Saphyr: a code system from reactor design to reference calculations
Energy Technology Data Exchange (ETDEWEB)
Akherraz, B.; Baudron, A.M.; Buiron, L.; Coste-Delclaux, M.; Fedon-Magnaud, C.; Lautard, J.J.; Moreau, F.; Nicolas, A.; Sanchez, R.; Zmijarevic, I. [CEA Saclay, Direction de l' Energie Nucleaire, Departement de Modelisation des Systemes et Structures, Service d' Etudes des Reacteurs et de Modelisation Avancee (DENDMSS/SERMA), 91 - Gif sur Yvette (France); Bergeron, A.; Caruge, D.; Fillion, P.; Gallo, D.; Royer, E. [CEA Saclay, Direction de l' Energie Nucleaire, Departement de Modelisation des Systemes et Structures, Service Fluides numeriques, Modelisations et Etudes (DEN/DMSS/SFNME), 91 - Gif sur Yvette (France); Loubiere, S. [CEA Saclay, Direction de l' Energie Nucleaire, Direction de la Simulation et des Outils Experimentaux, 91- Gif sur Yvette (France)
2003-07-01
In this paper we briefly present the package SAPHYR (in French Advanced System for Reactor Physics) which is devoted to reactor calculations, safety analysis and design. This package is composed of three main codes: APOLLO2 for lattice calculations, CRONOS2 for whole core neutronic calculations and FLICA4 for thermohydraulics. Thanks to a continuous development effort, the SAPHYR system is an outstanding tool covering a large domain of applications, from sophisticated 'research and development' studies that need state-of-the-art methodology to routine industrial calculations for reactor and criticality analysis. SAPHYR is powerful enough to carry out calculations for all types of reactors and is invaluable to understand complex phenomena. SAPHYR components are in use in various nuclear companies such as 'Electricite de France', Framatome-ANP, Cogema, SGN, Transnucleaire and Technicatome. Waiting for the next generation tools (DESCARTES for neutronics and NEPTUNE for thermohydraulics) to be available for such a variety of use, with a better level of flexibility and at least equivalent validation and qualification level, the improvement of SAPHYR is going on, to acquire new functions constantly required by users and to improve current performance levels.
Fuel management and core design code systems for pressurized water reactor neutronic calculations
Energy Technology Data Exchange (ETDEWEB)
Ahnert, C.; Arayones, J.M.
1985-06-01
A package of connected code systems for the neutronic calculations relevant in fuel management and core design has been developed and applied for validation to the startup tests and first operating cycle of a 900MW (electric) PWR. The package includes the MARIA code system for the modeling of the different types of PWR fuel assemblies, the CARMEN code system for detailed few group diffusion calculations for PWR cores at operating and burnup conditions, and the LOLA code system for core simulation using onegroup nodal theory parameters explicitly calculated from the detailed solutions.
Core Calculation of 1 MWatt PUSPATI TRIGA Reactor (RTP) using Monte Carlo MVP Code System
Karim, Julia Abdul
2008-05-01
The Monte Carlo MVP code system was adopted for the Reaktor TRIGA PUSAPTI (RTP) core calculation. The code was developed by a group of researcher of Japan Atomic Energy Agency (JAEA) first in 1994. MVP is a general multi-purpose Monte Carlo code for neutron and photon transport calculation and able to estimate an accurate simulation problems. The code calculation is based on the continuous energy method. This code is capable of adopting an accurate physics model, geometry description and variance reduction technique faster than conventional method as compared to the conventional scalar method. This code could achieve higher computational speed by several factors on the vector super-computer. In this calculation, RTP core was modeled as close as possible to the real core and results of keff flux, fission densities and others were obtained.
DIST: a computer code system for calculation of distribution ratios of solutes in the purex system
Energy Technology Data Exchange (ETDEWEB)
Tachimori, Shoichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1996-05-01
Purex is a solvent extraction process for reprocessing the spent nuclear fuel using tri n-butylphosphate (TBP). A computer code system DIST has been developed to calculate distribution ratios for the major solutes in the Purex process. The DIST system is composed of database storing experimental distribution data of U(IV), U(VI), Pu(III), Pu(IV), Pu(VI), Np(IV), Np(VI), HNO{sub 3} and HNO{sub 2}: DISTEX and of Zr(IV), Tc(VII): DISTEXFP and calculation programs to calculate distribution ratios of U(IV), U(VI), Pu(III), Pu(IV), Pu(VI), Np(IV), Np(VI), HNO{sub 3} and HNO{sub 2}(DIST1), and Zr(IV), Tc(VII)(DITS2). The DIST1 and DIST2 determine, by the best-fit procedures, the most appropriate values of many parameters put on empirical equations by using the DISTEX data which fulfill the assigned conditions and are applied to calculate distribution ratios of the respective solutes. Approximately 5,000 data were stored in the DISTEX and DISTEXFP. In the present report, the following items are described, 1) specific features of DIST1 and DIST2 codes and the examples of calculation 2) explanation of databases, DISTEX, DISTEXFP and a program DISTIN, which manages the data in the DISTEX and DISTEXFP by functions as input, search, correction and delete. and at the annex, 3) programs of DIST1, DIST2, and figure-drawing programs DIST1G and DIST2G 4) user manual for DISTIN. 5) source programs of DIST1 and DIST2. 6) the experimental data stored in the DISTEX and DISTEXFP. (author). 122 refs.
Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system
Energy Technology Data Exchange (ETDEWEB)
Iga, Kiminori [Kyushu Univ., Fukuoka (Japan); Takada, Hiroshi; Nagao, Tadashi
1998-01-01
In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B{sub 4}C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)
Calculation of the Novovoronezh Recriticality Experiment with the KARATE-440 code system
Energy Technology Data Exchange (ETDEWEB)
Hegyi, György, E-mail: ghegyi@aeki.kfki.hu [MTA KFKI Atomic Energy Research Institute, Budapest (Hungary)
2011-07-01
In this paper the results of KARATE-440 calculations on Novovoronezh NPP Recriticality Experiment are presented, the corresponding parameters are analyzed. The simulation of the processes and the comparison of the results with the measurements are of particular interest as these efforts make our code to be validated in a higher level. The KARATE-440 code system has been developed and applied for VVER-440 core analysis during near twenty years, as a close collaboration among the developers and the specialists at the 4 Hungarian nuclear power units. KARATE is now a mature, demonstrated, complete and integrated system of computer codes and procedures that provide full and independent VVER core analysis capabilities. Even if only some well defined states of the experiment were simulated, satisfactory agreement was found between measured and calculated data. The results present evidence that the KARATE- 440 code package can adequately model the reactor states in a wide range of performance parameters and the special core type referred in the experiment so it is acceptable for neutronic analysis of all the VVER-440 NPP's. (author)
Decay heat experiment and validation of calculation code systems for fusion reactor
Energy Technology Data Exchange (ETDEWEB)
Maekawa, Fujio; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Wada, Masayuki
1999-10-01
Although accurate estimation of decay heat value is essential for safety analyses of fusion reactors against loss of coolant accidents and so on, no experimental work has been devoted to validating the estimation. Hence, a decay heat measurement experiment was performed as a task (T-339) of ITER/EDA. A new detector, the Whole Energy Absorption Spectrometer (WEAS), was developed for accurate and efficient measurements of decay heat. Decay heat produced in the thirty-two sample materials which were irradiated by 14-MeV neutrons at FNS/JAERI were measured with WEAS for a wide cooling time period from 1 min to 400 days. The data presently obtained were the first experimental decay heat data in the field of fusion. Validity of decay heat calculation codes of ACT4 and CINAC-V4, activation cross section libraries of FENDL/A-2.0 and JENDL Activation File, and decay data was investigated through analyses of the experiment. As a result, several points that should be modified were found in the codes and data. After solving the problems, it was demonstrated that decay heat valued calculated for most of samples were in good agreement with the experimental data. Especially for stainless steel 316 and copper, which were important materials for ITER, decay heat could be predicted with accuracy of {+-}10%. (author)
The MARS15-based FermiCORD Code System for Calculation of the Accelerator-Induced Residual Dose
Energy Technology Data Exchange (ETDEWEB)
Grebe, A.; Leveling, A.; Lu, T.; Mokhov, N.; Pronskikh, V.
