function, {theta} and {var_phi} being the poloidal and toroidal angles, respectively. All basic physics, like wave refraction, reflection and diffraction are self consistently included, along with the fundamental ion and ion minority cyclotron resonances, two ion hybrid resonance, electron Landau and TTMP absorption. Antenna reactive impedance and loading resistance are also calculated and urgently needed for an antenna -generator matching. This is accomplished in a real confining magnetic field being varying in a plasma major radius direction, in toroidal and poloidal directions, through making use of the hot dense plasma wave induced currents with account to the finite Larmor radius effects. We expand the solution in Fourier series over the toroidal ({var_phi}) and poloidal ({theta}) angles and solve resulting ordinary differential equations in a radial like {Psi}-coordinate by finite difference method. The constructed discretization scheme is divergent-free one, thus retaining the basic properties of original equations. The Fourier expansion over the angle coordinates has given to us the possibility to correctly construct the ''parallel'' wave number k{sub //}, and thereby to correctly describe the ICRF waves absorption by a hot plasma. The toroidal harmonics are tightly

  18. Application of software quality assurance to a specific scientific code development task

    International Nuclear Information System (INIS)

    Dronkers, J.J.

    1986-03-01

    This paper describes an application of software quality assurance to a specific scientific code development program. The software quality assurance program consists of three major components: administrative control, configuration management, and user documentation. The program attempts to be consistent with existing local traditions of scientific code development while at the same time providing a controlled process of development

  19. Simulation of Alfvén eigenmode bursts using a hybrid code for nonlinear magnetohydrodynamics and energetic particles

    Science.gov (United States)

    Todo, Y.; Berk, H. L.; Breizman, B. N.

    2012-03-01

    A hybrid simulation code for nonlinear magnetohydrodynamics (MHD) and energetic-particle dynamics has been extended to simulate recurrent bursts of Alfvén eigenmodes by implementing the energetic-particle source, collisions and losses. The Alfvén eigenmode bursts with synchronization of multiple modes and beam ion losses at each burst are successfully simulated with nonlinear MHD effects for the physics condition similar to a reduced simulation for a TFTR experiment (Wong et al 1991 Phys. Rev. Lett. 66 1874, Todo et al 2003 Phys. Plasmas 10 2888). It is demonstrated with a comparison between nonlinear MHD and linear MHD simulation results that the nonlinear MHD effects significantly reduce both the saturation amplitude of the Alfvén eigenmodes and the beam ion losses. Two types of time evolution are found depending on the MHD dissipation coefficients, namely viscosity, resistivity and diffusivity. The Alfvén eigenmode bursts take place for higher dissipation coefficients with roughly 10% drop in stored beam energy and the maximum amplitude of the dominant magnetic fluctuation harmonic δBm/n/B ~ 5 × 10-3 at the mode peak location inside the plasma. Quadratic dependence of beam ion loss rate on magnetic fluctuation amplitude is found for the bursting evolution in the nonlinear MHD simulation. For lower dissipation coefficients, the amplitude of the Alfvén eigenmodes is at steady levels δBm/n/B ~ 2 × 10-3 and the beam ion losses take place continuously. The beam ion pressure profiles are similar among the different dissipation coefficients, and the stored beam energy is higher for higher dissipation coefficients.

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

    International Nuclear Information System (INIS)

    Tai Asayama; Masaki Morishita; Masanori Tashimo

    2006-01-01

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

  1. Development of statistical analysis code for meteorological data (W-View)

    International Nuclear Information System (INIS)

    Tachibana, Haruo; Sekita, Tsutomu; Yamaguchi, Takenori

    2003-03-01

    A computer code (W-View: Weather View) was developed to analyze the meteorological data statistically based on 'the guideline of meteorological statistics for the safety analysis of nuclear power reactor' (Nuclear Safety Commission on January 28, 1982; revised on March 29, 2001). The code gives statistical meteorological data to assess the public dose in case of normal operation and severe accident to get the license of nuclear reactor operation. This code was revised from the original code used in a large office computer code to enable a personal computer user to analyze the meteorological data simply and conveniently and to make the statistical data tables and figures of meteorology. (author)

  2. Development of a hybrid cooler; Udvikling af hybridkoeler

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, P.; Toftegaard, R.; Weinkauff Kristoffersen, J. [Teknologisk Institut, Aarhus (Denmark); Juel Skovrup, M. [IPU, Kgs. Lyngby (Denmark); Ibsen, C. [VP Industries, Lem (Denmark)

    2013-04-15

    The project aims to develop a hybrid cooler which acts as a dry cooler in the winter and as cooling tower in summer. Energy consumption for cooling systems with a dry cooler and a cooling tower, respectively, is comparable in the winter months. This phase 1 of the project shows that improvements of 50-100% on the performance of a hybrid cooler can be achieved as compared to a dry cooler. The improvement is achieved by humidifying the air with recirculated water through nozzles so that the air temperature decreases from the dry temperature to the wet temperature, and that the dry cooler surface is humidified with a film of water, which increases the heat transfer coefficient considerably compared to a dry surface. The experiments showed that a humidifier system cannot be used without further action. At face velocities less than 5 m/s the humidification does not yield any improvement, and in some cases the heat transfer in a standard dry cooler is decreased. This is due to entrainment of not fully vaporized droplets which are deposited between the dry cooler fins and form bridges that block parts of the cooler. By modifying the surface characteristics with a coating, it will be possible to drain the water away so that no bridges are formed. The company Accoat, which makes special surfaces, will therefore be associated to phase 2 of the project. Another aspect that was evident in the tests, is the formation of biofilm on the heat exchanger surface, which can reduce performance by up to 25%. Biofilm can be prevented by treating the feed water, and therefore Danish Clean Water A/S associated to phase 2 of the project, as they produce water purification systems for biofouling decomposition. (LN)

  3. DEVELOPMENT OF SALES APPLICATION OF PREPAID ELECTRICITY VOUCHER BASED ON ANFROID PLATFORM USING QUICK RESPONSE CODE (QR CODE

    Directory of Open Access Journals (Sweden)

    Ricky Akbar

    2017-09-01

    Full Text Available Perusahaan Listrik Negara (PLN has implemented a smart electricity system or prepaid electricity. The customers pay the electricity voucher first before use the electricity. The token contained in electricity voucher that has been purchased by the customer is inserted into the Meter Prabayar (MPB installed in the location of customers. When a customer purchases a voucher, it will get a receipt that contains all of the customer's identity and the 20-digit of voucher code (token to be entered into MPB as a substitute for electrical energy credit. Receipts obtained by the customer is certainly vulnerable to loss, or hijacked by unresponsible parties. In this study, authors designed and develop an android based application by utilizing QR code technology as a replacement for the receipt of prepaid electricity credit which contains the identity of the customer and the 20-digit voucher code. The application is developed by implemented waterfall methodology. The implementation process of the waterfall methods used, are (1 analysis of functional requirement of the system by conducting a preliminary study and data collection based on field studies and literature, (2 system design by using UML diagrams and Business Process Model Notation (BPMN and Entity Relationship diagram (ERD, (3 design implementation by using OOP (Object Oriented programming technique. Web application is developed by using laravel PHP framework and database MySQL while mobile application is developed by using B4A (4 developed system is tested by using blackbox method testing. Final result of this research is a Web and mobile applications for the sale of electricityvoucher by QR Code technology.

  4. Development of thermal hydraulic analysis code for IHX of FBR

    International Nuclear Information System (INIS)

    Kumagai, Hiromichi; Naohara, Nobuyuki

    1991-01-01

    In order to obtain flow resistance correlations for thermal-hydrauric analysis code concerned with an intermediate heat exchanger (IHX) of FBR, the hydraulic experiment by air was carried out through a bundle of tubes arranged in an in-line and staggard fashion. The main results are summarized as follows. (1) On pressure loss per unit length of a tube bundle, which is densely a regular triangle arrangement, the in-line fashion is almost the same as the staggard one. (2) In case of 30deg sector model for IHX tube bundle, pressure loss is 1/3 in comparison with the in-line or staggard arrangement. (3) By this experimental data, flow resistance correlations for thermalhydrauric analysis code are obtained. (author)

  5. Recent developments for the HEADTAIL code: updating and benchmarks

    CERN Document Server

    Quatraro, D; Salvant, B

    2010-01-01

    The HEADTAIL code models the evolution of a single bunch interacting with a localized impedance source or an electron cloud, optionally including space charge. The newest version of HEADTAIL relies on a more detailed optical model of the machine taken from MAD-X and is more flexible in handling and distributing the interaction and observation points along the simulated machine. In addition, the option of the interaction with the wake field of specific accelerator components has been added, such that the user can choose to load dipolar and quadrupolar components of the wake from the impedance database ZBASE. The case of a single LHC-type bunch interacting with the realistic distribution of the kicker wake fields inside the SPS has been successfully compared with a single integrated beta-weighted kick per turn. The current version of the code also contains a new module for the longitudinal dynamics to calculate the evolution of a bunch inside an accelerating bucket.

  6. Code Development for Control Design Applications: Phase I: Structural Modeling

    International Nuclear Information System (INIS)

    Bir, G. S.; Robinson, M.

    1998-01-01

    The design of integrated controls for a complex system like a wind turbine relies on a system model in an explicit format, e.g., state-space format. Current wind turbine codes focus on turbine simulation and not on system characterization, which is desired for controls design as well as applications like operating turbine model analysis, optimal design, and aeroelastic stability analysis. This paper reviews structural modeling that comprises three major steps: formation of component equations, assembly into system equations, and linearization

  7. Development of 'SKYSHINE-CG' code. A line-beam method code equipped with combinatorial geometry routine

    Energy Technology Data Exchange (ETDEWEB)

    Nakagawa, Takahiro; Ochiai, Katsuharu [Plant and System Planning Department, Toshiba Corporation, Yokohama, Kanagawa (Japan); Uematsu, Mikio; Hayashida, Yoshihisa [Department of Nuclear Engineering, Toshiba Engineering Corporation, Yokohama, Kanagawa (Japan)

    2000-03-01

    A boiling water reactor (BWR) plant has a single loop coolant system, in which main steam generated in the reactor core proceeds directly into turbines. Consequently, radioactive {sup 16}N (6.2 MeV photon emitter) contained in the steam contributes to gamma-ray skyshine dose in the vicinity of the BWR plant. The skyshine dose analysis is generally performed with the line-beam method code SKYSHINE, in which calculational geometry consists of a rectangular turbine building and a set of isotropic point sources corresponding to an actual distribution of {sup 16}N sources. For the purpose of upgrading calculational accuracy, the SKYSHINE-CG code has been developed by incorporating the combinatorial geometry (CG) routine into the SKYSHINE code, so that shielding effect of in-building equipment can be properly considered using a three-dimensional model composed of boxes, cylinders, spheres, etc. Skyshine dose rate around a 500 MWe BWR plant was calculated with both SKYSHINE and SKYSHINE-CG codes, and the calculated results were compared with measured data obtained with a NaI(Tl) scintillation detector. The C/E values for SKYSHINE-CG calculation were scattered around 4.0, whereas the ones for SKYSHINE calculation were as large as 6.0. Calculational error was found to be reduced by adopting three-dimensional model based on the combinatorial geometry method. (author)

  8. Development of an object oriented lattice QCD code ''Bridge++''

    International Nuclear Information System (INIS)

    Ueda, S; Aoki, S; Aoyama, T; Kanaya, K; Taniguchi, Y; Matsufuru, H; Motoki, S; Namekawa, Y; Nemura, H; Ukita, N

    2014-01-01

    We are developing a new lattice QCD code set ''Bridge++'' aiming at extensible, readable, and portable workbench for QCD simulations, while keeping a high performance at the same time. Bridge++ covers conventional lattice actions and numerical algorithms. The code set is constructed in C++ with an object oriented programming. In this paper we describe fundamental ingredients of the code and the current status of development

  9. ASME Section XI trends in developing nuclear codes and standards

    International Nuclear Information System (INIS)

    Hedden, O.F.

    1995-01-01

    When the author began working on nuclear power many years ago, he knew that perfection was the only acceptable technical standard. Unfortunately, this became an obsession with perfection that has had unfavorable consequences in some of the non-technical areas of work in ASME nuclear power Codes and Standards. However, the economic problems of the nuclear power industry now demand a more pragmatic approach if the industry is to continue. Not only does each item considered for action need to be evaluated to criteria that may in some cases be less than perfection, but one needs to consider whether it contributes tangibly to either safety or to reduction in technical or administrative burden. These should be the governing, criteria. The introduction of risk-based inspection methodologies will certainly be an important element in doing this successfully. One needs to consider these criteria collectively, as one discusses each item at the committee level, and individually, as one votes on each item. In the past, the author has been concerned that the industry was not acting quickly enough in taking advantage of opportunities offered by the Code to increase safety or to reduce cost. While he still has some concern, he thinks communication channels have been greatly improved. Now he is becoming more concerned with both the collective and individual actions that delay beneficial changes. The second part of the author's talk has to do with the relevance of the code committees in the nuclear power industry regulatory process

  10. Development of a parallelization strategy for the VARIANT code

    International Nuclear Information System (INIS)

    Hanebutte, U.R.; Khalil, H.S.; Palmiotti, G.; Tatsumi, M.

    1996-01-01

    The VARIANT code solves the multigroup steady-state neutron diffusion and transport equation in three-dimensional Cartesian and hexagonal geometries using the variational nodal method. VARIANT consists of four major parts that must be executed sequentially: input handling, calculation of response matrices, solution algorithm (i.e. inner-outer iteration), and output of results. The objective of the parallelization effort was to reduce the overall computing time by distributing the work of the two computationally intensive (sequential) tasks, the coupling coefficient calculation and the iterative solver, equally among a group of processors. This report describes the code's calculations and gives performance results on one of the benchmark problems used to test the code. The performance analysis in the IBM SPx system shows good efficiency for well-load-balanced programs. Even for relatively small problem sizes, respectable efficiencies are seen for the SPx. An extension to achieve a higher degree of parallelism will be addressed in future work. 7 refs., 1 tab

  11. Development of Ultrasonic Pulse Compression Using Golay Codes

    International Nuclear Information System (INIS)

    Kim, Young H.; Kim, Young Gil; Jeong, Peter

    1994-01-01

    Conventional ultrasonic flaw detection system uses a large amplitude narrow pulse to excite a transducer. However, these systems are limited in pulse energy. An excessively large amplitude causes a dielectric breakage of the transducer, and an excessively long pulse causes decrease of the resolution. Using the pulse compression, a long pulse of pseudorandom signal can be used without sacrificing resolution by signal correlation. In the present work, the pulse compression technique was implemented into an ultrasonic system. Golay code was used as a pseudorandom signal in this system, since pair sum of autocorrelations has no sidelobe. The equivalent input pulse of the Golay code was derived to analyze the pulse compression system. Throughout the experiment, the pulse compression technique has demonstrated for its improved SNR(signal to noise ratio) by reducing the system's white noise. And the experimental data also indicated that the SNR enhancement was proportional to the square root of the code length used. The technique seems to perform particularly well with highly energy-absorbent materials such as polymers, plastics and rubbers

  12. A Hybrid Fuzzy Model for Lean Product Development Performance Measurement

    Science.gov (United States)

    Osezua Aikhuele, Daniel; Mohd Turan, Faiz

    2016-02-01

    In the effort for manufacturing companies to meet up with the emerging consumer demands for mass customized products, many are turning to the application of lean in their product development process, and this is gradually moving from being a competitive advantage to a necessity. However, due to lack of clear understanding of the lean performance measurements, many of these companies are unable to implement and fully integrated the lean principle into their product development process. Extensive literature shows that only few studies have focus systematically on the lean product development performance (LPDP) evaluation. In order to fill this gap, the study therefore proposed a novel hybrid model based on Fuzzy Reasoning Approach (FRA), and the extension of Fuzzy-AHP and Fuzzy-TOPSIS methods for the assessment of the LPDP. Unlike the existing methods, the model considers the importance weight of each of the decision makers (Experts) since the performance criteria/attributes are required to be rated, and these experts have different level of expertise. The rating is done using a new fuzzy Likert rating scale (membership-scale) which is designed such that it can address problems resulting from information lost/distortion due to closed-form scaling and the ordinal nature of the existing Likert scale.