2016-09-01
The FermiCORD code system, a set of codes based on MARS15 that calculates the accelerator-induced residual doses at experimental facilities of arbitrary configurations, has been developed. FermiCORD is written in C++ as an add-on to Fortran-based MARS15. The FermiCORD algorithm consists of two stages: 1) simulation of residual doses on contact with the surfaces surrounding the studied location and of radionuclide inventories in the structures surrounding those locations using MARS15, and 2) simulation of the emission of the nuclear decay gamma-quanta by the residuals in the activated structures and scoring the prompt doses of these gamma-quanta at arbitrary distances from those structures. The FermiCORD code system has been benchmarked against similar algorithms based on other code systems and showed a good agreement. The code system has been applied for calculation of the residual dose of the target station for the Mu2e experiment and the results have been compared to approximate dosimetric approaches.
Decay heat measurement on fusion reactor materials and validation of calculation code system
Energy Technology Data Exchange (ETDEWEB)
Maekawa, Fujio; Ikeda, Yujiro; Wada, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1998-03-01
Decay heat rates for 32 fusion reactor relevant materials irradiated with 14-MeV neutrons were measured for the cooling time period between 1 minute and 400 days. With using the experimental data base, validity of decay heat calculation systems for fusion reactors were investigated. (author)
Energy Technology Data Exchange (ETDEWEB)
Christine Bals; Henrique Austregesilo; Klaus Trambauer [Gesellschaft fuer Anlagen- und Reaktorsicherheit - GRS - mbH Forschungsinstitute, 85748 Garching (Germany)
2005-07-01
Full text of publication follows: The QUENCH fuel bundle experiments, performed at the Forschungszentrum Karlsruhe in Germany, aim to investigate the hydrogen source term and the bundle degradation during reflood of an overheated reactor core. The test QUENCH-07, in which the bundle was cooled from high temperatures by steam injected from the bottom, was the first experiment in this test series with a boron carbide absorber rod in the bundle. One major objective of this test was to provide information on the B{sub 4}C/SS/Zry interactions, on the formation of gaseous reaction products during B{sub 4}C oxidation and control rod degradation, and on the impact of control rod degradation on surrounding rods. In the general frame of developmental assessment and code validation, a post-test calculation of test QUENCH-07 complemented by an uncertainty analysis was performed with the code ATHLET-CD. The system code ATHLET-CD is being developed for realistic simulation of accidents with core degradation and for evaluation of accident management procedures. It applies the detailed and validated models of the thermal-hydraulic code ATHLET in an efficient coupling with dedicated models for core degradation, including mechanical rod behaviour, zirconium and B{sub 4}C oxidation, melting and relocation of metallic and ceramic components, and for the release and transport of fission products and aerosols. The first step of the work was the simulation of the QUENCH-07 experiment, applying the modeling options recommended in the code User's Manual (reference calculation). The global results of this calculation, mainly with respect to the hydrogen release rate and to the time evolution of bundle temperatures in different elevations, showed a good agreement with the measured data. This calculation was complemented by a sensitivity analysis in order to investigate the influence of a combined variation of code input parameters on the simulation of the main phenomena observed in the
Energy Technology Data Exchange (ETDEWEB)
Panka, Istvan; Kereszturi, Andras [Hungarian Academy of Sciences, Budapest (Hungary). Reactor Analysis Dept.
2013-09-15
In this paper the uncertainties of the neutronic calculations at core level - originating from the uncertainties of the basic nuclear data - are presented. The investigations have been made for a VVER-1000 core (Kozloduy-6) defined in the frame of the OECD NEA UAM benchmark. In the first part of the paper, the uncertainties of the effective multiplication factor, the assembly-wise radial power distribution, the axial power distribution and the rod worth are shown. After that the preliminary evaluation of the uncertainties of the neutron kinetic calculations are presented for a rod movement transient at HZP (Hot Zero Power) state, where the uncertainties of the time dependent core and assembly powers and the dynamic reactivity were evaluated. In both cases, we will see that the most important quantities - at core level and at HZP state - have a considerable uncertainty which is originating from the uncertainties of the basic cross section library in these investigations. (orig.)
The Flutter Shutter Code Calculator
Directory of Open Access Journals (Sweden)
Yohann Tendero
2015-08-01
Full Text Available The goal of the flutter shutter is to make uniform motion blur invertible, by a"fluttering" shutter that opens and closes on a sequence of well chosen sub-intervals of the exposure time interval. In other words, the photon flux is modulated according to a well chosen sequence calledflutter shutter code. This article provides a numerical method that computes optimal flutter shutter codes in terms of mean square error (MSE. We assume that the observed objects follow a known (or learned random velocity distribution. In this paper, Gaussian and uniform velocity distributions are considered. Snapshots are also optimized taking the velocity distribution into account. For each velocity distribution, the gain of the optimal flutter shutter code with respectto the optimal snapshot in terms of MSE is computed. This symmetric optimization of theflutter shutter and of the snapshot allows to compare on an equal footing both solutions, i.e. camera designs. Optimal flutter shutter codes are demonstrated to improve substantially the MSE compared to classic (patented or not codes. A numerical method that permits to perform a reverse engineering of any existing (patented or not flutter shutter codes is also describedand an implementation is given. In this case we give the underlying velocity distribution fromwhich a given optimal flutter shutter code comes from. The combination of these two numerical methods furnishes a comprehensive study of the optimization of a flutter shutter that includes a forward and a backward numerical solution.
Ferrero, M.; Rérat, Michel; Orlando, R.; Dovesi, R.
2007-12-01
The Coupled Perturbed Hartree-Fock (CPHF) scheme has been implemented in CRYSTAL06, periodic ab initio computer code that uses a gaussian type basis set, in order to obtain the polarizability of crystalline systems (3D), slabs (2D), polymers (1D) and, as a limiting case, molecules (0D). The formulation is presented, together with applications referring to diamond, silicon and SiC. It turns out that, for a given hamiltonian and basis set, high numerical accuracy can be achieved at relatively low cost compared to the Finite Field perturbation method (FF), that uses a saw-tooth electric potential. Correctness of the CPHF results was tested with reference to FF calculations with CRYSTAL.
Energy Technology Data Exchange (ETDEWEB)
Cicero, R., E-mail: ciceror@unican.e [INESCO INGENIEROS S.L., Santander (Spain); Departamento de Ciencia e Ingenieria del Terreno y los Materiales, Universidad de Cantabria, Santander (Spain); Cicero, S. [Departamento de Ciencia e Ingenieria del Terreno y los Materiales, Universidad de Cantabria, Santander (Spain); Gorrochategui, I. [Centro Tecnologico de Componentes, Santander (Spain); Lacalle, R. [INESCO INGENIEROS S.L., Santander (Spain); Departamento de Ciencia e Ingenieria del Terreno y los Materiales, Universidad de Cantabria, Santander (Spain)
2010-01-15
Nuclear power plants are generally designed and inspected according to the ASME Code. This code indicates stress intensity (S{sub INT}) as the parameter to be used in the stress analysis of components. One of the particularities of S{sub INT} is that it always takes positive values, independently of the nature of the stress (tensile or compressive). This circumstance is relevant in the Fatigue Monitoring Systems used in nuclear power plants, due to the manner in which the different variable stresses are combined in order to obtain the final total stress range. This paper describes some situations derived from the application of the ASME Code, shows different ways of dealing with them and illustrates their influence on the evaluation of the fatigue usage factor through a case study.