  13. Developments in the Generation and Interpretation of Wire Codes (invited paper)

    International Nuclear Information System (INIS)

    Ebi, K.L.

    1999-01-01

    Three new developments in the generation and interpretation of wire codes are discussed. First, a method was developed to computer generate wire codes using data gathered from a utility database of the local distribution system and from tax assessor records. This method was used to wire code more than 250,000 residences in the greater Denver metropolitan area. There was an approximate 75% agreement with field wire coding. Other research in Denver suggests that wire codes predict some characteristics of a residence and its neighbourhood, including age, assessed value, street layout and traffic density. A third new development is the case-specular method to study the association between wire codes and childhood cancers. Recent results from applying the method to the Savitz et al and London et al studies suggest that the associations between childhood cancer and VHCC residences were strongest for residences with a backyard rather than street service drop, and for VHCC residences with LCC speculars. (author)

  14. CopenHybridDevelopment of a CO2 Neutral Hybrid Street Lighting System for the Danish Municipalities’ Illumination Classes

    DEFF Research Database (Denmark)

    Poulsen, Peter Behrensdorff; Dam-Hansen, Carsten; Thorseth, Anders

    2013-01-01

    A mathematical model has been developed for the energy system of the hybrid street lighting making it possible to simulate a given configuration (solar panel performance data, size and orientation - wind turbine performance data, projected area and height - battery data) over a year in an urban...... environment of a given configuration based on measured wind/solar/temperature data from nearby meteorological station or other relevant weather data. The simulation can show if it is possible to cover the consumption by the light source over the year. The model can be used to evaluate both commercial hybrid...

  15. Development of a code for description of sodium spray and pool fires. Pt. 1

    International Nuclear Information System (INIS)

    Alexas, A.

    1979-09-01

    In the scope of the development of a code to describe both, sodium pool- and spray fires, the well-known codes SOFIRE II and NABRAND have been compared. Regarding the program technique of both codes, the NABRAND-code seems to be the better one, though it includes some conservatisms in the modelling and in the transport coefficients used. For a realistic estimation of the consequences of large sodium fires in an LMFBR, an elimination of these conservatisms is necessary. After that it must be investigated if a combination of the modificated version of the NABRAND-code and of a spray fire-code (for example the code SPRAY) is efficient. (orig.) [de

  16. Development of M3C code for Monte Carlo reactor physics criticality calculations

    International Nuclear Information System (INIS)

    Kumar, Anek; Kannan, Umasankari; Krishanani, P.D.

    2015-06-01

    The development of Monte Carlo code (M3C) for reactor design entails use of continuous energy nuclear data and Monte Carlo simulations for each of the neutron interaction processes. BARC has started a concentrated effort for developing a new general geometry continuous energy Monte Carlo code for reactor physics calculation indigenously. The code development required a comprehensive understanding of the basic continuous energy cross section sets. The important features of this code are treatment of heterogeneous lattices by general geometry, use of point cross sections along with unionized energy grid approach, thermal scattering model for low energy treatment, capability of handling the microscopic fuel particles dispersed randomly. The capability of handling the randomly dispersed microscopic fuel particles which is very useful for the modeling of High-Temperature Gas-Cooled reactor fuels which are composed of thousands of microscopic fuel particle (TRISO fuel particle), randomly dispersed in a graphite matrix. The Monte Carlo code for criticality calculation is a pioneering effort and has been used to study several types of lattices including cluster geometries. The code has been verified for its accuracy against more than 60 sample problems covering a wide range from simple (like spherical) to complex geometry (like PHWR lattice). Benchmark results show that the code performs quite well for the criticality calculation of the system. In this report, the current status of the code, features of the code, some of the benchmark results for the testing of the code and input preparation etc. are discussed. (author)

  17. 77 FR 17460 - Multistakeholder Process To Develop Consumer Data Privacy Codes of Conduct

    Science.gov (United States)

    2012-03-26

    ..., 2012, NTIA requested public comments on (1) which consumer data privacy issues should be the focus of.... 120214135-2203-02] RIN 0660-XA27 Multistakeholder Process To Develop Consumer Data Privacy Codes of Conduct... request for public comments on the multistakeholder process to develop consumer data privacy codes of...

  18. Drought Stress Responses of Sunflower Germplasm Developed after Wide Hybridization

    Directory of Open Access Journals (Sweden)

    Roumiana Dimova Vassilevska-Ivanova

    2016-10-01

    Full Text Available Response of sunflower germplasms viz. cultivated sunflower H. annuus and two breeding lines H. annuus x T. rotundifolia and H. annuus x V. encelioides developed after wide hybridization were used for identification of drought tolerant sunflower genotypes at the seedling growth stage. Three water stress levels of zero (control, -0.4, and -0.8 MPa were developed using polyethyleneglycol-6000 (PEG-6000. Physiological and biochemical stress determining parameters such as root and shoots length, fresh weight, antioxidant enzyme activities (superoxide dismutase (SOD, catalase (CAT, guaiacol peroxidase (GPO, ascorbate peroxidase (APX and antioxidant metabolite content (total antioxidant capacity, total phenols and total flavonoids content were compared between seedlings of all three genotypes. Results revealed that sunflower genotypes have similar responses at two osmotic potentials for shoot and root length and fresh weight. The data also showed that drought stresss could induce oxidative stress, as indicated by the increase level of ascorbate peroxidase and guaiacol peroxidase at -04 MPa in H. annuus cv 1114. Although the activity of ascorbate peroxidase and guaiacol peroxidase was differentially influenced by drought, the changes of antioxidant enzyme activities such as catalase, superoxide dismutase, guaiacol peroxidase, and ascorbate peroxidase subjected to drought stress follow a similar pattern in both breeding lines, indicating that similar defense systems might be involved in the oxidative stress injury in sunflowers. Increase in content of phenols and flavonoids were detected for all three genotypes under stress, which showed that these were major antioxidant metabolites in scavenging cellular H2O2.

  19. Lower hybrid current drive in shaped tokamaks

    International Nuclear Information System (INIS)

    Kesner, J.

    1993-01-01

    A time dependent lower hybrid current drive tokamak simulation code has been developed. This code combines the BALDUR tokamak simulation code and the Bonoli/Englade lower hybrid current drive code and permits the study of the interaction of lower hybrid current drive with neutral beam heating in shaped cross-section plasmas. The code is time dependent and includes the beam driven and bootstrap currents in addition to the current driven by the lower hybrid system. Examples of simulations are shown for the PBX-M experiment which include the effect of cross section shaping on current drive, ballooning mode stabilization by current profile control and sawtooth stabilization. A critical question in current drive calculations is the radial transport of the energetic electrons. The authors have developed a response function technique to calculate radial transport in the presence of an electric field. The consequences of the combined influences of radial diffusion and electric field acceleration are discussed

  20. Development of Monte Carlo-based pebble bed reactor fuel management code

    International Nuclear Information System (INIS)

    Setiadipura, Topan; Obara, Toru

    2014-01-01

    Highlights: • A new Monte Carlo-based fuel management code for OTTO cycle pebble bed reactor was developed. • The double-heterogeneity was modeled using statistical method in MVP-BURN code. • The code can perform analysis of equilibrium and non-equilibrium phase. • Code-to-code comparisons for Once-Through-Then-Out case were investigated. • Ability of the code to accommodate the void cavity was confirmed. - Abstract: A fuel management code for pebble bed reactors (PBRs) based on the Monte Carlo method has been developed in this study. The code, named Monte Carlo burnup analysis code for PBR (MCPBR), enables a simulation of the Once-Through-Then-Out (OTTO) cycle of a PBR from the running-in phase to the equilibrium condition. In MCPBR, a burnup calculation based on a continuous-energy Monte Carlo code, MVP-BURN, is coupled with an additional utility code to be able to simulate the OTTO cycle of PBR. MCPBR has several advantages in modeling PBRs, namely its Monte Carlo neutron transport modeling, its capability of explicitly modeling the double heterogeneity of the PBR core, and its ability to model different axial fuel speeds in the PBR core. Analysis at the equilibrium condition of the simplified PBR was used as the validation test of MCPBR. The calculation results of the code were compared with the results of diffusion-based fuel management PBR codes, namely the VSOP and PEBBED codes. Using JENDL-4.0 nuclide library, MCPBR gave a 4.15% and 3.32% lower k eff value compared to VSOP and PEBBED, respectively. While using JENDL-3.3, MCPBR gave a 2.22% and 3.11% higher k eff value compared to VSOP and PEBBED, respectively. The ability of MCPBR to analyze neutron transport in the top void of the PBR core and its effects was also confirmed

  1. Development of an analysis code for pressure wave propagation, (1)

    International Nuclear Information System (INIS)

    Tanaka, Yoshihisa; Sakano, Kosuke; Shindo, Yoshihisa

    1974-11-01

    We analyzed the propagation of the pressure-wave in the piping system of SWAT-1B rig by using SWAC-5 Code. We carried out analyses on the following parts. 1) A straight pipe 2) Branches 3) A piping system The results obtained in these analyses are as follows. 1) The present our model simulates well the straight pipe and the branch with the same diameters. 2) The present our model simulates approximately the branch with the different diameters and the piping system. (auth.)

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

    International Nuclear Information System (INIS)

    Jobson, D.A.

    1983-01-01

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

  3. Linear engine development for series hybrid electric vehicles

    Science.gov (United States)

    Toth-Nagy, Csaba

    This dissertation argues that diminishing oil reserves, concern over global climate change, and desire to improve ambient air quality all demand the development of environment-friendly personal transportation. In certain applications, series hybrid electric vehicles offer an attractive solution to reducing fuel consumption and emissions. Furthermore, linear engines are emerging as a powerplant suited to series HEV applications. In this dissertation, a linear engine/alternator was considered as the auxiliary power unit of a range extender series hybrid electric vehicle. A prototype linear engine/alternator was developed, constructed and tested at West Virginia University. The engine was a 2-stroke, 2-cylinder, dual piston, direct injection, diesel engine. Experiment on the engine was performed to study its behavior. The study variables included mass of the translator, amount of fuel injected, injection timing, load, and stroke with operating frequency and mechanical efficiency as the basis of comparison. The linear engine was analyzed in detail and a simple simulation model was constructed to compare the trends of simulation with the experimental data and to expand on the area where the experimental data were lacking. The simulation was based on a simple and analytical model, rather than a detailed and intensely numerical one. The experimental and theoretical data showed similar trends. Increasing translator mass decreased the operating frequency and increased compression ratio. Larger mass and increased compression ratio improved the ability of the engine to sustain operation and the engine was able to idle on less fuel injected into the cylinder. Increasing the stroke length caused the operating frequency to drop. Increasing fueling or decreasing the load resulted in increased operating frequency. This projects the possibility of using the operating frequency as an input for feedback control of the engine. Injection timing was varied to investigate two different

  4. Methodology, status, and plans for development and assessment of the RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, G.W.; Riemke, R.A. [Idaho National Engineering Laboratory, Idaho Falls, ID (United States)

    1997-07-01

    RELAP/MOD3 is a computer code used for the simulation of transients and accidents in light-water nuclear power plants. The objective of the program to develop and maintain RELAP5 was and is to provide the U.S. Nuclear Regulatory Commission with an independent tool for assessing reactor safety. This paper describes code requirements, models, solution scheme, language and structure, user interface validation, and documentation. The paper also describes the current and near term development program and provides an assessment of the code`s strengths and limitations.

  5. Development of LMR basic design technology - Development of 3-D multi-group nodal kinetics code for liquid metal reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myung Hyun [Kyunghee University, Seoul (Korea, Republic of)

    1996-07-01

    A development project of 3-dimensional kinetics code for ALMR has three level of works. In the first level, a multi-group, nodal kinetics code for the HEX-Z geometry has been developed. A code showed very good results for the static analysis as well as for the kinetics problems. At the second level, a core thermal-hydraulic analysis code was developed for the temperature feedback calculation in ALMR transients analysis. This code is coupled with kinetics code. A sodium property table was programmed and tested to the KAERI data and thermal feedback model was developed and coupled in code. Benchmarking of T/H calculation has been performed and showed fairly good results. At the third level of research work, reactivity feedback model for structure thermal expansion is developed and added to the code. At present, basic model was studied. However, code development in now on going. Benchmarking of this model developed can not be done because of lack of data. 31 refs., 17 tabs., 38 figs. (author)

  6. Methodology, status and plans for development and assessment of TUF and CATHENA codes

    Energy Technology Data Exchange (ETDEWEB)

    Luxat, J.C.; Liu, W.S.; Leung, R.K. [Ontario Hydro, Toronto (Canada)] [and others

    1997-07-01

    An overview is presented of the Canadian two-fluid computer codes TUF and CATHENA with specific focus on the constraints imposed during development of these codes and the areas of application for which they are intended. Additionally a process for systematic assessment of these codes is described which is part of a broader, industry based initiative for validation of computer codes used in all major disciplines of safety analysis. This is intended to provide both the licensee and the regulator in Canada with an objective basis for assessing the adequacy of codes for use in specific applications. Although focused specifically on CANDU reactors, Canadian experience in developing advanced two-fluid codes to meet wide-ranging application needs while maintaining past investment in plant modelling provides a useful contribution to international efforts in this area.

  7. Micromagnetic Code Development of Advanced Magnetic Structures Final Report CRADA No. TC-1561-98

    Energy Technology Data Exchange (ETDEWEB)

    Cerjan, Charles J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Shi, Xizeng [Read-Rite Corporation, Fremont, CA (United States)

    2017-11-09

    The specific goals of this project were to: Further develop the previously written micromagnetic code DADIMAG (DOE code release number 980017); Validate the code. The resulting code was expected to be more realistic and useful for simulations of magnetic structures of specific interest to Read-Rite programs. We also planned to further the code for use in internal LLNL programs. This project complemented LLNL CRADA TC-840-94 between LLNL and Read-Rite, which allowed for simulations of the advanced magnetic head development completed under the CRADA. TC-1561-98 was effective concurrently with LLNL non-exclusive copyright license (TL-1552-98) to Read-Rite for DADIMAG Version 2 executable code.

  8. Methodology, status and plans for development and assessment of TUF and CATHENA codes

    International Nuclear Information System (INIS)

    Luxat, J.C.; Liu, W.S.; Leung, R.K.

    1997-01-01

    An overview is presented of the Canadian two-fluid computer codes TUF and CATHENA with specific focus on the constraints imposed during development of these codes and the areas of application for which they are intended. Additionally a process for systematic assessment of these codes is described which is part of a broader, industry based initiative for validation of computer codes used in all major disciplines of safety analysis. This is intended to provide both the licensee and the regulator in Canada with an objective basis for assessing the adequacy of codes for use in specific applications. Although focused specifically on CANDU reactors, Canadian experience in developing advanced two-fluid codes to meet wide-ranging application needs while maintaining past investment in plant modelling provides a useful contribution to international efforts in this area

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

    International Nuclear Information System (INIS)

    Konstantin Mikityuk; Sandro Pelloni; Paul Coddington

    2005-01-01

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

  10. Development of thin-film Si HYBRID solar module

    Energy Technology Data Exchange (ETDEWEB)

    Nakajima, Akihiko; Gotoh, Masahiro; Sawada, Toru; Fukuda, Susumu; Yoshimi, Masashi; Yamamoto, Kenji; Nomura, Takuji [Kaneka Corporation, 2-1-1, Hieitsuji, Otsu, Shiga 520-0104 (Japan)

    2009-06-15

    The device current-voltage (I-V) characteristics of thin-film silicon stacked tandem solar modules (HYBRID modules), consisting of a hydrogenated amorphous silicon (a-Si:H) cell and a thin-film crystalline silicon solar cell ({mu}c-Si), have been investigated under various spectral irradiance distributions. The performance of the HYBRID module varied periodically in natural sunlight due to the current-limiting property of the HYBRID module and the environmental effects. The behavior based on the current-limiting property was demonstrated by the modelling of the I-V curves using the linear interpolation method for each component cell. The improvement of the performance for the HYBRID module in natural sunlight will also be discussed from the viewpoint of the device design of the component cells. (author)

  11. Sex-specific asymmetry in eye development in interspecific hybrids ...

    Indian Academy of Sciences (India)

    hybrids having D. pseudoananassae as one of the parental species. One of them was the .... that recessive epistatic X-autosomal interactions are involved as none of the ... The X chromosome of the maternal species,. D. pseudoananassae ...

  12. Towards the optimization of a gyrokinetic Particle-In-Cell (PIC) code on large-scale hybrid architectures

    International Nuclear Information System (INIS)

    Ohana, N; Lanti, E; Tran, T M; Brunner, S; Hariri, F; Villard, L; Jocksch, A; Gheller, C

    2016-01-01

    With the aim of enabling state-of-the-art gyrokinetic PIC codes to benefit from the performance of recent multithreaded devices, we developed an application from a platform called the “PIC-engine” [1, 2, 3] embedding simplified basic features of the PIC method. The application solves the gyrokinetic equations in a sheared plasma slab using B-spline finite elements up to fourth order to represent the self-consistent electrostatic field. Preliminary studies of the so-called Particle-In-Fourier (PIF) approach, which uses Fourier modes as basis functions in the periodic dimensions of the system instead of the real-space grid, show that this method can be faster than PIC for simulations with a small number of Fourier modes. Similarly to the PIC-engine, multiple levels of parallelism have been implemented using MPI+OpenMP [2] and MPI+OpenACC [1], the latter exploiting the computational power of GPUs without requiring complete code rewriting. It is shown that sorting particles [3] can lead to performance improvement by increasing data locality and vectorizing grid memory access. Weak scalability tests have been successfully run on the GPU-equipped Cray XC30 Piz Daint (at CSCS) up to 4,096 nodes. The reduced time-to-solution will enable more realistic and thus more computationally intensive simulations of turbulent transport in magnetic fusion devices. (paper)

  13. Proceedings of the 8th topical meeting on nuclear code development

    International Nuclear Information System (INIS)

    1993-03-01

    The 8th Topical Meeting on Nuclear Code Development, organized by Committee on Reactor Physics and Nuclear Codes Committee of Japan Atomic Energy Research Institute (JAERI), was held at Tokai Research Establishment of JAERI, on 11th and 12th of November, 1992. In the meeting, 14 papers were presented on the topics of (1) the next generation nuclear reactor design system and (2) advances of the nuclear fuel reprocessing safety analysis codes. These papers are compiled in this proceedings. (author)

  14. Development of standards, codes of practice and guidelines at the national level

    International Nuclear Information System (INIS)

    Swindon, T.N.

    1989-01-01

    Standards, codes of practice and guidelines are defined and their different roles in radiation protection specified. The work of the major bodies that develop such documents in Australia - the National Health and Medical Research Council and the Standards Association of Australia - is discussed. The codes of practice prepared under the Environment Protection (Nuclear Codes) Act, 1978, an act of the Australian Federal Parliament, are described and the guidelines associated with them outlined. 5 refs

  15. Development of a computer code for Dalat research reactor transient analysis

    International Nuclear Information System (INIS)

    Le Vinh Vinh; Nguyen Thai Sinh; Huynh Ton Nghiem; Luong Ba Vien; Pham Van Lam; Nguyen Kien Cuong

    2003-01-01

    DRSIM (Dalat Reactor SIMulation) computer code has been developed for Dalat reactor transient analysis. It is basically a coupled neutronics-hydrodynamics-heat transfer code employing point kinetics, one dimensional hydrodynamics and one dimensional heat transfer. The work was financed by VAEC and DNRI in the framework of institutional R and D programme. Some transient problems related to reactivity and loss of coolant flow was carried out by DRSIM using temperature and void coefficients calculated by WIMS and HEXNOD2D codes. (author)

  16. Development and application of best-estimate LWR safety analysis codes

    International Nuclear Information System (INIS)

    Reocreux, M.

    1997-01-01

    This paper is a review of the status and the future orientations of the development and application of best estimate LWR safety analysis codes. The present status of these codes exhibits a large success and almost a complete fulfillment of the objectives which were assigned in the 70s. The applications of Best Estimate codes are numerous and cover a large variety of safety questions. However these applications raised a number of problems. The first ones concern the need to have a better control of the quality of the results. This means requirements on code assessment and on uncertainties evaluation. The second ones concern needs for code development and specifically regarding physical models, numerics, coupling with other codes and programming. The analysis of the orientations for code developments and applications in the next years, shows that some developments should be made without delay in order to solve today questions whereas some others are more long term and should be tested for example in some pilot programmes before being eventually applied in main code development. Each of these development programmes are analyzed in the paper by detailing their main content and their possible interest. (author)

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

    International Nuclear Information System (INIS)

    Liu, X.J.; Cheng, X.

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-15

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

  19. Development of hybrid low-pressure MSGC neutron detectors

    International Nuclear Information System (INIS)

    Gebauer, B.; Alimov, S.S.; Klimov, A.Yu.; Levchanovski, F.V.; Litvinenko, E.I.; Nikiforov, A.S.; Prikhodko, V.I.; Richter, G.; Rogov, V.; Schulz, Ch.; Shashkin, V.I.; Wilhelm, M.; Wilpert, Th.

    2004-01-01

    For very high rate and resolution time-resolved experiments at next generation pulsed spallation neutron sources like ESS large-area hybrid low-pressure micro-strip gas chamber detectors are being developed. Due to their thin composite converter foil and exponential gas multiplication commencing at the converter surfaces the detectors are free of parallax, and according to detailed modeling the very high transverse and longitudinal localization accuracies in the conversion and gas multiplication processes allow position and time resolutions of ∼100 μm and 8 cps. This will open up novel applications based on time-of-flight (TOF) and single-event detection with very high dynamic range, replacing integrating CCD and image plate detectors, e.g. in radiography/tomography, TOF Laue diffraction, single crystal diffraction and focusing low-Q SANS. In this conference report new results concerning the technical realization of this detector system are reported in conjunction with a brief summary of the detector principle and with reference to earlier results

  20. Development of commercial hybrid electric vehicle with native key components

    Directory of Open Access Journals (Sweden)

    S. V. Bakhmutov

    2014-01-01

    Full Text Available The perspectives of development of medium weight cargo vehicles with hybrid powertrain including Russian native key components are considered in this article. Series-parallel scheme of HEV is more relevant owing to limitations of series and parallel schemes. An example of this technology is described. This technical solution has good facilities for variation of HEV and AWD type. The authors have patented it. In addition, another main issue is to choose the types of key components with good correlation for parameters of ICE, electric motors, batteries, and inverter. Using mathematical model of the vehicle a selection and correlation of technical characteristics were carried out to meet ecological and economical requirements. After computing calculation two control strategies were accepted. The first strategy contributes to good fuel consumption, while the other one is aimed at ecology. Researchers use test benches to confirm the results of calculation, and this one was built by the authors applying native components. The result of experiment on the test bench is the growth of fuel consumption of the medium weight cargo vehicle by 25% and compliance with ecological class Euro-4.

  1. Recent Developments of JAEA's Monte Carlo Code MVP for Reactor Physics Applications

    Science.gov (United States)

    Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa

    2014-06-01

    This paper describes the recent development status of a Monte Carlo code MVP developed at Japan Atomic Energy Agency. The basic features and capabilities of MVP are overviewed. In addition, new capabilities useful for reactor analysis are also described.

  2. Development of an integrated thermal-hydraulics capability incorporating RELAP5 and PANTHER neutronics code

    Energy Technology Data Exchange (ETDEWEB)

    Page, R.; Jones, J.R.

    1997-07-01

    Ensuring that safety analysis needs are met in the future is likely to lead to the development of new codes and the further development of existing codes. It is therefore advantageous to define standards for data interfaces and to develop software interfacing techniques which can readily accommodate changes when they are made. Defining interface standards is beneficial but is necessarily restricted in application if future requirements are not known in detail. Code interfacing methods are of particular relevance with the move towards automatic grid frequency response operation where the integration of plant dynamic, core follow and fault study calculation tools is considered advantageous. This paper describes the background and features of a new code TALINK (Transient Analysis code LINKage program) used to provide a flexible interface to link the RELAP5 thermal hydraulics code with the PANTHER neutron kinetics and the SIBDYM whole plant dynamic modelling codes used by Nuclear Electric. The complete package enables the codes to be executed in parallel and provides an integrated whole plant thermal-hydraulics and neutron kinetics model. In addition the paper discusses the capabilities and pedigree of the component codes used to form the integrated transient analysis package and the details of the calculation of a postulated Sizewell `B` Loss of offsite power fault transient.

  3. Development of an integrated thermal-hydraulics capability incorporating RELAP5 and PANTHER neutronics code

    International Nuclear Information System (INIS)

    Page, R.; Jones, J.R.

    1997-01-01

    Ensuring that safety analysis needs are met in the future is likely to lead to the development of new codes and the further development of existing codes. It is therefore advantageous to define standards for data interfaces and to develop software interfacing techniques which can readily accommodate changes when they are made. Defining interface standards is beneficial but is necessarily restricted in application if future requirements are not known in detail. Code interfacing methods are of particular relevance with the move towards automatic grid frequency response operation where the integration of plant dynamic, core follow and fault study calculation tools is considered advantageous. This paper describes the background and features of a new code TALINK (Transient Analysis code LINKage program) used to provide a flexible interface to link the RELAP5 thermal hydraulics code with the PANTHER neutron kinetics and the SIBDYM whole plant dynamic modelling codes used by Nuclear Electric. The complete package enables the codes to be executed in parallel and provides an integrated whole plant thermal-hydraulics and neutron kinetics model. In addition the paper discusses the capabilities and pedigree of the component codes used to form the integrated transient analysis package and the details of the calculation of a postulated Sizewell 'B' Loss of offsite power fault transient

  4. Burnup calculation for a tokamak commercial hybrid reactor

    International Nuclear Information System (INIS)

    Feng Kaiming; Xie Zhongyou

    1990-08-01

    A computer code ISOGEN-III and its associated data library BULIB have been developed for fusion-fission hybrid reactor burnup calculations. These are used to calcuate burnup of a tokamak commercial hybrid reactor. The code and library are introduced briefly, and burnup calculation results are given

  5. Development of the PRO-LOCA Probabilistic Fracture Mechanics Code, MERIT Final Report

    International Nuclear Information System (INIS)

    Scott, Paul; Kurth, Robert; Cox, Andrew; Olson, Rick; Rudland, Dave

    2010-12-01

    The MERIT project has been an internationally financed program with the main purpose of developing probabilistic models for piping failure of nuclear components and to include these models in a probabilistic code named PRO-LOCA. The principal objective of the project has been to develop probabilistic models for piping failure of nuclear components and to include these models in a probabilistic code. The MERIT program has produced a code named PRO-LOCA with the following features: - Crack initiation models for fatigue or stress corrosion cracking for previously unflawed material. - Subcritical crack growth models for fatigue and stress corrosion cracking for both initiated and pre-existing circumferential defects. - Models for flaw detection by inspections and leak detection. - Crack stability. The PRO-LOCA code can thus predict the leak or break frequency for the whole sequence of initiation, subcritical crack growth until wall penetration and leakage, instability of the through-wall crack (pipe rupture). The outcome of the PRO-LOCA code are a sequence of failure frequencies which represents the probability of surface crack developing, a through-wall crack developing and six different sizes of crack opening areas corresponding to different leak flow rates or LOCA categories. Note that the level of quality assurance of the PRO-LOCA code is such that the code in its current state of development is considered to be more of a research code than a regulatory tool.

  6. Development of the PRO-LOCA Probabilistic Fracture Mechanics Code, MERIT Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Scott, Paul; Kurth, Robert; Cox, Andrew; Olson, Rick (Battelle Columbus (United States)); Rudland, Dave (Nuclear Regulatory Commission (United States))

    2010-12-15

    The MERIT project has been an internationally financed program with the main purpose of developing probabilistic models for piping failure of nuclear components and to include these models in a probabilistic code named PRO-LOCA. The principal objective of the project has been to develop probabilistic models for piping failure of nuclear components and to include these models in a probabilistic code. The MERIT program has produced a code named PRO-LOCA with the following features: - Crack initiation models for fatigue or stress corrosion cracking for previously unflawed material. - Subcritical crack growth models for fatigue and stress corrosion cracking for both initiated and pre-existing circumferential defects. - Models for flaw detection by inspections and leak detection. - Crack stability. The PRO-LOCA code can thus predict the leak or break frequency for the whole sequence of initiation, subcritical crack growth until wall penetration and leakage, instability of the through-wall crack (pipe rupture). The outcome of the PRO-LOCA code are a sequence of failure frequencies which represents the probability of surface crack developing, a through-wall crack developing and six different sizes of crack opening areas corresponding to different leak flow rates or LOCA categories. Note that the level of quality assurance of the PRO-LOCA code is such that the code in its current state of development is considered to be more of a research code than a regulatory tool.

  7. The development of depletion program coupled with Monte Carlo computer code

    International Nuclear Information System (INIS)

    Nguyen Kien Cuong; Huynh Ton Nghiem; Vuong Huu Tan

    2015-01-01

    The paper presents the development of depletion code for light water reactor coupled with MCNP5 code called the MCDL code (Monte Carlo Depletion for Light Water Reactor). The first order differential depletion system equations of 21 actinide isotopes and 50 fission product isotopes are solved by the Radau IIA Implicit Runge Kutta (IRK) method after receiving neutron flux, reaction rates in one group energy and multiplication factors for fuel pin, fuel assembly or whole reactor core from the calculation results of the MCNP5 code. The calculation for beryllium poisoning and cooling time is also integrated in the code. To verify and validate the MCDL code, high enriched uranium (HEU) and low enriched uranium (LEU) fuel assemblies VVR-M2 types and 89 fresh HEU fuel assemblies, 92 LEU fresh fuel assemblies cores of the Dalat Nuclear Research Reactor (DNRR) have been investigated and compared with the results calculated by the SRAC code and the MCNP R EBUS linkage system code. The results show good agreement between calculated data of the MCDL code and reference codes. (author)

  8. Development of dose assessment code for release of tritium during normal operation of nuclear power plants

    International Nuclear Information System (INIS)

    Duran, J.; Malatova, I.