Energy Technology Data Exchange (ETDEWEB)
Behler, Matthias; Hannstein, Volker; Kilger, Robert; Moser, Franz-Eberhard; Pfeiffer, Arndt; Stuke, Maik
2014-06-15
In order to account for the reactivity-reducing effect of burn-up in the criticality safety analysis for systems with irradiated nuclear fuel (''burnup credit''), numerical methods to determine the enrichment and burnup dependent nuclide inventory (''burnup code'') and its resulting multiplication factor k{sub eff} (''criticality code'') are applied. To allow for reliable conclusions, for both calculation systems the systematic deviations of the calculation results from the respective true values, the bias and its uncertainty, are being quantified by calculation and analysis of a sufficient number of suitable experiments. This quantification is specific for the application case under scope and is also called validation. GRS has developed a methodology to validate a calculation system for the application of burnup credit in the criticality safety analysis for irradiated fuel assemblies from pressurized water reactors. This methodology was demonstrated by applying the GRS home-built KENOREST burnup code and the criticality calculation sequence CSAS5 from SCALE code package. It comprises a bounding approach and alternatively a stochastic, which both have been exemplarily demonstrated by use of a generic spent fuel pool rack and a generic dry storage cask, respectively. Based on publicly available post irradiation examination and criticality experiments, currently the isotopes of uranium and plutonium elements can be regarded for.
INVAP's Nuclear Calculation System
Directory of Open Access Journals (Sweden)
Ignacio Mochi
2011-01-01
Full Text Available Since its origins in 1976, INVAP has been on continuous development of the calculation system used for design and optimization of nuclear reactors. The calculation codes have been polished and enhanced with new capabilities as they were needed or useful for the new challenges that the market imposed. The actual state of the code packages enables INVAP to design nuclear installations with complex geometries using a set of easy-to-use input files that minimize user errors due to confusion or misinterpretation. A set of intuitive graphic postprocessors have also been developed providing a fast and complete visualization tool for the parameters obtained in the calculations. The capabilities and general characteristics of this deterministic software package are presented throughout the paper including several examples of its recent application.
TEA: A CODE CALCULATING THERMOCHEMICAL EQUILIBRIUM ABUNDANCES
Energy Technology Data Exchange (ETDEWEB)
Blecic, Jasmina; Harrington, Joseph; Bowman, M. Oliver, E-mail: jasmina@physics.ucf.edu [Planetary Sciences Group, Department of Physics, University of Central Florida, Orlando, FL 32816-2385 (United States)
2016-07-01
We present an open-source Thermochemical Equilibrium Abundances (TEA) code that calculates the abundances of gaseous molecular species. The code is based on the methodology of White et al. and Eriksson. It applies Gibbs free-energy minimization using an iterative, Lagrangian optimization scheme. Given elemental abundances, TEA calculates molecular abundances for a particular temperature and pressure or a list of temperature–pressure pairs. We tested the code against the method of Burrows and Sharp, the free thermochemical equilibrium code Chemical Equilibrium with Applications (CEA), and the example given by Burrows and Sharp. Using their thermodynamic data, TEA reproduces their final abundances, but with higher precision. We also applied the TEA abundance calculations to models of several hot-Jupiter exoplanets, producing expected results. TEA is written in Python in a modular format. There is a start guide, a user manual, and a code document in addition to this theory paper. TEA is available under a reproducible-research, open-source license via https://github.com/dzesmin/TEA.
Energy Technology Data Exchange (ETDEWEB)
Boyarinov, V.F.; Davidenko, V.D.; Polismakov, A.A.; Tsybulsky, V.F. [RRC Kurchatov Institute, Moscow (Russian Federation)
2005-07-01
At the present time, the new code system SUHAM-U for calculation of the neutron-physical processes in nuclear reactor core with triangular and square lattices based both on the modern micro-group (about 7000 groups) cross-sections library of code system UNK and on solving the multigroup (up to 89 groups) neutron transport equation by Surface Harmonics Method is elaborated. In this paper the procedure for generation of multigroup cross-sections from micro-group ones for calculation of VVER-1000 reactor core with MOX loading is described. The validation has consisted in computing VVER-1000 fuel assemblies with uranium and MOX fuel and has shown enough high accuracy under corresponding selection of the number and boundaries of the energy groups. This work has been fulfilled in the frame of ISTC project 'System Analyses of Nuclear Safety for VVER Reactors with MOX Fuels'.
Energy Technology Data Exchange (ETDEWEB)
Blazy-Aubignac, L
2007-09-15
The treatment planning systems (T.P.S.) occupy a key position in the radiotherapy service: they realize the projected calculation of the dose distribution and the treatment duration. Traditionally, the quality control of the calculated distribution doses relies on their comparisons with dose distributions measured under the device of treatment. This thesis proposes to substitute these dosimetry measures to the profile of reference dosimetry calculations got by the Penelope Monte-Carlo code. The Monte-Carlo simulations give a broad choice of test configurations and allow to envisage a quality control of dosimetry aspects of T.P.S. without monopolizing the treatment devices. This quality control, based on the Monte-Carlo simulations has been tested on a clinical T.P.S. and has allowed to simplify the quality procedures of the T.P.S.. This quality control, in depth, more precise and simpler to implement could be generalized to every center of radiotherapy. (N.C.)
Energy Technology Data Exchange (ETDEWEB)
Mazurier, J
1999-05-28
This thesis has been performed in the framework of national reference setting-up for absorbed dose in water and high energy photon beam provided with the SATURNE-43 medical accelerator of the BNM-LPRI (acronym for National Bureau of Metrology and Primary standard laboratory of ionising radiation). The aim of this work has been to develop and validate different user codes, based on PENELOPE Monte Carlo code system, to determine the photon beam characteristics and calculate the correction factors of reference dosimeters such as Fricke dosimeters and graphite calorimeter. In the first step, the developed user codes have permitted the influence study of different components constituting the irradiation head. Variance reduction techniques have been used to reduce the calculation time. The phase space has been calculated for 6, 12 and 25 MV at the output surface level of the accelerator head, then used for calculating energy spectra and dose distributions in the reference water phantom. Results obtained have been compared with experimental measurements. The second step has been devoted to develop an user code allowing calculation correction factors associated with both BNM-LPRI's graphite and Fricke dosimeters thanks to a correlated sampling method starting with energy spectra obtained in the first step. Then the calculated correction factors have been compared with experimental and calculated results obtained with the Monte Carlo EGS4 code system. The good agreement, between experimental and calculated results, leads to validate simulations performed with the PENELOPE code system. (author)
Data calculation program for RELAP 5 code
Energy Technology Data Exchange (ETDEWEB)
Silvestre, Larissa J.B.; Sabundjian, Gaiane, E-mail: larissajbs@usp.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)
2015-07-01
As the criteria and requirements for a nuclear power plant are extremely rigid, computer programs for simulation and safety analysis are required for certifying and licensing a plant. Based on this scenario, some sophisticated computational tools have been used such as the Reactor Excursion and Leak Analysis Program (RELAP5), which is the most used code for the thermo-hydraulic analysis of accidents and transients in nuclear reactors. A major difficulty in the simulation using RELAP5 code is the amount of information required for the simulation of thermal-hydraulic accidents or transients. The preparation of the input data leads to a very large number of mathematical operations for calculating the geometry of the components. Therefore, a mathematical friendly preprocessor was developed in order to perform these calculations and prepare RELAP5 input data. The Visual Basic for Application (VBA) combined with Microsoft EXCEL demonstrated to be an efficient tool to perform a number of tasks in the development of the program. Due to the absence of necessary information about some RELAP5 components, this work aims to make improvements to the Mathematic Preprocessor for RELAP5 code (PREREL5). For the new version of the preprocessor, new screens of some components that were not programmed in the original version were designed; moreover, screens of pre-existing components were redesigned to improve the program. In addition, an English version was provided for the new version of the PREREL5. The new design of PREREL5 contributes for saving time and minimizing mistakes made by users of the RELAP5 code. The final version of this preprocessor will be applied to Angra 2. (author)
Exposure calculation code module for reactor core analysis: BURNER
Energy Technology Data Exchange (ETDEWEB)
Vondy, D.R.; Cunningham, G.W.
1979-02-01
The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.
Electrical Conductivity Calculations from the Purgatorio Code
Energy Technology Data Exchange (ETDEWEB)
Hansen, S B; Isaacs, W A; Sterne, P A; Wilson, B G; Sonnad, V; Young, D A
2006-01-09
The Purgatorio code [Wilson et al., JQSRT 99, 658-679 (2006)] is a new implementation of the Inferno model describing a spherically symmetric average atom embedded in a uniform plasma. Bound and continuum electrons are treated using a fully relativistic quantum mechanical description, giving the electron-thermal contribution to the equation of state (EOS). The free-electron density of states can also be used to calculate scattering cross sections for electron transport. Using the extended Ziman formulation, electrical conductivities are then obtained by convolving these transport cross sections with externally-imposed ion-ion structure factors.
Research on Primary Shielding Calculation Source Generation Codes
Zheng Zheng; Mei Qiliang; Li Hui; Shangguan Danhua; Zhang Guangchun
2017-01-01
Primary Shielding Calculation (PSC) plays an important role in reactor shielding design and analysis. In order to facilitate PSC, a source generation code is developed to generate cumulative distribution functions (CDF) for the source particle sample code of the J Monte Carlo Transport (JMCT) code, and a source particle sample code is deveoped to sample source particle directions, types, coordinates, energy and weights from the CDFs. A source generation code is developed to transform three di...
Development of Reactivity Calculation Code for HANARO
Energy Technology Data Exchange (ETDEWEB)
Park, B. G.; Kim, M. S. [KAERI, Daejeon (Korea, Republic of)
2016-05-15
Reactivity of the reactor core is measured by a multi-channel wide range reactivity computer (or called reactivity meter), which uses current signals from the compensated ion chamber (CIC) mounted on the courtside wall of the reflector tank in the pool [1]. Because there were a few difficulties in operating the reactivity meter in the MS-DOS environment, some researches have been carried out to improve and upgrade it on the Windows environment [2]. Nevertheless, it is still hard for reactor operators to immediately check the time-dependent reactivity in case of power excursion because of some limitations such as aging of devices and compatibility issues. In this study, a simple off-line tool which can estimate the timedependent reactivity by using the fission chamber signals has been developed, and utilized to the case of loose parts of dummy rod in the 86-2th cycles of HANARO. In order to check the reactivity quickly for reactor operators, the inverse kinetic quations have been incorporated, and several useful functions have been implemented to the code. in the case of 86-2th cycles of HANARO, the developed code showed good performance to estimate time dependent reactivity. In the future, the on line analysis modules will be implanted to the code with upgrade of the measrement equipment such as current meters and data acquisition devices. Additionally, reactivity will be estimated by using the reactivity meter in the MS-DOS enviroment, and the new Windows version, for the verification of the developed code.
Quasiparticle GW calculations within the GPAW electronic structure code
DEFF Research Database (Denmark)
Hüser, Falco
are explained in detail and many examples are given. This provides a full understanding of how the code works and how the outcome should be interpreted. Secondly, it gives an extensive discussion of calculated results for the electronic structure of 3-dimensional, 2-dimensional and finite systems and comparison......The GPAW electronic structure code, developed at the physics department at the Technical University of Denmark, is used today by researchers all over the world to model the structural, electronic, optical and chemical properties of materials. They address fundamental questions in material science...... and use their knowledge to design new materials for a vast range of applications. Todays hottest topics are, amongst many others, better materials for energy conversion (e.g. solar cells), energy storage (batteries) and catalysts for the removal of environmentally dangerous exhausts. The mentioned...
Calculating Reuse Distance from Source Code
Energy Technology Data Exchange (ETDEWEB)
Narayanan, Sri Hari Krishna; Hovland, Paul
2016-01-26
The efficient use of a system is of paramount importance in high-performance computing. Applications need to be engineered for future systems even before the architecture of such a system is clearly known. Static performance analysis that generates performance bounds is one way to approach the task of understanding application behavior. Performance bounds provide an upper limit on the performance of an application on a given architecture. Predicting cache hierarchy behavior and accesses to main memory is a requirement for accurate performance bounds. This work presents our static reuse distance algorithm to generate reuse distance histograms. We then use these histograms to predict cache miss rates. Experimental results for kernels studied show that the approach is accurate.
Gao, Wen
2015-01-01
This comprehensive and accessible text/reference presents an overview of the state of the art in video coding technology. Specifically, the book introduces the tools of the AVS2 standard, describing how AVS2 can help to achieve a significant improvement in coding efficiency for future video networks and applications by incorporating smarter coding tools such as scene video coding. Topics and features: introduces the basic concepts in video coding, and presents a short history of video coding technology and standards; reviews the coding framework, main coding tools, and syntax structure of AV
Burnup calculations using serpent code in accelerator driven thorium reactors
Energy Technology Data Exchange (ETDEWEB)
Korkmaz, M.E.; Agar, O. [Karamanoglu Mehmetbey Univ., Karaman (Turkey). Physics Dept.; Yigit, M. [Aksaray Univ. (Turkey). Physics Dept.
2013-07-15
In this study, burnup calculations have been performed for a sodium cooled Accelerator Driven Thorium Reactor (ADTR) using the Serpent 1.1.16 Monte Carlo code. The ADTR has been designed for burning minor actinides, mixed {sup 232}Th and mixed {sup 233}U fuels. A solid Pb-Bi spallation target in the center of the core is used and sodium as coolant. The system is designed for a heating power of 2 000 MW and for an operation time of 600 days. For burnup calculations the Advanced Matrix Exponential Method CRAM (Chebyshev Rational Approximation Method) and different nuclear data libraries (ENDF7, JEF2.2, JEFF3.1.1) were used. The effective multiplication factor change from 0.93 to 0.97 for different nuclear data libraries during the reactor operation period. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Kozmenkov, Y. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Kliem, S. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany)]. E-mail: S.Kliem@fzd.de; Grundmann, U. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Rohde, U. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Weiss, F.-P. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany)
2007-09-15
Plant-measured data provided by the OECD/NEA VVER-1000 coolant transient benchmark programme were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to an experiment that was conducted during the commissioning of the Kozloduy NPP Unit 6 in Bulgaria. In this experiment, the fourth main coolant pump was switched on whilst the remaining three were running normal operating conditions. The experiment was conducted at 27.5% of the nominal level of the reactor power. The transient is characterized by a rapid increase in the primary coolant flow through the core, and as a consequence, a decrease of the space-dependent core inlet temperature. The control rods were kept in their original positions during the entire transient. The coupled simulations performed on both DYN3D/RELAP5 and DYN3D/ATHLET were based on the same reactor model, including identical main coolant pump characteristics, boundary conditions, benchmark-specified nuclear data library and nearly identical nodalization schemes. In addition to validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has also been performed to evaluate the respective thermal hydraulic models of the system codes RELAP5 and ATHLET.
Radiative transfer code: Application to the calculation of PAR
Indian Academy of Sciences (India)
Radiative transfer code: Application to the calculation of PAR. DEVRED EMMANUEL, DUBUISSON .... water vapor from Leckner (1978), with absorption coefficients for these gases taken from Gregg and. Carder (1990). ... law for the size distribution. The particle size ranges from 0.01 "m to 50 "m. 3. Validation of the code.