    2009-01-01

    A computer code PTM H TO has been developed to assess tritium doses to the general public. The code enables to simulate the behavior of tritium in the environment released into the atmosphere under normal operation of nuclear power plants. Code can calculate the doses for the three chemical and physical forms: tritium gas (HT), tritiated water vapor and water drops (HTO). The models in this code consist of the tritium transfer model including oxidation of HT to HTO and reemission of HTO from soil to the atmosphere, and the dose calculation model

  9. 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)

  10. Development of a graphical interface computer code for reactor fuel reloading optimization

    International Nuclear Information System (INIS)

    Do Quang Binh; Nguyen Phuoc Lan; Bui Xuan Huy

    2007-01-01

    This report represents the results of the project performed in 2007. The aim of this project is to develop a graphical interface computer code that allows refueling engineers to design fuel reloading patterns for research reactor using simulated graphical model of reactor core. Besides, this code can perform refueling optimization calculations based on genetic algorithms as well as simulated annealing. The computer code was verified based on a sample problem, which relies on operational and experimental data of Dalat research reactor. This code can play a significant role in in-core fuel management practice at nuclear research reactor centers and in training. (author)

  11. Development of a tracer transport option for the NAPSAC fracture network computer code

    International Nuclear Information System (INIS)

    Herbert, A.W.

    1990-06-01

    The Napsac computer code predicts groundwater flow through fractured rock using a direct fracture network approach. This paper describes the development of a tracer transport algorithm for the NAPSAC code. A very efficient particle-following approach is used enabling tracer transport to be predicted through large fracture networks. The new algorithm is tested against three test examples. These demonstrations confirm the accuracy of the code for simple networks, where there is an analytical solution to the transport problem, and illustrates the use of the computer code on a more realistic problem. (author)

  12. Development and application of sub-channel analysis code based on SCWR core

    International Nuclear Information System (INIS)

    Fu Shengwei; Xu Zhihong; Yang Yanhua

    2011-01-01

    The sub-channel analysis code SABER was developed for thermal-hydraulic analysis of supercritical water-cooled reactor (SCWR) fuel assembly. The extended computational cell structure, a new boundary conditions, 3 dimensional heat conduction model and water properties package were implemented in SABER code, which could be used to simulate the thermal fuel assembly of SCWR. To evaluate the applicability of the code, a steady state calculation of the fuel assembly was performed. The results indicate good applicability of the SABER code to simulate the counter-current flow and the heat exchange between coolant and moderator channels. (authors)

  13. 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)

  14. Development of DUST: A computer code that calculates release rates from a LLW disposal unit

    International Nuclear Information System (INIS)

    Sullivan, T.M.

    1992-01-01

    Performance assessment of a Low-Level Waste (LLW) disposal facility begins with an estimation of the rate at which radionuclides migrate out of the facility (i.e., the disposal unit source term). The major physical processes that influence the source term are water flow, container degradation, waste form leaching, and radionuclide transport. A computer code, DUST (Disposal Unit Source Term) has been developed which incorporates these processes in a unified manner. The DUST code improves upon existing codes as it has the capability to model multiple container failure times, multiple waste form release properties, and radionuclide specific transport properties. Verification studies performed on the code are discussed

  15. Recent X-ray hybrid CMOS detector developments and measurements

    Science.gov (United States)

    Hull, Samuel V.; Falcone, Abraham D.; Burrows, David N.; Wages, Mitchell; Chattopadhyay, Tanmoy; McQuaide, Maria; Bray, Evan; Kern, Matthew

    2017-08-01

    The Penn State X-ray detector lab, in collaboration with Teledyne Imaging Sensors (TIS), have progressed their efforts to improve soft X-ray Hybrid CMOS detector (HCD) technology on multiple fronts. Having newly acquired a Teledyne cryogenic SIDECARTM ASIC for use with HxRG devices, measurements were performed with an H2RG HCD and the cooled SIDECARTM. We report new energy resolution and read noise measurements, which show a significant improvement over room temperature SIDECARTM operation. Further, in order to meet the demands of future high-throughput and high spatial resolution X-ray observatories, detectors with fast readout and small pixel sizes are being developed. We report on characteristics of new X-ray HCDs with 12.5 micron pitch that include in-pixel CDS circuitry and crosstalk-eliminating CTIA amplifiers. In addition, PSU and TIS are developing a new large-scale array Speedster-EXD device. The original 64 × 64 pixel Speedster-EXD prototype used comparators in each pixel to enable event driven readout with order of magnitude higher effective readout rates, which will now be implemented in a 550 × 550 pixel device. Finally, the detector lab is involved in a sounding rocket mission that is slated to fly in 2018 with an off-plane reflection grating array and an H2RG X-ray HCD. We report on the planned detector configuration for this mission, which will increase the NASA technology readiness level of X-ray HCDs to TRL 9.

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

    CERN Document Server

    Taki, M; Kobayashi, H; Yamaguchi, T

    2003-01-01

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

  17. Development and validation of GWHEAD, a three-dimensional groundwater head computer code

    International Nuclear Information System (INIS)

    Beckmeyer, R.R.; Root, R.W.; Routt, K.R.

    1980-03-01

    A computer code has been developed to solve the groundwater flow equation in three dimensions. The code has finite-difference approximations solved by the strongly implicit solution procedure. Input parameters to the code include hydraulic conductivity, specific storage, porosity, accretion (recharge), and initial hydralic head. These parameters may be input as varying spatially. The hydraulic conductivity may be input as isotropic or anisotropic. The boundaries either may permit flow across them or may be impermeable. The code has been used to model leaky confined groundwater conditions and spherical flow to a continuous point sink, both of which have exact analytical solutions. The results generated by the computer code compare well with those of the analytical solutions. The code was designed to be used to model groundwater flow beneath fuel reprocessing and waste storage areas at the Savannah River Plant

  18. Development and Validation of A Nuclear Fuel Cycle Analysis Tool: A FUTURE Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. K.; Ko, W. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Yoon Hee [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2013-10-15

    This paper presents the development and validation methods of the FUTURE (FUel cycle analysis Tool for nUcleaR Energy) code, which was developed for a dynamic material flow evaluation and economic analysis of the nuclear fuel cycle. This code enables an evaluation of a nuclear material flow and its economy for diverse nuclear fuel cycles based on a predictable scenario. The most notable virtue of this FUTURE code, which was developed using C and MICROSOFT SQL DBMS, is that a program user can design a nuclear fuel cycle process easily using a standard process on the canvas screen through a drag-and-drop method. From the user's point of view, this code is very easy to use thanks to its high flexibility. In addition, the new code also enables the maintenance of data integrity by constructing a database environment of the results of the nuclear fuel cycle analyses.

  19. DEVELOPMENT AND VALIDATION OF A NUCLEAR FUEL CYCLE ANALYSIS TOOL: A FUTURE CODE

    Directory of Open Access Journals (Sweden)

    S.K. KIM

    2013-10-01

    Full Text Available This paper presents the development and validation methods of the FUTURE (FUel cycle analysis Tool for nUcleaR Energy code, which was developed for a dynamic material flow evaluation and economic analysis of the nuclear fuel cycle. This code enables an evaluation of a nuclear material flow and its economy for diverse nuclear fuel cycles based on a predictable scenario. The most notable virtue of this FUTURE code, which was developed using C# and MICROSOFT SQL DBMS, is that a program user can design a nuclear fuel cycle process easily using a standard process on the canvas screen through a drag-and-drop method. From the user's point of view, this code is very easy to use thanks to its high flexibility. In addition, the new code also enables the maintenance of data integrity by constructing a database environment of the results of the nuclear fuel cycle analyses.

  20. Development of a relativistic Particle In Cell code PARTDYN for linear accelerator beam transport

    Energy Technology Data Exchange (ETDEWEB)

    Phadte, D., E-mail: deepraj@rrcat.gov.in [LPD, Raja Ramanna Centre for Advanced Technology, Indore 452013 (India); Patidar, C.B.; Pal, M.K. [MAASD, Raja Ramanna Centre for Advanced Technology, Indore (India)

    2017-04-11

    A relativistic Particle In Cell (PIC) code PARTDYN is developed for the beam dynamics simulation of z-continuous and bunched beams. The code is implemented in MATLAB using its MEX functionality which allows both ease of development as well higher performance similar to a compiled language like C. The beam dynamics calculations carried out by the code are compared with analytical results and with other well developed codes like PARMELA and BEAMPATH. The effect of finite number of simulation particles on the emittance growth of intense beams has been studied. Corrections to the RF cavity field expressions were incorporated in the code so that the fields could be calculated correctly. The deviations of the beam dynamics results between PARTDYN and BEAMPATH for a cavity driven in zero-mode have been discussed. The beam dynamics studies of the Low Energy Beam Transport (LEBT) using PARTDYN have been presented.

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

    International Nuclear Information System (INIS)

    Sheron, B.W.; Rosztoczy, Z.R.

    1980-08-01

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

  2. Extension of hybrid micro-depletion model for decay heat calculation in the DYN3D code

    International Nuclear Information System (INIS)

    Bilodid, Yurii; Fridman, Emil; Shwageraus, E.

    2017-01-01

    This work extends the hybrid micro-depletion methodology, recently implemented in DYN3D, to the decay heat calculation by accounting explicitly for the heat contribution from the decay of each nuclide in the fuel.

  3. Extension of hybrid micro-depletion model for decay heat calculation in the DYN3D code

    Energy Technology Data Exchange (ETDEWEB)

    Bilodid, Yurii; Fridman, Emil [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety; Kotlyar, D. [Georgia Institute of Technology, Atlanta, GA (United States); Shwageraus, E. [Cambridge Univ. (United Kingdom)

    2017-06-01

    This work extends the hybrid micro-depletion methodology, recently implemented in DYN3D, to the decay heat calculation by accounting explicitly for the heat contribution from the decay of each nuclide in the fuel.

  4. Gas cloud explosions and their effect on nuclear power plant, basic development of explosion codes

    International Nuclear Information System (INIS)

    Hall, S.F.; Martin, D.; MacKenzie, J.

    1985-01-01

    The study of factors influencing the pressure and velocity fields produced by the burning of flammable substances has been in progress at SRD for some years. This paper describes an extension of these studies by using existing codes for a parametric survey, and modifying codes to produce more realistic representations of explosions and developing a two dimensional combustion code, FLARE. The one dimensional combustion code, GASEX1, has been used to determine the pressure from a burning gas cloud for a number of different fuels, concentrations and burning velocities. The code was modified so that gas concentrations could be modelled. Results for concentration gradients showed the pressure depended on local conditions and the burning velocity. The two dimensional code, GASEX2, was modified to model the interaction of pressure waves with structures. It was used, with results from GASEX1, to model the interaction of a pressure wave from the combustion of a gas cloud with a rigid structure representing a nuclear power plant. The two dimensional code FLARE has been developed to model the interaction of flames and pressure waves with structures. The code incorporates a simple turbulence model with a turbulence dependent reaction rate. Validation calculations have been carried out for the code. (author)

  5. Safety analysis code SCTRAN development for SCWR and its application to CGNPC SCWR

    International Nuclear Information System (INIS)

    Wu, Pan; Gou, Junli; Shan, Jianqiang; Jiang, Yang; Yang, Jue; Zhang, Bo

    2013-01-01

    Highlights: ► A new safety analysis code named SCTRAN is developed for SCWRs. ► Capability of SCTRAN is verified by comparing with code APROS and RELAP5-3D. ► A new passive safety system is proposed for CGNPC SCWR and analyzed with SCTRAN. ► CGNPC SCWR is able to cope with two critical accidents for SCWRs, LOFA and LOCA. - Abstract: Design analysis is one of the main difficulties during the research and design of SCWRs. Currently, the development of safety analysis code for SCWR is still in its infancy all around the world, and very few computer codes could carry out the trans-critical calculations where significant changes in water properties would take place. In this paper, a safety analysis code SCTRAN for SCWRs has been developed based on code RETRAN-02, the best estimate code used for safety analysis of light water reactors. The ability of SCTRAN code to simulate transients where both supercritical and subcritical regimes are encountered has been verified by comparing with APROS and RELAP5-3D codes. Furthermore, the LOFA and LOCA transients for the CGNPC SCWR design were analyzed with SCTRAN code. The characteristics and performance of the passive safety systems applied to CGNPC SCWR were evaluated. The results show that: (1) The SCTRAN computer code developed in this study is capable to perform design analysis for SCWRs; (2) During LOFA and LOCA accidents in a CGNPC SCWR, the passive safety systems would significantly mitigate the consequences of these transients and enhance the inherent safety

  6. Development of 2D particle-in-cell code to simulate high current, low ...

    Indian Academy of Sciences (India)

    Abstract. A code for 2D space-charge dominated beam dynamics study in beam trans- port lines is developed. The code is used for particle-in-cell (PIC) simulation of z-uniform beam in a channel containing solenoids and drift space. It can also simulate a transport line where quadrupoles are used for focusing the beam.

  7. The development of fluid codes for the laser compression of plasma

    International Nuclear Information System (INIS)

    Nicholas, D.J.

    1982-08-01

    Notes are given on the construction and use of simulation codes in plasma physics requiring only a limited background knowledge in numerical analysis and finite-difference techniques. The development of a 1-D Eulerian codes to source form is followed as an example. (U.K.)

  8. Developing a coding scheme for detecting usability and fun problems in computer games for young children

    NARCIS (Netherlands)

    Barendregt, W.; Bekker, M.M.

    2006-01-01

    This article describes the development and assessment of a coding scheme for finding both usability and fun problems through observations of young children playing computer games during user tests. The proposed coding scheme is based on an existing list of breakdown indication types of the detailed

  9. The gradual development steps of the external coupled RELAP5 - DYN3D code

    International Nuclear Information System (INIS)

    Strmensky, C.