Research on Primary Shielding Calculation Source Generation Codes
Directory of Open Access Journals (Sweden)
Zheng Zheng
2017-01-01
Full Text Available Primary Shielding Calculation (PSC plays an important role in reactor shielding design and analysis. In order to facilitate PSC, a source generation code is developed to generate cumulative distribution functions (CDF for the source particle sample code of the J Monte Carlo Transport (JMCT code, and a source particle sample code is deveoped to sample source particle directions, types, coordinates, energy and weights from the CDFs. A source generation code is developed to transform three dimensional (3D power distributions in xyz geometry to source distributions in r θ z geometry for the J Discrete Ordinate Transport (JSNT code. Validation on PSC model of Qinshan No.1 nuclear power plant (NPP, CAP1400 and CAP1700 reactors are performed. Numerical results show that the theoretical model and the codes are both correct.
Research on Primary Shielding Calculation Source Generation Codes
Zheng, Zheng; Mei, Qiliang; Li, Hui; Shangguan, Danhua; Zhang, Guangchun
2017-09-01
Primary Shielding Calculation (PSC) plays an important role in reactor shielding design and analysis. In order to facilitate PSC, a source generation code is developed to generate cumulative distribution functions (CDF) for the source particle sample code of the J Monte Carlo Transport (JMCT) code, and a source particle sample code is deveoped to sample source particle directions, types, coordinates, energy and weights from the CDFs. A source generation code is developed to transform three dimensional (3D) power distributions in xyz geometry to source distributions in r θ z geometry for the J Discrete Ordinate Transport (JSNT) code. Validation on PSC model of Qinshan No.1 nuclear power plant (NPP), CAP1400 and CAP1700 reactors are performed. Numerical results show that the theoretical model and the codes are both correct.
LMR core disassembly calculation using the VENUS-II code
Energy Technology Data Exchange (ETDEWEB)
Suk, S. D.; Lee, Y. B. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
2004-07-01
In this study, core disassembly analysis in the sodium-voided core of the KALIMER were performed using the VENUS-II code for various reactivity insertion rates up to 100 $/s, which has been widely considered to be the upper limit of ramp rates due to fuel compaction. The VENUS-II code is a two-dimensional coupled neutronics-hydrodynamics program that calculates the dynamic behavior of an LMFR during a prompt critical excursion. Core disruptive accidents have been investigated at Korea Atomic Energy Research Institute(KAERI) as part of the work to demonstrate the inherent and ultimate safety of conceptual design of the Korea Advanced Liquid Metal Reactor(KALIMER), a pool-type sodium cooled prototype fast reactor that uses U-TRU-Zr metallic fuel. Results of previous scoping studies using a modified Bethe-Tait methods showed that the core disruptive power excursions are terminated without any significant pressure rise or energy release to threaten the integrity of the reactor system.This study is a continual to the previous scoping studies, analyzing core disruptive accidents in more detail using the VENUS-II code.
Energy Technology Data Exchange (ETDEWEB)
Hernandez Valle, S.; Lopez Aldama, D. [Centro de Investigaciones Nucleares, Tecnologicas y Ambientales, La Habana (Cuba). E-mail: svalle@ctn.isctn.edu.cu
2000-07-01
The following paper contributes to validate the Nuclear Engineering Department methods to carry out calculations in TRIGA reactors solving a Benchmark. The benchmark is analyzed with the WIMS-D/4-SNAP/3D code system and using the cross section library WIMS-TRIGA. A brief description of the DSN method is presented used in WIMS/d{sup 4} code and also the SNAP-3d code is shortly explained. The results are presented and compared with the experimental values. In other hand the possible error sources are analyzed. (author)
Energy Technology Data Exchange (ETDEWEB)
NONE
2003-07-01
These scientific days had for objective to draw up the situation of calculation codes of radiation transport, of sources estimation, of radiation doses managements and to draw the future perspectives. (N.C.)
Verification test calculations for the Source Term Code Package
Energy Technology Data Exchange (ETDEWEB)
Denning, R S; Wooton, R O; Alexander, C A; Curtis, L A; Cybulskis, P; Gieseke, J A; Jordan, H; Lee, K W; Nicolosi, S L
1986-07-01
The purpose of this report is to demonstrate the reasonableness of the Source Term Code Package (STCP) results. Hand calculations have been performed spanning a wide variety of phenomena within the context of a single accident sequence, a loss of all ac power with late containment failure, in the Peach Bottom (BWR) plant, and compared with STCP results. The report identifies some of the limitations of the hand calculation effort. The processes involved in a core meltdown accident are complex and coupled. Hand calculations by their nature must deal with gross simplifications of these processes. Their greatest strength is as an indicator that a computer code contains an error, for example that it doesn't satisfy basic conservation laws, rather than in showing the analysis accurately represents reality. Hand calculations are an important element of verification but they do not satisfy the need for code validation. The code validation program for the STCP is a separate effort. In general the hand calculation results show that models used in the STCP codes (e.g., MARCH, TRAP-MELT, VANESA) obey basic conservation laws and produce reasonable results. The degree of agreement and significance of the comparisons differ among the models evaluated. 20 figs., 26 tabs.
Hot zero power reactor calculations using the Insilico code
Hamilton, Steven P.; Evans, Thomas M.; Davidson, Gregory G.; Johnson, Seth R.; Pandya, Tara M.; Godfrey, Andrew T.
2016-06-01
In this paper we describe the reactor physics simulation capabilities of the Insilico code. A description of the various capabilities of the code is provided, including detailed discussion of the geometry, meshing, cross section processing, and neutron transport options. Numerical results demonstrate that Insilico using an SPN solver with pin-homogenized cross section generation is capable of delivering highly accurate full-core simulation of various pressurized water reactor problems. Comparison to both Monte Carlo calculations and measured plant data is provided.
Elements of algebraic coding systems
Cardoso da Rocha, Jr, Valdemar
2014-01-01
Elements of Algebraic Coding Systems is an introductory text to algebraic coding theory. In the first chapter, you'll gain inside knowledge of coding fundamentals, which is essential for a deeper understanding of state-of-the-art coding systems. This book is a quick reference for those who are unfamiliar with this topic, as well as for use with specific applications such as cryptography and communication. Linear error-correcting block codes through elementary principles span eleven chapters of the text. Cyclic codes, some finite field algebra, Goppa codes, algebraic decoding algorithms, and applications in public-key cryptography and secret-key cryptography are discussed, including problems and solutions at the end of each chapter. Three appendices cover the Gilbert bound and some related derivations, a derivation of the Mac- Williams' identities based on the probability of undetected error, and two important tools for algebraic decoding-namely, the finite field Fourier transform and the Euclidean algorithm f...
WIPP Benchmark calculations with the large strain SPECTROM codes
Energy Technology Data Exchange (ETDEWEB)
Callahan, G.D.; DeVries, K.L. [RE/SPEC, Inc., Rapid City, SD (United States)
1995-08-01
This report provides calculational results from the updated Lagrangian structural finite-element programs SPECTROM-32 and SPECTROM-333 for the purpose of qualifying these codes to perform analyses of structural situations in the Waste Isolation Pilot Plant (WIPP). Results are presented for the Second WIPP Benchmark (Benchmark II) Problems and for a simplified heated room problem used in a parallel design calculation study. The Benchmark II problems consist of an isothermal room problem and a heated room problem. The stratigraphy involves 27 distinct geologic layers including ten clay seams of which four are modeled as frictionless sliding interfaces. The analyses of the Benchmark II problems consider a 10-year simulation period. The evaluation of nine structural codes used in the Benchmark II problems shows that inclusion of finite-strain effects is not as significant as observed for the simplified heated room problem, and a variety of finite-strain and small-strain formulations produced similar results. The simplified heated room problem provides stratigraphic complexity equivalent to the Benchmark II problems but neglects sliding along the clay seams. The simplified heated problem does, however, provide a calculational check case where the small strain-formulation produced room closures about 20 percent greater than those obtained using finite-strain formulations. A discussion is given of each of the solved problems, and the computational results are compared with available published results. In general, the results of the two SPECTROM large strain codes compare favorably with results from other codes used to solve the problems.
Hot zero power reactor calculations using the Insilico code
Energy Technology Data Exchange (ETDEWEB)
Hamilton, Steven P., E-mail: hamiltonsp@ornl.gov; Evans, Thomas M., E-mail: evanstm@ornl.gov; Davidson, Gregory G., E-mail: davidsongg@ornl.gov; Johnson, Seth R., E-mail: johnsonsr@ornl.gov; Pandya, Tara M., E-mail: pandyatm@ornl.gov; Godfrey, Andrew T., E-mail: godfreyat@ornl.gov
2016-06-01
In this paper we describe the reactor physics simulation capabilities of the Insilico code. A description of the various capabilities of the code is provided, including detailed discussion of the geometry, meshing, cross section processing, and neutron transport options. Numerical results demonstrate that Insilico using an SP{sub N} solver with pin-homogenized cross section generation is capable of delivering highly accurate full-core simulation of various pressurized water reactor problems. Comparison to both Monte Carlo calculations and measured plant data is provided.
Fuel rod under power oscillations; calculations with the ENIGMA code
Energy Technology Data Exchange (ETDEWEB)
Ranta-Puska, Kari
1999-05-15
Power oscillations in a BWR may result from a series of events starting from a re-circulation pump trip or can be initiated during start-up at low-flow conditions by other perturbations. Whole core and regional oscillations have been observed. Severe consequences may be anticipated if the instability diverges and the reactor protection system fails (no scram) in all phases of the incident (ATWS). Power peaks higher than ten times of the pre-transient power level have been speculated to appear. Low-magnitude oscillations have been observed at the TVO plant, Olkiluoto 1987, and at the Lasalle-2 plant, 1988, and in other BWRs world-wide. Typically, a boiling water reactor has an unstable operational point at low flow and high power conditions. The physical phenomenon behind the instability is density wave oscillations leading to boiling boundary oscillations and void fraction fluctuations across the heated channel. These in turn, make the fission power vary. The typical frequency of the oscillations seems to be of the order of 0.5 Hz, and thus the power peak for a fuel rod is considerably wider than a RIA-pulse, for instance. Large oscillations can result in elevated fuel temperatures, accelerated fission gas release and additional internal loads on the cladding. These effects may be more severe for a high burnup rod with a large fission gas inventory and a closed gap. Therefore, an experiment has been proposed to be conducted at Halden reactor for simulating the fuel rod response under power oscillations. As there is lack of knowledge also on the relevant boundary conditions, pre-calculations with various input options have been performed and are further suggested. Calculations with FRAPTRAN code have shown the importance of the cladding-coolant heat transfer to the fuel temperature. The applicability of the ENIGMA code to this kind of transients was confirmed. To support the planning of the proposed Halden test, estimates on fuel and cladding temperatures as well as
Neutronic calculation of fast reactors by the EUCLID/V1 integrated code
Koltashev, D. A.; Stakhanova, A. A.
2017-01-01
This article considers neutronic calculation of a fast-neutron lead-cooled reactor BREST-OD-300 by the EUCLID/V1 integrated code. The main goal of development and application of integrated codes is a nuclear power plant safety justification. EUCLID/V1 is integrated code designed for coupled neutronics, thermomechanical and thermohydraulic fast reactor calculations under normal and abnormal operating conditions. EUCLID/V1 code is being developed in the Nuclear Safety Institute of the Russian Academy of Sciences. The integrated code has a modular structure and consists of three main modules: thermohydraulic module HYDRA-IBRAE/LM/V1, thermomechanical module BERKUT and neutronic module DN3D. In addition, the integrated code includes databases with fuel, coolant and structural materials properties. Neutronic module DN3D provides full-scale simulation of neutronic processes in fast reactors. Heat sources distribution, control rods movement, reactivity level changes and other processes can be simulated. Neutron transport equation in multigroup diffusion approximation is solved. This paper contains some calculations implemented as a part of EUCLID/V1 code validation. A fast-neutron lead-cooled reactor BREST-OD-300 transient simulation (fuel assembly floating, decompression of passive feedback system channel) and cross-validation with MCU-FR code results are presented in this paper. The calculations demonstrate EUCLID/V1 code application for BREST-OD-300 simulating and safety justification.
Usov, E. V.; Butov, A. A.; Dugarov, G. A.; Kudasov, I. G.; Lezhnin, S. I.; Mosunova, N. A.; Pribaturin, N. A.
2017-07-01
The system of equations from a two-fluid model is widely used in modeling thermohydraulic processes during accidents in nuclear reactors. The model includes conservation equations governing the balance of mass, momentum, and energy in each phase of the coolant. The features of heat and mass transfer, as well as of mechanical interaction between phases or with the channel wall, are described by a system of closing relations. Properly verified foreign and Russian codes with a comprehensive system of closing relations are available to predict processes in water coolant. As to the sodium coolant, only a few open publications on this subject are known. A complete system of closing relations used in the HYDRA-IBRAE/LM/V1 thermohydraulic code for calculation of sodium boiling in channels of power equipment is presented. The selection of these relations is corroborated on the basis of results of analysis of available publications with an account taken of the processes occurring in liquid sodium. A comparison with approaches outlined in foreign publications is presented. Particular attention has been given to the calculation of the sodium two-phase flow boiling. The flow regime map and a procedure for the calculation of interfacial friction and heat transfer in a sodium flow with account taken of high conductivity of sodium are described in sufficient detail. Correlations are presented for calculation of heat transfer for a single-phase sodium flow, sodium flow boiling, and sodium flow boiling crisis. A method is proposed for prediction of flow boiling crisis initiation.
TEMP: a computer code to calculate fuel pin temperatures during a transient. [LMFBR
Energy Technology Data Exchange (ETDEWEB)
Bard, F E; Christensen, B Y; Gneiting, B C
1980-04-01
The computer code TEMP calculates fuel pin temperatures during a transient. It was developed to accommodate temperature calculations in any system of axi-symmetric concentric cylinders. When used to calculate fuel pin temperatures, the code will handle a fuel pin as simple as a solid cylinder or as complex as a central void surrounded by fuel that is broken into three regions by two circumferential cracks. Any fuel situation between these two extremes can be analyzed along with additional cladding, heat sink, coolant or capsule regions surrounding the fuel. The one-region version of the code accurately calculates the solution to two problems having closed-form solutions. The code uses an implicit method, an explicit method and a Crank-Nicolson (implicit-explicit) method.
Energy Technology Data Exchange (ETDEWEB)
Rojas C, E.L.; Varon T, C.F.; Pedraza N, R. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: elrc@nuclear.inin.mx
2007-07-01
The treatment of the breast cancer at early stages is of vital importance. For that, most of the investigations are dedicated to the early detection of the suffering and their treatment. As investigation consequence and clinical practice, in 2002 it was developed in U.S.A. an irradiation system of high dose rate known as Mammosite. In this work we carry out dose calculations for a simplified Mammosite system with the Monte Carlo Penelope simulation code and MCNPX, varying the concentration of the contrast material that it is used in the one. (Author)
KARATE - a code for VVER-440 core calculation
Energy Technology Data Exchange (ETDEWEB)
Gado, J.; Hegedus, Cs.J.; Hegyi, Gy.; Kereszturi, A.; Makai, M.; Maraczi, Cs.; Telbisz, M.