    2001-01-01

    This paper describes the on-going and finished parts of project: 'The external coupled RELAP5-DYN3D code'. The development progress was divided into four steps. In present time, second and third steps are performed and four step is started. The two parameters of coolant was selected and are exchanged between codes RELAP5 and DYN3D. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    We are estimating that the debris containing fuel are piled in the containment and the pressure vessel bottoms of Fukushima-Daiichi NPPs. A radioactive Xe concentration discharged in recriticality is being monitored by utilizing the gas management system set up in NPPs unit 1-3. For this reason, we can confirm the recriticality might not be broken out. However, the debris conditions distributed in the containment vessel and the pressure vessel bottoms are not clear. The internal and external surrounding changes will make recriticality become possible. According to TEPCO's roadmap, TEPCO will launch extracting task within 10 years. Even in the case that the fuel condition changes due to debris relocation and mixture, subcriticality must be secured. Criticality safety analysis with non-uniform effect should therefore be essential for the molten debris. For above reasons, we studies the optimum distributions with some parameters that have a large reactivity change were assessed with OPT-DANT code. Finally, the boron concentration was estimated in order to keep subcriticality. (author)

  11. A study on the development of photoelastic experimental hybrid method for colour isochromatics (I)

    International Nuclear Information System (INIS)

    Hawong, Jai Sug; Nam, Jeong Hwan; Kim, Kyo Hyoung; Kwon, O Sung; Kwon, Gun; Park, Sung Han

    2010-01-01

    Isochromatics obtained from photoelastic experiment shows the stress distributions of the full field of a structure under load. Therefore, stress distributions of the structure can be read at a glance through isochromatics. Many experimental data can be obtained from isochromatics which are then used in various photoelastic experimental hybrid methods for stress analysis. Monochromatic light has however, until now been used in the photoelastic experimental hybrid method to produce black and white isochromatics. The use of black and white isochromatics in photoelastic experimental hybrid method for black and white isochromatics requires high fringe orders in order to obtain sufficient experimental data for photoelastic hybrid techniques. Accordingly, this paper develops the photoelastic experimental hybrid method for color isochromatics in which a fringe order of 1 is enough to gather the experimental data of the photoelastic experimental hybrid method. The method was applied to validate stress concentration problems. Experimental results from this study indicated that the photoelastic experimental hybrid method for color isochromatics is more precise than the photoelastic experimental hybrid method for black and white isochromatics. The use of few fringe orders in photoelastic experimental hybrid method for color isochromatics can offer significant advantages in stress analysis of real components using reflective-type photoelastic experimental method

  12. Hybrid2 - The hybrid power system simulation model

    Energy Technology Data Exchange (ETDEWEB)

    Baring-Gould, E.I.; Green, H.J.; Dijk, V.A.P. van [National Renewable Energy Lab., Golden, CO (United States); Manwell, J.F. [Univ. of Massachusetts, Amherst, MA (United States)

    1996-12-31

    There is a large-scale need and desire for energy in remote communities, especially in the developing world; however the lack of a user friendly, flexible performance prediction model for hybrid power systems incorporating renewables hindered the analysis of hybrids as options to conventional solutions. A user friendly model was needed with the versatility to simulate the many system locations, widely varying hardware configurations, and differing control options for potential hybrid power systems. To meet these ends, researchers from the National Renewable Energy Laboratory (NREL) and the University of Massachusetts (UMass) developed the Hybrid2 software. This paper provides an overview of the capabilities, features, and functionality of the Hybrid2 code, discusses its validation and future plans. Model availability and technical support provided to Hybrid2 users are also discussed. 12 refs., 3 figs., 4 tabs.

  13. Recent developments of JAEA’s Monte Carlo code MVP for reactor physics applications

    International Nuclear Information System (INIS)

    Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa

    2015-01-01

    Highlights: • This paper describes the recent development status of the Monte Carlo code MVP. • The basic features and capabilities of MVP are briefly described. • New capabilities useful for reactor analysis are also described. - Abstract: This paper describes the recent development status of a Monte Carlo code MVP developed at Japan Atomic Energy Agency. The basic features and capabilities of MVP are overviewed. In addition, new capabilities useful for reactor analysis are also described

  14. Development of a system of computer codes for severe accident analyses and its applications

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Soon Hong; Cheon, Moon Heon; Cho, Nam jin; No, Hui Cheon; Chang, Hyeon Seop; Moon, Sang Kee; Park, Seok Jeong; Chung, Jee Hwan [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1991-12-15

    The objectives of this study is to develop a system of computer codes for postulated severe accident analyses in Nuclear Power Plants. This system of codes is necessary to conduct individual plant examination for domestic nuclear power plants. As a result of this study, one can conduct severe accident assessments more easily, and can extract the plant-specific vulnerabilities for severe accidents and at the same time the ideas for enhancing overall accident resistance. The scope and contents of this study are as follows : development of a system of computer codes for severe accident analyses, development of severe accident management strategy.

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

    International Nuclear Information System (INIS)

    Chang, Soon Hong; Cheon, Moon Heon; Cho, Nam jin; No, Hui Cheon; Chang, Hyeon Seop; Moon, Sang Kee; Park, Seok Jeong; Chung, Jee Hwan

    1991-12-01

    The objectives of this study is to develop a system of computer codes for postulated severe accident analyses in Nuclear Power Plants. This system of codes is necessary to conduct individual plant examination for domestic nuclear power plants. As a result of this study, one can conduct severe accident assessments more easily, and can extract the plant-specific vulnerabilities for severe accidents and at the same time the ideas for enhancing overall accident resistance. The scope and contents of this study are as follows : development of a system of computer codes for severe accident analyses, development of severe accident management strategy

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

    International Nuclear Information System (INIS)

    Zhang, H.

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hyun; Seo, K. W

    2006-01-15

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

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

    International Nuclear Information System (INIS)

    Hwang, Dae Hyun; Seo, K. W.

    2006-01-01

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

  20. Development of hybrid fluid jet/float polishing process

    Science.gov (United States)

    Beaucamp, Anthony T. H.; Namba, Yoshiharu; Freeman, Richard R.

    2013-09-01

    On one hand, the "float polishing" process consists of a tin lap having many concentric grooves, cut from a flat by single point diamond turning. This lap is rotated above a hydrostatic bearing spindle of high rigidity, damping and rotational accuracy. The optical surface thus floats above a thin layer of abrasive particles. But whilst surface texture can be smoothed to ~0.1nm rms (as measured by atomic force microscopy), this process can only be used on flat surfaces. On the other hand, the CNC "fluid jet polishing" process consists of pumping a mixture of water and abrasive particles to a converging nozzle, thus generating a polishing spot that can be moved along a tool path with tight track spacing. But whilst tool path feed can be moderated to ultra-precisely correct form error on freeform optical surfaces, surface finish improvement is generally limited to ~1.5nm rms (with fine abrasives). This paper reports on the development of a novel finishing method, that combines the advantages of "fluid jet polishing" (i.e. freeform corrective capability) with "float polishing" (i.e. super-smooth surface finish of 0.1nm rms or less). To come up with this new "hybrid" method, computational fluid dynamic modeling of both processes in COMSOL is being used to characterize abrasion conditions and adapt the process parameters of experimental fluid jet polishing equipment, including: (1) geometrical shape of nozzle, (2) position relative to the surface, (3) control of inlet pressure. This new process is aimed at finishing of next generation X-Ray / Gamma Ray focusing optics.

  1. Modified hybrid subcarrier/amplitude/ phase/polarization LDPC-coded modulation for 400 Gb/s optical transmission and beyond.

    Science.gov (United States)

    Batshon, Hussam G; Djordjevic, Ivan; Xu, Lei; Wang, Ting

    2010-06-21

    In this paper, we present a modified coded hybrid subcarrier/ amplitude/phase/polarization (H-SAPP) modulation scheme as a technique capable of achieving beyond 400 Gb/s single-channel transmission over optical channels. The modified H-SAPP scheme profits from the available resources in addition to geometry to increase the bandwidth efficiency of the transmission system, and so increases the aggregate rate of the system. In this report we present the modified H-SAPP scheme and focus on an example that allows 11 bits/Symbol that can achieve 440 Gb/s transmission using components of 50 Giga Symbol/s (GS/s).

  2. Krylov solvers preconditioned with the low-order red-black algorithm for the PN hybrid FEM for the instant code

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Yaqi; Rabiti, Cristian; Palmiotti, Giuseppe, E-mail: yaqi.wang@inl.gov, E-mail: cristian.rabiti@inl.gov, E-mail: giuseppe.palmiotti@inl.gov [Idaho National Laboratory, Idaho Falls, ID (United States)

    2011-07-01

    This paper proposes a new set of Krylov solvers, CG and GMRes, as an alternative of the Red-Black (RB) algorithm on on solving the steady-state one-speed neutron transport equation discretized with PN in angle and hybrid FEM (Finite Element Method) in space. A pre conditioner with the low-order RB iteration is designed to improve their convergence. These Krylov solvers can reduce the cost of pre-assembling the response matrices greatly. Numerical results with the INSTANT code are presented in order to show that they can be a good supplement on solving the PN-HFEM system. (author)

  3. Krylov solvers preconditioned with the low-order red-black algorithm for the PN hybrid FEM for the instant code

    International Nuclear Information System (INIS)

    Wang, Yaqi; Rabiti, Cristian; Palmiotti, Giuseppe

    2011-01-01

    This paper proposes a new set of Krylov solvers, CG and GMRes, as an alternative of the Red-Black (RB) algorithm on on solving the steady-state one-speed neutron transport equation discretized with PN in angle and hybrid FEM (Finite Element Method) in space. A pre conditioner with the low-order RB iteration is designed to improve their convergence. These Krylov solvers can reduce the cost of pre-assembling the response matrices greatly. Numerical results with the INSTANT code are presented in order to show that they can be a good supplement on solving the PN-HFEM system. (author)

  4. A code of mines for the sustainable development

    International Nuclear Information System (INIS)

    Gonzalez Serna, Carmen Lucia

    2000-01-01

    In accordance with the article 80 of the political constitution of Colombia, the state will plan the handling and use of the natural resources, to guarantee its sustainable development, its conservation, restoration or substitution. When developing this constitutional norm for the renewable natural resources, the law 99 of 1993 in their first article determined that the process of economic and social development of the country will be guided according to the universal principles and of the sustainable development, contents in the declaration of Rio de Janeiro of June of 1992 on environment and development and in the third article defined the concept of sustainable development as that that drives to the economic growth, to the elevation of the quality of life and the social well-being, without draining the base of renewable natural resources in that it is sustained. The author includes antecedents, modification to the mining and law legislation among others

  5. The development and application of a sub-channel code in ocean environment

    International Nuclear Information System (INIS)

    Wu, Pan; Shan, Jianqiang; Xiang, Xiong; Zhang, Bo; Gou, Junli; Zhang, Bin

    2016-01-01

    Highlights: • A sub-channel code named ATHAS/OE is developed for nuclear reactors in ocean environment. • ATHAS/OE is verified by another modified sub-channel code based on COBRA-IV. • ATHAS/OE is used to analyze thermal hydraulic of a typical SMR in heaving and rolling motion. • Calculation results show that ocean condition affect the thermal hydraulic of a reactor significantly. - Abstract: An upgraded version of ATHAS sub-channel code ATHAS/OE is developed for the investigation of the thermal hydraulic behavior of nuclear reactor core in ocean environment with consideration of heaving and rolling motion effect. The code is verified by another modified sub-channel code based on COBRA-IV and used to analyze the thermal hydraulic characteristics of a typical SMR under heaving and rolling motion condition. The calculation results show that the heaving and rolling motion affect the thermal hydraulic behavior of a reactor significantly.

  6. 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)

  7. Thermal-hydraulic analysis code development and application to passive safety reactor at JAERI

    International Nuclear Information System (INIS)

    Araya, F.

    1995-01-01

    After a brief overview of safety assessment process, the author describes the LOCA analysis code system developed in JAERI. It comprises audit calculation code (WREM, WREM-J2, Japanese own code and BE codes (2D/3D, ICAP, ROSA). The codes are applied to development of Japanese passive safety reactor concept JPSR. Special attention is paid to the passive heat removal system and phenomena considered to occur under loss of heat sink event. Examples of LOCA analysis based on operation of JPSR for the cases of heat removal by upper RHR and heat removal from core to atmosphere are given. Experiments for multi-dimensional flow field in RPV and steam condensation in water pool are used for understanding the phenomena in passive safety reactors. The report is in a poster form only. 1 tab., 13 figs

  8. Safety, codes and standards for hydrogen installations. Metrics development and benchmarking

    Energy Technology Data Exchange (ETDEWEB)

    Harris, Aaron P. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Dedrick, Daniel E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); LaFleur, Angela Christine [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); San Marchi, Christopher W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-04-01

    Automakers and fuel providers have made public commitments to commercialize light duty fuel cell electric vehicles and fueling infrastructure in select US regions beginning in 2014. The development, implementation, and advancement of meaningful codes and standards is critical to enable the effective deployment of clean and efficient fuel cell and hydrogen solutions in the energy technology marketplace. Metrics pertaining to the development and implementation of safety knowledge, codes, and standards are important to communicate progress and inform future R&D investments. This document describes the development and benchmarking of metrics specific to the development of hydrogen specific codes relevant for hydrogen refueling stations. These metrics will be most useful as the hydrogen fuel market transitions from pre-commercial to early-commercial phases. The target regions in California will serve as benchmarking case studies to quantify the success of past investments in research and development supporting safety codes and standards R&D.

  9. Code development of the national hemovigilance system and expansion strategies for hospital blood banks

    Directory of Open Access Journals (Sweden)

    Kim Jeongeun

    2012-01-01

    Full Text Available Objectives : The aims of this study were to develop reportable event codes that are applicable to the national hemovigilance systems for hospital blood banks, and to present expansion strategies for the blood banks. Materials and Methods : The data were obtained from a literature review and expert consultation, followed by adding to and revising the established hemovigilance code system and guidelines to develop reportable event codes for hospital blood banks. The Medical Error Reporting System-Transfusion Medicine developed in the US and other codes of reportable events were added to the Korean version of the Biologic Products Deviation Report (BPDR developed by the Korean Red Cross Blood Safety Administration, then using these codes, mapping work was conducted. We deduced outcomes suitable for practice, referred to the results of the advisory councils, and conducted a survey with experts and blood banks practitioners. Results : We developed reportable event codes that were applicable to hospital blood banks and could cover blood safety - from blood product safety to blood transfusion safety - and also presented expansion strategies for hospital blood banks. Conclusion : It was necessary to add 10 major categories to the blood transfusion safety stage and 97 reportable event codes to the blood safety stage. Contextualized solutions were presented on 9 categories of expansion strategies of hemovigilance system for the hospital blood banks.

  10. Developing a Code of Practice for Learning Analytics

    Science.gov (United States)

    Sclater, Niall

    2016-01-01

    Ethical and legal objections to learning analytics are barriers to development of the field, thus potentially denying students the benefits of predictive analytics and adaptive learning. Jisc, a charitable organization that champions the use of digital technologies in UK education and research, has attempted to address this with the development of…

  11. Open-Source Development of the Petascale Reactive Flow and Transport Code PFLOTRAN

    Science.gov (United States)

    Hammond, G. E.; Andre, B.; Bisht, G.; Johnson, T.; Karra, S.; Lichtner, P. C.; Mills, R. T.

    2013-12-01

    Open-source software development has become increasingly popular in recent years. Open-source encourages collaborative and transparent software development and promotes unlimited free redistribution of source code to the public. Open-source development is good for science as it reveals implementation details that are critical to scientific reproducibility, but generally excluded from journal publications. In addition, research funds that would have been spent on licensing fees can be redirected to code development that benefits more scientists. In 2006, the developers of PFLOTRAN open-sourced their code under the U.S. Department of Energy SciDAC-II program. Since that time, the code has gained popularity among code developers and users from around the world seeking to employ PFLOTRAN to simulate thermal, hydraulic, mechanical and biogeochemical processes in the Earth's surface/subsurface environment. PFLOTRAN is a massively-parallel subsurface reactive multiphase flow and transport simulator designed from the ground up to run efficiently on computing platforms ranging from the laptop to leadership-class supercomputers, all from a single code base. The code employs domain decomposition for parallelism and is founded upon the well-established and open-source parallel PETSc and HDF5 frameworks. PFLOTRAN leverages modern Fortran (i.e. Fortran 2003-2008) in its extensible object-oriented design. The use of this progressive, yet domain-friendly programming language has greatly facilitated collaboration in the code's software development. Over the past year, PFLOTRAN's top-level data structures were refactored as Fortran classes (i.e. extendible derived types) to improve the flexibility of the code, ease the addition of new process models, and enable coupling to external simulators. For instance, PFLOTRAN has been coupled to the parallel electrical resistivity tomography code E4D to enable hydrogeophysical inversion while the same code base can be used as a third

  12. Development and application of a hybrid transport methodology for active interrogation systems

    Energy Technology Data Exchange (ETDEWEB)

    Royston, K.; Walters, W.; Haghighat, A. [Nuclear Engineering Program, Department of Mechanical Engineering, Virginia Tech., 900 N Glebe Rd., Arlington, VA 22203 (United States); Yi, C.; Sjoden, G. [Nuclear and Radiological Engineering, Georgia Tech, 801 Ferst Drive, Atlanta, GA 30332 (United States)

    2013-07-01

    A hybrid Monte Carlo and deterministic methodology has been developed for application to active interrogation systems. The methodology consists of four steps: i) neutron flux distribution due to neutron source transport and subcritical multiplication; ii) generation of gamma source distribution from (n, 7) interactions; iii) determination of gamma current at a detector window; iv) detection of gammas by the detector. This paper discusses the theory and results of the first three steps for the case of a cargo container with a sphere of HEU in third-density water cargo. To complete the first step, a response-function formulation has been developed to calculate the subcritical multiplication and neutron flux distribution. Response coefficients are pre-calculated using the MCNP5 Monte Carlo code. The second step uses the calculated neutron flux distribution and Bugle-96 (n, 7) cross sections to find the resulting gamma source distribution. In the third step the gamma source distribution is coupled with a pre-calculated adjoint function to determine the gamma current at a detector window. The AIMS (Active Interrogation for Monitoring Special-Nuclear-Materials) software has been written to output the gamma current for a source-detector assembly scanning across a cargo container using the pre-calculated values and taking significantly less time than a reference MCNP5 calculation. (authors)

  13. What Are They Talking About? Analyzing Code Reviews in Pull-Based Development Model

    Institute of Scientific and Technical Information of China (English)

    Zhi-Xing Li; Yue Yu; Gang Yin; Tao Wang; Huai-Min Wang

    2017-01-01

    Code reviews in pull-based model are open to community users on GitHub. Various participants are taking part in the review discussions and the review topics are not only about the improvement of code contributions but also about project evolution and social interaction. A comprehensive understanding of the review topics in pull-based model would be useful to better organize the code review process and optimize review tasks such as reviewer recommendation and pull-request prioritization. In this paper, we first conduct a qualitative study on three popular open-source software projects hosted on GitHub and construct a fine-grained two-level taxonomy covering four level-1 categories (code correctness, pull-request decision-making, project management, and social interaction) and 11 level-2 subcategories (e.g., defect detecting, reviewer assigning, contribution encouraging). Second, we conduct preliminary quantitative analysis on a large set of review comments that were labeled by TSHC (a two-stage hybrid classification algorithm), which is able to automatically classify review comments by combining rule-based and machine-learning techniques. Through the quantitative study, we explore the typical review patterns. We find that the three projects present similar comments distribution on each subcategory. Pull-requests submitted by inexperienced contributors tend to contain potential issues even though they have passed the tests. Furthermore, external contributors are more likely to break project conventions in their early contributions.

  14. Development of a 1 D hybrid HTC model using CFD simulations for the analysis of direct contact condensation as the driving force for water hammers

    Energy Technology Data Exchange (ETDEWEB)

    Ceuca, Christian Sabin; Macian-Juan, Rafael [Technische Univ. Muenchen (Germany). Lehrstuhl fuer Nukleartechnik

    2013-03-15

    A Hybrid Heat Transfer Coefficient module has been developed based on two Surface Renewal Theory models using CFD simulations. The validation of the model has been done on a meso-scale computational grid for CFD simulations and on a macro-scale computational grid for System Code analysis. The CFD simulation was performed for a stratified co-current two phase flow between saturated steam and sub-cooled water while the System Code analysis was performed for a Condensation Induced Water Hammer experiment. (orig.)

  15. Development of tools for automatic generation of PLC code

    OpenAIRE

    Koutli, Maria; Chasapis, Georgios; Rochez, Jacques

    2014-01-01

    This Master thesis was performed at CERN and more specifically in the EN-ICE-PLC section. The Thesis describes the integration of two PLC platforms, that are based on CODESYS development tool, to the CERN defined industrial framework, UNICOS. CODESYS is a development tool for PLC programming, based on IEC 61131-3 standard, and is adopted by many PLC manufacturers. The two PLC development environments are, the SoMachine from Schneider and the TwinCAT from Beckhoff. The two CODESYS compatible P...

  16. The development, qualification and availability of AECL analytical, scientific and design codes

    International Nuclear Information System (INIS)

    Kupferschmidt, W.C.H.; Fehrenbach, P.J.; Wolgemuth, G.A.; McDonald, B.H.; Snell, V.G.

    2001-01-01

    Over the past several years, AECL has embarked on a comprehensive program to develop, qualify and support its key safety and licensing codes, and to make executable versions of these codes available to the international nuclear community. To this end, we have instituted a company-wide Software Quality Assurance (SQA) Program for Analytical, Scientific and Design Computer Programs to ensure that the design, development, maintenance, modification, procurement and use of computer codes within AECL is consistent with today's quality assurance standards. In addition, we have established a comprehensive Code Validation Project (CVP) with the goal of qualifying AECL's 'front-line' safety and licensing codes by 2001 December. The outcome of this initiative will be qualified codes, which are properly verified and validated for the expected range of applications, with associated statements of accuracy and uncertainty for each application. The code qualification program, based on the CSA N286.7 standard, is intended to ensure (1) that errors are not introduced into safety analyses because of deficiencies in the software, (2) that an auditable documentation base is assembled that demonstrates to the regulator that the codes are of acceptable quality, and (3) that these codes are formally qualified for their intended applications. Because AECL and the Canadian nuclear utilities (i.e., Ontario Power Generation, Bruce Power, Hydro Quebec and New Brunswick Power) generally use the same safety and licensing codes, the nuclear industry in Canada has agreed to work cooperatively together towards the development, qualification and maintenance of a common set of analysis tools, referred to as the Industry Standard Toolset (IST). This paper provides an overview of the AECL Software Quality Assurance Program and the Code Validation Project, and their associated linkages to the Canadian nuclear community's Industry Standard Toolset initiative to cooperatively qualify and support commonly

  17. 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.

  18. 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

  19. Development of a coupling code for PWR reactor cavity radiation streaming calculation

    International Nuclear Information System (INIS)

    Zheng, Z.; Wu, H.; Cao, L.; Zheng, Y.; Zhang, H.; Wang, M.

    2012-01-01

    PWR reactor cavity radiation streaming is important for the safe of the personnel and equipment, thus calculation has to be performed to evaluate the neutron flux distribution around the reactor. For this calculation, the deterministic codes have difficulties in fine geometrical modeling and need huge computer resource; and the Monte Carlo codes require very long sampling time to obtain results with acceptable precision. Therefore, a coupling method has been developed to eliminate the two problems mentioned above in each code. In this study, we develop a coupling code named DORT2MCNP to link the Sn code DORT and Monte Carlo code MCNP. DORT2MCNP is used to produce a combined surface source containing top, bottom and side surface simultaneously. Because SDEF card is unsuitable for the combined surface source, we modify the SOURCE subroutine of MCNP and compile MCNP for this application. Numerical results demonstrate the correctness of the coupling code DORT2MCNP and show reasonable agreement between the coupling method and the other two codes (DORT and MCNP). (authors)

  20. Development and characterization of hybrids from native wine yeasts

    Directory of Open Access Journals (Sweden)

    Verónica García

    2012-06-01

    Full Text Available For commercial purposes, the winemaking industry is constantly searching for new yeast strains. Historically, this has been achieved by collecting wild strains and selecting the best for industrial use through an enological evaluation. Furthermore, the increasing consumer demands have forced the industry to incorporate new strategies such as genetic engineering to obtain improved strains. In response to the lack of public acceptance of this methodology, alternative strategies based on breeding have gained acceptance in recent years. Through the use of conjugation of individual spores without the support of genetic engineering methods we generated intraspecific hybrids from wild strains with outstanding enological characteristics and interdelta fingerprinting was used to confirm the hybrid condition. A detailed enological characterization of the hybrids in synthetic and natural must indicates that physiological parameters such as sporulation, residual sugar, ethanol yield and total nitrogen uptake are within the levels determined for the parental strains, however, other parameters such as growth rate, lag phase and ethanol production show statistical differences with some parental or commercial strains. These findings allow us to propose these hybrids as new wine-making strains.

  1. Development of probabilistic fracture mechanics code PASCAL and user's manual

    Energy Technology Data Exchange (ETDEWEB)

    Shibata, Katsuyuki; Onizawa, Kunio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Li, Yinsheng; Kato, Daisuke [Fuji Research Institute Corporation, Tokyo (Japan)

    2001-03-01

    As a part of the aging and structural integrity research for LWR components, a new PFM (Probabilistic Fracture Mechanics) code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed since FY1996. This code evaluates the failure probability of an aged reactor pressure vessel subjected to transient loading such as PTS (Pressurized Thermal Shock). The development of the code has been aimed to improve the accuracy and reliability of analysis by introducing new analysis methodologies and algorithms considering the recent development in the fracture mechanics methodologies and computer performance. The code has some new functions in optimized sampling and cell dividing procedure in stratified Monte Carlo simulation, elastic-plastic fracture criterion of R6 method, extension analysis models in semi-elliptical crack, evaluation of effect of thermal annealing and etc. In addition, an input data generator of temperature and stress distribution time histories was also prepared in the code. Functions and performance of the code have been confirmed based on the verification analyses and some case studies on the influence parameters. The present phase of the development will be completed in FY2000. Thus this report provides the user's manual and theoretical background of the code. (author)

  2. Development of thermal hydraulic models for the reliable regulatory auditing code

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Song, C. H.; Lee, Y. J.; Kwon, T. S.; Lee, S. W. [Korea Automic Energy Research Institute, Taejon (Korea, Republic of)

    2004-02-15

    The objective of this project is to develop thermal hydraulic models for use in improving the reliability of the regulatory auditing codes. The current year fall under the second step of the 3 year project, and the main researches were focused on the development of downcorner boiling model. During the current year, the bubble stream model of downcorner has been developed and installed in he auditing code. The model sensitivity analysis has been performed for APR1400 LBLOCA scenario using the modified code. The preliminary calculation has been performed for the experimental test facility using FLUENT and MARS code. The facility for air bubble experiment has been installed. The thermal hydraulic phenomena for VHTR and super critical reactor have been identified for the future application and model development.

  3. Development of Learning Management in Moral Ethics and Code of Ethics of the Teaching Profession Course

    Science.gov (United States)

    Boonsong, S.; Siharak, S.; Srikanok, V.

    2018-02-01

    The purposes of this research were to develop the learning management, which was prepared for the enhancement of students’ Moral Ethics and Code of Ethics in Rajamangala University of Technology Thanyaburi (RMUTT). The contextual study and the ideas for learning management development was conducted by the document study, focus group method and content analysis from the document about moral ethics and code of ethics of the teaching profession concerning Graduate Diploma for Teaching Profession Program. The main tools of this research were the summarize papers and analyse papers. The results of development showed the learning management for the development of moral ethics and code of ethics of the teaching profession for Graduate Diploma for Teaching Profession students could promote desired moral ethics and code of ethics of the teaching profession character by the integrated learning techniques which consisted of Service Learning, Contract System, Value Clarification, Role Playing, and Concept Mapping. The learning management was presented in 3 steps.

  4. Development Perspective of Regulatory Audit Code System for SFR Nuclear Safety Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Moo Hoon; Lee, Gil Soo; Shin, An Dong; Suh, Nam Duk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2012-05-15

    A sodium-cooled fast reactor (SFR) in Korea is based on the KALIMER-600 concept developed by KAERI. Based on 'Long-term R and D Plan for Future Reactor Systems' which was approved by the Korea Atomic Energy Commission in 2008, the KAERI designer is scheduled to apply the design certification of the prototype SFR in 2017. In order to establish regulatory infrastructure for the licensing of a prototype SFR, KINS has develop the regulatory requirements for the demonstration SFR since 2010, and are scheduled to develop the regulatory audit code systems in regard to core, fuel, and system, etc. since 2012. In this study, the domestic code systems used for core design and safety evaluation of PWRs and the nuclear physics and code system for SFRs were briefly reviewed, and the development perspective of regulatory audit code system for SFR nuclear safety evaluation were derived

  5. Development of 3D CFD code based on structured non-orthogonal grids

    International Nuclear Information System (INIS)

    Vaidya, Abhijeet Mohan; Maheshwari, Naresh Kumar; Rama Rao, A.

    2016-01-01

    Most of the nuclear industry problems involve complex geometries. Solution of flow and heat transfer over complex geometries is a very important requirement for designing new reactor systems. Hence development of a general purpose three dimensional (3D) CFD code is undertaken. For handling complex shape of computational domain, implementation on structured non-orthogonal coordinates is being done. The code is validated by comparing its results for 3D inclined lid driven cavity at different inclination angles and Reynolds numbers with OpenFOAM results. This paper contains formulation and validation of the new code developed. (author)

  6. A framework for developing finite element codes for multi-disciplinary applications.

    OpenAIRE

    Dadvand, Pooyan

    2007-01-01

    The world of computing simulation has experienced great progresses in recent years and requires more exigent multidisciplinary challenges to satisfy the new upcoming demands. Increasing the importance of solving multi-disciplinary problems makes developers put more attention to these problems and deal with difficulties involved in developing software in this area. Conventional finite element codes have several difficulties in dealing with multi-disciplinary problems. Many of these codes are d...

  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. Recent developments in ASSERT-PV code for subchannel thermalhydraulics

    International Nuclear Information System (INIS)

    Rao, Y.F.; Hammouda, N.

    2003-01-01

    This paper summarises recent development of ASSERT-PV, and provides examples of applications to CANDU fuel bundles in predicting flow, heat transfer and sheath temperatures. The development work is intended to improve computational and phenomenological modelling capabilities of ASSERT-PV in simulating various flow scenarios in CANDU fuel bundles. The latest version of ASSERT-PV can be used for simulations of steady state or transient, subchannel thermalhydraulics in CANDU bundles under conditions up to and including post-dryout heat transfer. (author)

  9. Development of the multistep compound process calculation code

    Energy Technology Data Exchange (ETDEWEB)

    Kawano, Toshihiko [Kyushu Univ., Fukuoka (Japan)

    1998-03-01

    A program `cmc` has been developed to calculate the multistep compound (MSC) process by Feshback-Kerman-Koonin. A radial overlap integral in the transition matrix element is calculated microscopically, and comparisons are made for neutron induced {sup 93}Nb reactions. Strengths of the two-body interaction V{sub 0} are estimated from the total MSC cross sections. (author)

  10. Development and Application of a Plant Code to the Analysis of Transients in Integrated Reactors

    International Nuclear Information System (INIS)

    Rabiti, A.; Gimenez, M.; Delmastro, D.; Zanocco, P.