1994-12-31
A modular calculation system has been elaborated at the KFKI Atomic Energy Research Institute for VVER-440 cores. The purpose of KARATE is the calculation of neutron physical and thermal-hydraulic processes in the core at normal, startup, and slow transient conditions. KARATE is under validation and verification (V&V) against mathematical, experimental, and operational data.
Energy Technology Data Exchange (ETDEWEB)
Barr, W.L.; Bathke, C.G.; Brooks, J.N.; Bulmer, R.H.; Busigin, A.; DuBois, P.F.; Fenstermacher, M.E.; Fink, J.; Finn, P.A.; Galambos, J.D.; Gohar, Y.; Gorker, G.E.; Haines, J.R.; Hassanein, A.M.; Hicks, D.R.; Ho, S.K.; Kalsi, S.S.; Kalyanam, K.M.; Kerns, J.A.; Lee, J.D.; Miller, J.R.; Miller, R.L.; Myall, J.O.; Peng, Y-K.M.; Perkins, L.J.; Spampinato, P.T.; Strickler, D.J.; Thomson, S.L.; Wagner, C.E.; Willms, R.S.; Reid, R.L. (ed.)
1988-04-01
A tokamak systems code capable of modeling experimental test reactors has been developed and is described in this document. The code, named TETRA (for Tokamak Engineering Test Reactor Analysis), consists of a series of modules, each describing a tokamak system or component, controlled by an optimizer/driver. This code development was a national effort in that the modules were contributed by members of the fusion community and integrated into a code by the Fusion Engineering Design Center. The code has been checked out on the Cray computers at the National Magnetic Fusion Energy Computing Center and has satisfactorily simulated the Tokamak Ignition/Burn Experimental Reactor II (TIBER) design. A feature of this code is the ability to perform optimization studies through the use of a numerical software package, which iterates prescribed variables to satisfy a set of prescribed equations or constraints. This code will be used to perform sensitivity studies for the proposed International Thermonuclear Experimental Reactor (ITER). 22 figs., 29 tabs.
Donval, Gaël; Moreau, Philippe; Danet, Julien; Larbi, Séverine Jouanneau-Si; Bayle-Guillemaud, Pascale; Boucher, Florent
2017-01-04
Most of the recent developments in EELS modelling has been focused on getting a better agreement with measurements. Less work however has been dedicated to bringing EELS calculations to larger structures that can more realistically describe actual systems. The purpose of this paper is to present a hybrid approach well adapted to calculating the whole set of localised EELS core-loss edges (at the XAS level of theory) on larger systems using only standard tools, namely the WIEN2k and VASP codes. We illustrate the usefulness of this method by applying it to a set of amorphous silicon structures in order to explain the flattening of the silicon L 2,3 EELS edge peak at the onset. We show that the peak flattening is actually caused by the collective contribution of each of the atoms to the average spectrum, as opposed to a flattening occurring on each individual spectrum. This method allowed us to reduce the execution time by a factor of 3 compared to a usual-carefully optimised-WIEN2k calculation. It provided even greater speed-ups on more complex systems (interfaces, ∼300 atoms) that will be presented in a future paper. This method is suited to calculate all the localized edges of all the atoms of a structure in a single calculation for light atoms as long as the core-hole effects can be neglected.
SRAC95; general purpose neutronics code system
Energy Technology Data Exchange (ETDEWEB)
Okumura, Keisuke; Tsuchihashi, Keichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio
1996-03-01
SRAC is a general purpose neutronics code system applicable to core analyses of various types of reactors. Since the publication of JAERI-1302 for the revised SRAC in 1986, a number of additions and modifications have been made for nuclear data libraries and programs. Thus, the new version SRAC95 has been completed. The system consists of six kinds of nuclear data libraries(ENDF/B-IV, -V, -VI, JENDL-2, -3.1, -3.2), five modular codes integrated into SRAC95; collision probability calculation module (PIJ) for 16 types of lattice geometries, Sn transport calculation modules(ANISN, TWOTRAN), diffusion calculation modules(TUD, CITATION) and two optional codes for fuel assembly and core burn-up calculations(newly developed ASMBURN, revised COREBN). In this version, many new functions and data are implemented to support nuclear design studies of advanced reactors, especially for burn-up calculations. SRAC95 is available not only on conventional IBM-compatible computers but also on scalar or vector computers with the UNIX operating system. This report is the SRAC95 users manual which contains general description, contents of revisions, input data requirements, detail information on usage, sample input data and list of available libraries. (author).
Energy Technology Data Exchange (ETDEWEB)
Doyas, R.J.; Dye, R.E.; Howerton, R.J.; Perkins, S.T.
1975-09-30
In the past several years, the Lawrence Livermore Laboratory CLYDE code has been revised and modified extensively. Processing of photon production was incorporated, and the processing of higher-order S/sub n/ transfer matrices was speeded up. A Doppler broadening option was also added. On the other hand, the CLYDE routines that process evaporation models and cumulative probability distributions (I = 5,6) were deleted. The processing of Monte Carlo output was spun off into a separate code, CTART. 3 figures, 11 tables. (auth)
Relative efficiency calculation of a HPGe detector using MCNPX code
Energy Technology Data Exchange (ETDEWEB)
Medeiros, Marcos P.C.; Rebello, Wilson F., E-mail: eng.cavaliere@ime.eb.br, E-mail: rebello@ime.eb.br [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Secao de Engenharia Nuclear; Lopes, Jose M.; Silva, Ademir X., E-mail: marqueslopez@yahoo.com.br, E-mail: ademir@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear
2015-07-01
High-purity germanium detectors (HPGe) are mandatory tools for spectrometry because of their excellent energy resolution. The efficiency of such detectors, quoted in the list of specifications by the manufacturer, frequently refers to the relative full-energy peak efficiency, related to the absolute full-energy peak efficiency of a 7.6 cm x 7.6 cm (diameter x height) NaI(Tl) crystal, based on the 1.33 MeV peak of a {sup 60}Co source positioned 25 cm from the detector. In this study, we used MCNPX code to simulate a HPGe detector (Canberra GC3020), from Real-Time Neutrongraphy Laboratory of UFRJ, to survey the spectrum of a {sup 60}Co source located 25 cm from the detector in order to calculate and confirm the efficiency declared by the manufacturer. Agreement between experimental and simulated data was achieved. The model under development will be used for calculating and comparison purposes with the detector calibration curve from software Genie2000™, also serving as a reference for future studies. (author)
Recent developments in the Los Alamos radiation transport code system
Energy Technology Data Exchange (ETDEWEB)
Forster, R.A.; Parsons, K. [Los Alamos National Lab., NM (United States)
1997-06-01
A brief progress report on updates to the Los Alamos Radiation Transport Code System (LARTCS) for solving criticality and fixed-source problems is provided. LARTCS integrates the Diffusion Accelerated Neutral Transport (DANT) discrete ordinates codes with the Monte Carlo N-Particle (MCNP) code. The LARCTS code is being developed with a graphical user interface for problem setup and analysis. Progress in the DANT system for criticality applications include a two-dimensional module which can be linked to a mesh-generation code and a faster iteration scheme. Updates to MCNP Version 4A allow statistical checks of calculated Monte Carlo results.
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.
LOLA SYSTEM: A code block for nodal PWR simulation. Part. I - Simula-3 Code
Energy Technology Data Exchange (ETDEWEB)
Aragones, J. M.; Ahnert, C.; Gomez Santamaria, J.; Rodriguez Olabarria, I.
1985-07-01
Description of the theory and users manual of the SIMULA-3 code, which is part of the core calculation system by nodal theory in one group, called LOLA SYSTEM. SIMULA-3 is the main module of the system, it uses a modified nodal theory, with interface leakages equivalent to the diffusion theory. (Author) 4 refs.