    2003-01-01

    In this work, a secondary system model for a CAREM-25 type nuclear power plant was developed.A two-phase flow homogenous model was used and found adequate for the scope of the present work.A finite difference scheme was used for the numerical implementation of the model.This model was coupled to the HUARPE code, a primary circuit code, in order to obtain a plant code.This plant code was used to analyze the inherent response of the system, without control feedback loops, for a transient of steam generator feed-water mass flow reduction.The results obtained are satisfactory, but a validation against other plant codes is still necessary

  11. Development of statistical analysis code for meteorological data (W-View)

    Energy Technology Data Exchange (ETDEWEB)

    Tachibana, Haruo; Sekita, Tsutomu; Yamaguchi, Takenori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    A computer code (W-View: Weather View) was developed to analyze the meteorological data statistically based on 'the guideline of meteorological statistics for the safety analysis of nuclear power reactor' (Nuclear Safety Commission on January 28, 1982; revised on March 29, 2001). The code gives statistical meteorological data to assess the public dose in case of normal operation and severe accident to get the license of nuclear reactor operation. This code was revised from the original code used in a large office computer code to enable a personal computer user to analyze the meteorological data simply and conveniently and to make the statistical data tables and figures of meteorology. (author)

  12. Benchmark Simulation for the Development of the Regulatory Audit Subchannel Analysis Code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, G. H.; Song, C.; Woo, S. W. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2012-05-15

    For the safe and reliable operation of a reactor, it is important to predict accurately the flow and temperature distributions in the thermal-hydraulic design of a reactor core. A subchannel approach can give the reasonable flow and temperature distributions with the short computing time. Korea Institute of Nuclear Safety (KINS) is presently reviewing new subchannel code, THALES, which will substitute for both THINC-IV and TORC code. To assess the prediction performance of THALES, KINS is developing the subchannel analysis code for the independent audit calculation. The code is based on workstation version of COBRA-IV-I. The main objective of the present study is to assess the performance of COBRA-IV-I code by comparing the simulation results with experimental ones for the sample problems

  13. Parametric systems analysis for ICF hybrid reactors

    International Nuclear Information System (INIS)

    Berwald, D.H.; Maniscalco, J.A.; Chapin, D.L.

    1981-01-01

    Parametric design and systems analysis for inertial confinement fusion-fission hybrids are presented. These results were generated as part of the Electric Power Research Institute (EPRI) sponsored Feasibility Assessment of Fusion-Fission Hybrids, using an Inertial Confinement Fusion (ICF) hybrid power plant design code developed in conjunction with the feasibility assessment. The SYMECON systems analysis code, developed by Westinghouse, was used to generate economic results for symbiotic electricity generation systems consisting of the hybrid and its client Light Water Reactors (LWRs). These results explore the entire fusion parameter space for uranium fast fission blanket hybrids, thorium fast fission blanket hybrids, and thorium suppressed fission blanket types are discussed, and system sensitivities to design uncertainties are explored

  14. Noise and vibration reduction technology in hybrid vehicle development; Hybrid sha kaihatsu ni okeru shindo soon teigen gijutsu

    Energy Technology Data Exchange (ETDEWEB)

    Yoshioa, T.; Sugita, H. [Toyota Motor Corp., Aichi (Japan)

    2000-03-01

    Accomplishing both environmental protection and good NVH performance has become a significant task in automotive development The first-in-the-world hybrid passenger car of mass production. 'Prius', has achieved superior NV performance compared with conventional vehicles with a 1.5-liter engine along with 50% reduction of fuel consumption and CO{sub 2} emissions. low HC, CO and NO{sub x} emissions. This paper describes NV reduction technology for solving problems peculiar to the hybrid vehicle such as engine start/stop vibration, drone noise at low engine speed and motor/generator noise and vibration. It also mentions application technology of low rolling resistance tires with light weight wheels and recycled material for sound proofing. (author)

  15. Development of chiral terminal-alkene-phosphine hybrid ligands for palladium-catalyzed asymmetric allylic substitutions.

    Science.gov (United States)

    Liu, Zhaoqun; Du, Haifeng

    2010-07-02

    A variety of novel chiral terminal-alkene-phosphine hybrid ligands were successfully developed from diethyl L-tartrate for palladium-catalyzed asymmetric allylic alkylations, etherifications, and amination to give the desired products in excellent yields and ee's.

  16. Development of a Fully-Automated Monte Carlo Burnup Code Monteburns

    International Nuclear Information System (INIS)

    Poston, D.I.; Trellue, H.R.

    1999-01-01

    Several computer codes have been developed to perform nuclear burnup calculations over the past few decades. In addition, because of advances in computer technology, it recently has become more desirable to use Monte Carlo techniques for such problems. Monte Carlo techniques generally offer two distinct advantages over discrete ordinate methods: (1) the use of continuous energy cross sections and (2) the ability to model detailed, complex, three-dimensional (3-D) geometries. These advantages allow more accurate burnup results to be obtained, provided that the user possesses the required computing power (which is required for discrete ordinate methods as well). Several linkage codes have been written that combine a Monte Carlo N-particle transport code (such as MCNP TM ) with a radioactive decay and burnup code. This paper describes one such code that was written at Los Alamos National Laboratory: monteburns. Monteburns links MCNP with the isotope generation and depletion code ORIGEN2. The basis for the development of monteburns was the need for a fully automated code that could perform accurate burnup (and other) calculations for any 3-D system (accelerator-driven or a full reactor core). Before the initial development of monteburns, a list of desired attributes was made and is given below. o The code should be fully automated (that is, after the input is set up, no further user interaction is required). . The code should allow for the irradiation of several materials concurrently (each material is evaluated collectively in MCNP and burned separately in 0RIGEN2). o The code should allow the transfer of materials (shuffling) between regions in MCNP. . The code should allow any materials to be added or removed before, during, or after each step in an automated fashion. . The code should not require the user to provide input for 0RIGEN2 and should have minimal MCNP input file requirements (other than a working MCNP deck). . The code should be relatively easy to use

  17. Online test application development using framework CodeIgniter

    Science.gov (United States)

    Wibawa, S. C.; Wahyuningsih, Y.; Sulistyowati, R.; Abidin, R.; Lestari, Y.; Noviyanti; Maulana, D. A.

    2018-01-01

    The purpose of this study is developing application an online test for vocational students and to know the user acceptance testing on the application. The method used in this research is the Research and Development (R & D) only up to the pilot phase of the product. The stage of the procedure of the research namely: (1) Analyze the exam using paper compared to using web-based application test online. (2) Design the media in accordance with the design of the author. (3) To test the product by including a questionnaire instrument against the application that has been done. Researchers carried out tests on class X on the computer and network engineering Vocational High School (SMK) Darul Ma’wa Plumpang. It can be concluded that: (1) application online test was created gets the value of the validator with the percentage of lowest value and the highest value for the validation of products: 25% and 100%. With a total number of 14 questions, after validation of the products obtained from the three aspects of the assessment scale from 81.25 to 100 obtained from 2 different validators with the meaning of an application that has been developed and very suitable for use in school. (2) Based on User Acceptance Testing (UAT), applications can be very well received by the students and recommend to replay the final semester and others. With the successful acquisition of a category which means it’s ready and qualified.

  18. Development of a Computer Code for the Estimation of Fuel Rod Failure

    Energy Technology Data Exchange (ETDEWEB)

    Rhee, I.H.; Ahn, H.J. [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    1997-12-31

    Much research has already been performed to obtain the information on the degree of failed fuel rods from the primary coolant activities of operating PWRs in the last few decades. The computer codes that are currently in use for domestic nuclear power plants, such as CADE code and ABB-CE codes developed by Westinghouse and ABB-CE, respectively, still give significant overall errors in estimating the failed fuel rods. In addition, with the CADE code, it is difficult to predict the degree of fuel rod failures during the transient period of nuclear reactor operation, where as the ABB-CE codes are relatively more difficult to use for end-users. In particular, the rapid progresses made recently in the area of the computer hardware and software systems that their computer programs be more versatile and user-friendly. While the MS windows system that is centered on the graphic user interface and multitasking is now in widespread use, the computer codes currently employed at the nuclear power plants, such as CADE and ABB-CE codes, can only be run on the DOS system. Moreover, it is desirable to have a computer code for the fuel rod failure estimation that can directly use the radioactivity data obtained from the on-line monitoring system of the primary coolant activity. The main purpose of this study is, therefore, to develop a Windows computer code that can predict the location, the number of failed fuel rods,and the degree of failures using the radioactivity data obtained from the primary coolant activity for PWRs. Another objective is to combine this computer code with the on-line monitoring system of the primary coolant radioactivity at Kori 3 and 4 operating nuclear power plants and enable their combined use for on-line evaluation of the number and degree of fuel rod failures. (author). 49 refs., 85 figs., 30 tabs.

  19. Development of ECOREA-II code for the evaluation of exposures from radionuclides through food Chain

    International Nuclear Information System (INIS)

    Yu, Dong Han; Lee, Han Soo

    2002-01-01

    The release of radionuclides from nuclear facilities following an accident into air results in human exposures by intakes of plant products such as rice, vegetables and/or animal products including meat, milk and eggs from contaminated soil. In order to evaluate such exposures from radioactive substances, it is essential to mathematically predict the behavior of these substances in the environments. A computer code, named 'ECOREA-II' is developing to assess human exposures through food chain of such substances in Korea. ECOREA-II code has a dynamic compartment-based model at its core, the graphical user interface (GUI) for the selection of input parameters and result displays on personal computers, and generation of data files for a GIS (Graphical Information System). Even the code is developed mostly based currently available models and/or codes, a new model is included for the time-dependent growth dilution in a vegetation part. Effort on The development of the code is towards the prediction of the behavior and pattern of radionuclides in a specific food chain condition in Korea. Finally, it provides a more user-friendly environment such as GUI developed based on the VBA(Visual Basic Application) for personal users. Therefore, the current code, when more fully developed, is expected to increase the understanding of environmental safety assessment of nuclear facilities following an accident and provide a reasonable regulatory guideline with respect to food safety issues

  20. Development and validation of the ENIGMA code for MOX fuel performance modelling

    International Nuclear Information System (INIS)

    Palmer, I.; Rossiter, G.; White, R.J.

    2000-01-01

    The ENIGMA fuel performance code has been under development in the UK since the mid-1980s with contributions made by both the fuel vendor (BNFL) and the utility (British Energy). In recent years it has become the principal code for UO 2 fuel licensing for both PWR and AGR reactor systems in the UK and has also been used by BNFL in support of overseas UO 2 and MOX fuel business. A significant new programme of work has recently been initiated by BNFL to further develop the code specifically for MOX fuel application. Model development is proceeding hand in hand with a major programme of MOX fuel testing and PIE studies, with the objective of producing a fuel modelling code suitable for mechanistic analysis, as well as for licensing applications. This paper gives an overview of the model developments being undertaken and of the experimental data being used to underpin and to validate the code. The paper provides a summary of the code development programme together with specific examples of new models produced. (author)

  1. PURE OR HYBRID? THE DEVELOPMENT OF MIXED DICTIONARY GENRES

    OpenAIRE

    Reinhard R. K. Hartmann

    2005-01-01

    This paper explores 'hybrid' genres of dictionaries and other reference works. Against the tradition of general dictionaries becoming ever more specialised, there has also been a growing trend of mixing two or more 'pure' dictionary types for achieving specific purposes, e.g. the combination of alphabetic and thematic dictionary,general dictionary and technical glossary, dictionary and thesaurus, dictionary and encyclopedia, monolingual and bilingual dictionary, etc. Examples of these various...

  2. Panorama 2011: The development of hybrid and electric vehicles

    International Nuclear Information System (INIS)

    Vinot, S.

    2011-01-01

    Car manufacturers are having to deal with increasingly stringent norms and customers who are increasingly demanding with respect to fuel savings. As a result, large numbers of them are now looking into solutions that involve electrifying their vehicles. Hybrid vehicles, some of which can be recharged, and electric vehicles are the new stars of the auto trade shows. But not all manufacturers are necessarily using the same strategies. (author)

  3. Development of hybrid system for mass productive passenger car; Joyoshayo ryosangata hybrid system no kaihatsu

    Energy Technology Data Exchange (ETDEWEB)

    Abe, S; Sasaki, S; Matsui, H; Kubo, K [Toyota Motor Corp., Aichi (Japan)

    1997-10-01

    The Toyota Hybrid System has two power sources which engage depending on driving conditions. An improved efficiency gasoline engine provides the main power to drive the wheels, as in conventional automobiles. Drive power can also be supplied by an electric motor, which derives its electricity from a battery and a Generator. Fuel efficiency is further boosted by other measures, such as automatically cutting the engine when the vehicle is at a stop and regenerating the energy accumulate during deceleration. Compared with conventional vehicles with a 1.5-liter engine running in 10 {center_dot} 15 mode, the CO2 emissions are reduced by half. Moreover, with the use of a motor to reduce engine load during acceleration, HC, CO, and NOx emissions are cut significantly low levels. 13 figs.

  4. A development and an application of Mixset-X computer code for simulating the Purex solvent extraction system

    International Nuclear Information System (INIS)

    Shida, M.; Naito, M.; Suto, T.; Omori, E.; Nojiri, T.

    2001-01-01

    MIXSET is a FORTRAN code developed to simulate the Purex solvent extraction system using mixer-settler extractors. Japan Nuclear Cycle Development Institute (JNC) has been developing the MIXSET code since the years 1970 to analyze the behavior of nuclides in the solvent extraction processes in Tokai Reprocessing Plant (TRP). This paper describes the history of MIXSET code development, the features of the latest version, called MIXSET-X and the application of the code for safety evaluation work. (author)

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  6. Development of aromatic VOC control technology by electron beam hybrid

    International Nuclear Information System (INIS)

    Kim, Jo-Chun; Kim, Ki-Joon

    2006-01-01

    As a fundamental study, the decomposition of volatile organic compounds (VOCs) using electron beam (EB) irradiation has been extensively investigated. EB treatments of VOCs such as toluene and styrene are discussed. The degradation characteristics were intensively investigated under various concentrations and irradiation doses to determine and improve VOC removal efficiencies. This work illustrates that the removal efficiencies of aromatic VOCs generally increase as their concentrations decrease and the irradiation doses increase. Based on these basic studies, it was found that by-products produced from EB irradiation of VOCs would cause a secondary pollution problem. Therefore, a novel hybrid technology has been applied to control aromatic VOC emissions by annexing the catalyst technique with conventional treatment study using EB technology. The experiments were carried out using a bench-scale at first, then a pilot-scale system was followed. Toluene was selected as a typical VOC for EB hybrid control to investigate by-products, effects of ceramic and catalyst, and factors affecting overall efficiency of degradation. It was concluded that VOCs could be destroyed more effectively by a novel hybrid system than single EB irradiation. (author)

  7. Development of a hybrid mode linear transformer driver stage

    Science.gov (United States)

    Zhang, Le; Wang, Meng; Zhou, Liangji; Tian, Qing; Guo, Fan; Wang, Lingyun; Qing, Yanling; Zhao, Yue; Dai, Yingmin; Han, Wenhui; Chen, Lin; Xie, Weiping

    2018-02-01

    At present, the mainstream technologies of primary power sources of large pulse power devices adopt Marx or linear transformer driver (LTD) designs. Based on the analysis of the characteristics of these two types of circuit topologies, the concept of a hybrid mode LTD stage based on Marx branches is proposed. The analysis shows that the hybrid mode LTD stage can realize the following goals: (a) to reduce the energy and power handled by the basic components (switch and capacitor) to lengthen their lifetime; (b) to reduce the requirements of the multipath synchronous trigger system; and (c) to improve the maintainability of the LTD stage by using independent Marx generators instead of "traditional LTD bricks." To verify the technique, a hybrid mode LTD stage consisting of 50 branches (four-stage compact Marx generators) was designed, manufactured and tested. The stage has a radius of about 3.3 m and a height of 0.6 m. The single Marx circuit's load current is about 21 kA, with a rise time of ˜90 ns (10%-90%), under the conditions of capacitors charged to ±40 kV and a 6.9 Ω matched load. The whole stage's load current is ˜1 MA , with a rise time of ˜112 ns (10%-90%), when the capacitors are charged to ±45 kV and the matched load is 0.14 Ω .

  8. Experiment-specific analyses in support of code development

    International Nuclear Information System (INIS)

    Ott, L.J.

    1990-01-01

    Experiment-specific models have been developed since 1986 by Oak Ridge National Laboratory Boiling Water Reactor (BWR) severe accident analysis programs for the purpose of BWR experimental planning and optimum interpretation of experimental results. These experiment-specific models have been applied to large integral tests (ergo, experiments) which start from an initial undamaged core state. The tests performed to date in BWR geometry have had significantly different-from-prototypic boundary and experimental conditions because of either normal facility limitations or specific experimental constraints. These experiments (ACRR: DF-4, NRU: FLHT-6, and CORA) were designed to obtain specific phenomenological information such as the degradation and interaction of prototypic components and the effects on melt progression of control-blade materials and channel boxes. Applications of ORNL models specific to the ACRR DF-4 and KfK CORA-16 experiments are discussed and significant findings from the experimental analyses are presented. 32 refs., 16 figs

  9. Practical Salesforce.com development without code customizing salesforce on the Force.com platform

    CERN Document Server

    Weinmeister, Philip

    2014-01-01

    Are you facing a challenging Salesforce.com problem-say, relating to customization, configuration, reporting, dashboards, or formulation-that you can't quite crack? Or maybe you are hoping to infuse some creativity into your solution design strategy to solve problems faster or make solutions more efficient? Practical Salesforce.com Development Without Code shows you how to unlock the power of the Force.com platform to solve real business problems-and all without writing a line of code. Adhering to Salesforce.com's ""Clicks, not code"" mantra, Salesforce.com expert Phil Weinmeister walks you t

  10. Development of the computer code system for the analyses of PWR core

    International Nuclear Information System (INIS)

    Tsujimoto, Iwao; Naito, Yoshitaka.

    1992-11-01

    This report is one of the materials for the work titled 'Development of the computer code system for the analyses of PWR core phenomena', which is performed under contracts between Shikoku Electric Power Company and JAERI. In this report, the numerical method adopted in our computer code system are described, that is, 'The basic course and the summary of the analysing method', 'Numerical method for solving the Boltzmann equation', 'Numerical method for solving the thermo-hydraulic equations' and 'Description on the computer code system'. (author)

  11. Strategies for developing subchannel capability in an advanced system thermalhydraulic code: a literature review

    International Nuclear Information System (INIS)

    Cheng, J.; Rao, Y.F.

    2015-01-01

    In the framework of developing next generation safety analysis tools, Canadian Nuclear Laboratories (CNL) has planned to incorporate subchannel analysis capability into its advanced system thermalhydraulic code CATHENA 4. This paper provides a literature review and an assessment of current subchannel codes. It also evaluates three code-development methods: (i) static coupling of CATHENA 4 with the subchannel code ASSERT-PV, (ii) dynamic coupling of the two codes, and (iii) fully implicit implementation for a new, standalone CATHENA 4 version with subchannel capability. Results of the review and assessment suggest that the current ASSERT-PV modules can be used as the base for the fully implicit implementation of subchannel capability in CATHENA 4, and that this option may be the most cost-effective in the long run, resulting in savings in user application and maintenance costs. In addition, improved versatility of the tool could be accomplished by the addition of new features that could be added as part of its development. The new features would improve the capabilities of the existing subchannel code in handling low, reverse, and stagnant flows often encountered in system thermalhydraulic analysis. Therefore, the method of fully implicit implementation is preliminarily recommended for further exploration. A feasibility study will be performed in an attempt to extend the present work into a preliminary development plan. (author)

  12. Development of a PC code package for the analysis of research and power reactors

    International Nuclear Information System (INIS)

    Urli, N.

    1992-06-01

    Computer codes available for performing reactor physics calculations for nuclear research reactors and power reactors are normally suited for running on mainframe computers. With the fast development in speed and memory of the PCs and affordable prices it became feasible to develop PC versions of commonly used codes. The present work performed under an IAEA sponsored research contract has successfully developed a code package for running on a PC. This package includes a cross-section generating code PSU-LEOPARD and 2D and 1D spatial diffusion codes, MCRAC and MCYC 1D. For adapting PSU-LEOPARD for a PC, the binary library has been reorganized to decimal form, upgraded to FORTRAN-77 standard and arrays and subroutines reorganized to conform to PC compiler. Similarly PC version of MCRAC for FORTRAN-77 and 1D code MCYC 1D have been developed. Tests, verification and bench mark results show excellent agreement with the results obtained from mainframe calculations. The execution speeds are also very satisfactory. 12 refs, 4 figs, 3 tabs

  13. Development of a thermal-hydraulic code for reflood analysis in a PWR experimental loop

    International Nuclear Information System (INIS)

    Alves, Sabrina P.; Mesquita, Amir Z.; Rezende, Hugo C.; Palma, Daniel A.P.

    2017-01-01

    A process of fundamental importance in the event of Loss of Coolant Accident (LOCA) in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Center (CDTN) has been developing since the 70’s programs to allow Brazil to become independent in the field of reactor safety analysis. To that end, in the 80’s was designed, assembled and commissioned one Rewetting Test Facility (ITR in Portuguese). This facility aims to investigate the phenomena involved in the thermal hydraulic reflood phase of a Loss of Coolant Accident in a PWR nuclear reactor. This work aim is the analysis of physical and mathematical models governing the rewetting phenomenon, and the development a thermo-hydraulic simulation code of a representative experimental circuit of the PWR reactors core cooling channels. It was possible to elaborate and develop a code called REWET. The results obtained with REWET were compared with the experimental results of the ITR, and with the results of the Hydroflut code, that was the old program previously used. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the front for two typical tests using the two codes calculation, and experimental results. The result simulated by REWET code for the rewetting time also came closer to the experimental results more than those calculated by Hydroflut code. (author)

  14. Development of a thermal-hydraulic code for reflood analysis in a PWR experimental loop

    Energy Technology Data Exchange (ETDEWEB)

    Alves, Sabrina P.; Mesquita, Amir Z.; Rezende, Hugo C., E-mail: sabrinapral@gmail.com, E-mail: amir@cdtn.brm, E-mail: hcr@cdtn.br, E-mail: hcr@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissão Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    A process of fundamental importance in the event of Loss of Coolant Accident (LOCA) in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Center (CDTN) has been developing since the 70’s programs to allow Brazil to become independent in the field of reactor safety analysis. To that end, in the 80’s was designed, assembled and commissioned one Rewetting Test Facility (ITR in Portuguese). This facility aims to investigate the phenomena involved in the thermal hydraulic reflood phase of a Loss of Coolant Accident in a PWR nuclear reactor. This work aim is the analysis of physical and mathematical models governing the rewetting phenomenon, and the development a thermo-hydraulic simulation code of a representative experimental circuit of the PWR reactors core cooling channels. It was possible to elaborate and develop a code called REWET. The results obtained with REWET were compared with the experimental results of the ITR, and with the results of the Hydroflut code, that was the old program previously used. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the front for two typical tests using the two codes calculation, and experimental results. The result simulated by REWET code for the rewetting time also came closer to the experimental results more than those calculated by Hydroflut code. (author)

  15. Development of eggplant hybrid cultivar ‘BATEM FILIZI’ and determination of yield performance

    Directory of Open Access Journals (Sweden)

    Hatice Filiz BOYACI

    2014-12-01

    Full Text Available Eggplant (Solanum melongena L. cultivation is an important part of vegetable production in Turkey. It is highly affected by abiotic stress factors such as low temperatures during growth and development. The present study was aimed to improve new hybrid varieties with high quality and fruit setting under undesirable environmental conditions in eggplant. The 39 inbred lines were developed from eggplant genepool by using pedigree breeding method at the Bati Akdeniz Agricultural Research Institute. They were characterized morphologically and evaluated for phylogenetic relationship among them. Selected seven lines were crossed each other to obtain F1 hybrids. Thirteen hybrids from these crosses were examined for their rate of heterosis and heterobeltiosis. These hybrids were not suitable for market demands although F1 hybrids including P350 as parental line had high heterosis rate. Oval fruit shaped P350 inbred line as female parent showed good performance and crossed with long fruit shaped P599 inbred line. The resultant F1 hybrid cultivar higher yielding, suitable for production under greenhouse and highly adaptable was registered with the name of BATEM FILIZI. It is recommended that this hybrid can be used in greenhouse production especially in single crop season.

  16. Development of a power train for the hybrid automobile - the Civic IMA

    Energy Technology Data Exchange (ETDEWEB)

    Matsuki, Masato; Sato, Toshiyuki; Wakashiro, Teruo; Kaku, Toshiaki; Kamiyama, Toshihiro; Kanda, Masahiro [Tochigi R and D Center (Japan); Brachmann, T. [Tochigi Offenbach R und D Center (Germany)

    2003-07-01

    The Civic Hybrid was developed as a compact passenger hybrid car that achieves both low fuel consumption and cleaner operation from the viewpoints of preserving the global environment and conserving resources. The engine has been improved for Hybrid applications, which were added to the base i-DSI, 4-cylinder, 1.3-liter SOHC, 2-ignition plugs/cylinder engine mounted in the Honda 'Jazz'. In addition, the cylinder idling system has been adopted to increase the regenerated energy during deceleration. The hybrid system is based on the Honda IMA system, and the maximum regenerative torque has been increased by approximately 30% by improving the magnetic circuits of an ultra-thin DC brushless motor and adopting a new rotor manufacturing method. Fuel economy is improved by a new hybrid power train, thus achieving low fuel consumption of 4.9 1/100 km in the European UDC+EUCD combined mode by at the same time meeting EURO IV standards. The power control unit, which is the IMA system control unit, was downsized and located behind the rear seat, thus ensuring comparable trunk capacity to the base vehicle of the Civic 4-Door. Hybrid vehicles have a lot to offer. This paper introduces evolutionary developments of Hybrid vehicles within the Honda Motor Company. (orig.)

  17. Conceptual Approach to Forming the Basic Code of Neo-Industrial Development of a Region

    Directory of Open Access Journals (Sweden)

    Elena Leonidovna Andreeva

    2017-09-01

    Full Text Available In the article, the authors propose the conceptual fundamentals of the “code approach” to the regional neo-industrial development. The purpose of the research is to reveal the essence of the transition to a new type of industrial and economic relations through a prism of “genetic codes” of the region. We consider these codes as a system of the “racial memory” of a territory, which determines the specificity and features of neo-industrialization realization. We substantiated the hypothesis about the influence of the “genetic codes” of the region on the effectiveness of the neo-industrialization. We have defined the participants, or else the carriers of the codes in the transformation of regional inheritance for the stimulation of the neoindustrial development of region’s economy. The subject matter of the research is the distinctive features of the functioning of the determinative region’s codes. Their content determines the socio-economic specificity of the region and the features of innovative, informational, value-based and competence-based development of the territory. The determinative codes generate the dynamic codes of the region, which are understood as their derivatives. They have a high probability of occurrence, higher speed of development and distribution, internal forces that make possible the self-development of the region. The scientific contribution is the substantiation of the basic code of the regional neo-industrial development. It represents the evolutionary accumulation of the rapid changes of its innovative, informational, value-based and competence-based codes stimulating the generation and implementation of new ideas regarding to economic entities adapted to the historical and cultural conditions. The article presents the code model of neo-industrial development of the region described by formulas. We applied the system analysis methods, historical and civilization approaches, evolutionary and

  18. CODE SWITCHING AND THE DEVELOPMENT OF LINGUISTIC SYSTEM OF SIMULTANEOUS BILINGUAL CHILDREN

    Directory of Open Access Journals (Sweden)

    Leni Amelia Suek

    2017-11-01

    Full Text Available Code switching and code mixing are the phenomena commonly seen done by a bilingual. This behavior is influenced by several aspects such as the linguistic system, sociolinguistics, pragmatics, and language competence of the bilingual. If children are able to distinguish two different languages since early age, they will be considered simultaneous bilinguals. They show that they develop multiple, rather than single, linguistic systems. However, it was understood that code switching and code mixing were due to the failure in using proper words, language features, and sociolinguistic competence. Yet, recent studies have shown that bilingual children are able to use both languages proficiently with no signs of confusion or failure in language use. This ability also does not hinder their cognitive development.

  19. Development of OCDMA system based on Flexible Cross Correlation (FCC) code with OFDM modulation

    Science.gov (United States)

    Aldhaibani, A. O.; Aljunid, S. A.; Anuar, M. S.; Arief, A. R.; Rashidi, C. B. M.

    2015-03-01

    The performance of the OCDMA systems is governed by numerous quantitative parameters such as the data rate, simultaneous number of users, the powers of transmitter and receiver, and the type of codes. This paper analyzes the performance of the OCDMA system using OFDM technique to enhance the channel data rate, to save power and increase the number of user of OSCDMA systems compared with previous hybrid subcarrier multiplexing/optical spectrum code division multiplexing (SCM/OSCDM) system. The average received signal to noise ratio (SNR) with the nonlinearity of subcarriers is derived. The theoretical results have been evaluated based on BER and number of users as well as amount of power saved. The proposed system gave better performance and save around -6 dBm of the power as well as increase the number of users twice compare to SCM/OCDMA system. In addition it is robust against interference and much more spectrally efficient than SCM/OCDMA system. The system was designed based on Flexible Cross Correlation (FCC) code which is easier construction, less complexity of encoder/decoder design and flexible in-phase cross-correlation for uncomplicated to implement using Fiber Bragg Gratings (FBGs) for the OCDMA systems for any number of users and weights. The OCDMA-FCC_OFDM improves the number of users (cardinality) 108% compare to SCM/ODCMA-FCC system.

  20. A New Hybrid Power Configuration Used In HEV And Development of Its Driving BLDC

    Directory of Open Access Journals (Sweden)

    Yanliang Xu

    2013-07-01

    Full Text Available Hybrid electric vehicle (HEV that has the advantages of high performance, high fuel efficiency, low emissions, and long operating range is focused on nowadays. A new hybrid power configuration used in HEV is presented with different control strategies and HEV performances when equipping different weight of fuel power and battery one. In order to realize the new hybrid power configuration especially in the refitted HEV, fractional-slot concentrated-wingding BLDC with higher rated spinning speed is given and the prototyped one is developed and fabricated which satisfy successfully the requirements of high performance and demission restriction for the refitted HEV.