Computer codes used in the calculation of high-temperature thermodynamic properties of sodium
Energy Technology Data Exchange (ETDEWEB)
Fink, J.K.
1979-12-01
Three computer codes - SODIPROP, NAVAPOR, and NASUPER - were written in order to calculate a self-consistent set of thermodynamic properties for saturated, subcooled, and superheated sodium. These calculations incorporate new critical parameters (temperature, pressure, and density) and recently derived single equations for enthalpy and vapor pressure. The following thermodynamic properties have been calculated in these codes: enthalpy, heat capacity, entropy, vapor pressure, heat of vaporization, density, volumetric thermal expansion coefficient, compressibility, and thermal pressure coefficient. In the code SODIPROP, these properties are calculated for saturated and subcooled liquid sodium. Thermodynamic properties of saturated sodium vapor are calculated in the code NAVAPOR. The code NASUPER calculates thermodynamic properties for super-heated sodium vapor only for low (< 1644 K) temperatures. No calculations were made for the supercritical region.
Improved decoding for a concatenated coding system
DEFF Research Database (Denmark)
Paaske, Erik
1990-01-01
The concatenated coding system recommended by CCSDS (Consultative Committee for Space Data Systems) uses an outer (255,233) Reed-Solomon (RS) code based on 8-b symbols, followed by the block interleaver and an inner rate 1/2 convolutional code with memory 6. Viterbi decoding is assumed. Two new...
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)
A Sample Calculation of Tritium Production and Distribution at VHTR by using TRITGO Code
Energy Technology Data Exchange (ETDEWEB)
Park, Ik Kyu; Kim, D. H.; Lee, W. J
2007-03-15
TRITGO code was developed for estimating the tritium production and distribution of high temperature gas cooled reactor(HTGR), especially GTMHR350 by General Atomics. In this study, the tritium production and distribution of NHDD was analyzed by using TRITGO Code. The TRITGO code was improved by a simple method to calculate the tritium amount in IS Loop. The improved TRITGO input for the sample calculation was prepared based on GTMHR600 because the NHDD has been designed referring GTMHR600. The GTMHR350 input with related to the tritium distribution was directly used. The calculated tritium activity among the hydrogen produced in IS-Loop is 0.56 Bq/g- H2. This is a very satisfying result considering that the limited tritium activity of Japanese Regulation Guide is 5.6 Bq/g-H2. The basic system to analyze the tritium production and the distribution by using TRITGO was successfully constructed. However, there exists some uncertainties in tritium distribution models, the suggested method for IS-Loop, and the current input was not for NHDD but for GTMHR600. The qualitative analysis for the distribution model and the IS-Loop model and the quantitative analysis for the input should be done in the future.
Tandem Mirror Reactor Systems Code (Version I)
Energy Technology Data Exchange (ETDEWEB)
Reid, R.L.; Finn, P.A.; Gohar, M.Y.; Barrett, R.J.; Gorker, G.E.; Spampinaton, P.T.; Bulmer, R.H.; Dorn, D.W.; Perkins, L.J.; Ghose, S.
1985-09-01
A computer code was developed to model a Tandem Mirror Reactor. Ths is the first Tandem Mirror Reactor model to couple, in detail, the highly linked physics, magnetics, and neutronic analysis into a single code. This report describes the code architecture, provides a summary description of the modules comprising the code, and includes an example execution of the Tandem Mirror Reactor Systems Code. Results from this code for two sensitivity studies are also included. These studies are: (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power; and (2) to determine the impact of reactor power level on cost.
Fission cross section calculations of actinides with EMPIRE code
Energy Technology Data Exchange (ETDEWEB)
Sin, M.; Oblozinsky, P.; Herman,M.; Capote,R.
2010-04-30
The cross sections of the neutron induced reactions on {sup 233,234,236}U, {sup 237}Np, {sup 238,242}Pu, {sup 241,243}Am, {sup 242,246}Cm carried out in the energy range 1 keV-20 MeV with EMPIRE code are presented, emphasizing the fission channel. Beside a consistent, accurate set of evaluations, the paper contains arguments supporting the choice of the reaction models and input parameters. A special attention is paid to the fission parameters and their uncertainties.
Energy Technology Data Exchange (ETDEWEB)
Rearden, Bradley T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jessee, Matthew Anderson [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
2016-08-01
The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.
Energy Technology Data Exchange (ETDEWEB)
Rearden, Bradley T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jessee, Matthew Anderson [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
2017-05-01
The SCALE Code System is a widely used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including 3 deterministic and 3 Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results. SCALE 6.2 represents one of the most comprehensive revisions in the history of SCALE, providing several new capabilities and significant improvements in many existing features.
SINFAC - SYSTEMS IMPROVED NUMERICAL FLUIDS ANALYSIS CODE
Costello, F. A.
1994-01-01
. On the first pass, the user finds that the calculated outlet conditions of the last component do not match the estimated inlet conditions of the first. The user then modifies the estimated inlet conditions of the first component in an attempt to match the calculated values. The user estimated values are called State Variables. The differences between the user estimated values and calculated values are called the Error Variables. The procedure systematically changes the State Variables until all of the Error Variables are less than the user-specified iteration limits. The solution procedure is referred to as SCX. It consists of two phases, the Systems phase and the Controller phase. The X is to imply experimental. SCX computes each next set of State Variables in two phases. In the first phase, SCX fixes the controller positions and modifies the other State Variables by the Newton-Raphson method. This first phase is the Systems phase. Once the Newton-Raphson method has solved the problem for the fixed controller positions, SCX next calculates new controller positions based on Newton's method while treating each sensor-controller pair independently but allowing all to change in one iteration. This phase is the Controller phase. SINFAC is available by license for a period of ten (10) years to approved licensees. The licenced program product includes the source code for the additional routines to SINDA, the SINDA object code, command procedures, sample data and supporting documentation. Additional documentation may be purchased at the price below. SINFAC was created for use on a DEC VAX under VMS. Source code is written in FORTRAN 77, requires 180k of memory, and should be fully transportable. The program was developed in 1988.
Energy Technology Data Exchange (ETDEWEB)
Liu, J
2004-07-12
Significantly improved upon its predecessor PHOTON, STAC8 is a valuable analytic code for quick and conservative beamline shielding designs for synchrotron radiation (SR) facilities. To check the applicability, accuracy, and limitations of STAC8, studies were conducted to compare STAC8 and PHOTON results with calculations using the FLUKA and EGS4 Monte Carlo codes. Doses and spectra for scattered SR in a few beam-target-shield geometries were calculated, with and without photon linear polarization effects. Areas for expanding the STAC8 capabilities, e.g., features of the mirror-reflected lights and double-Compton light calculations, and use of monochromatic light, etc., have been identified. Some of these features have been implemented and benchmarked against Monte Carlo calculations. Reasonable agreements were found between the STAC8 and Monte Carlo calculations.
Energy Technology Data Exchange (ETDEWEB)
Holly R. Trellue
1998-12-01
Monteburns is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code 0RIGEN2. Monteburns produces many criticality and burnup computational parameters based on material feed/removal specifications, power(s), and time intervals. This code processes input from the user indicating the system geometry, initial material compositions, feed/removal, and other code-specific parameters. Results from MCNP, 0RIGEN2, and other calculations are then output successively as the code runs. The principle function of monteburns is to first transfer one-group cross sections and fluxes from MCNP to 0RIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from 0RIGEN2 back to MCNP in a repeated, cyclic fashion. The main requirement of the code is that the user have a working MCNP input file and other input parameters; all interaction with 0RIGEN2 and other calculations are performed by monteburns. This report presents the results obtained from the benchmarking of monteburns to measured and previously obtained data from traditional Light Water Reactor systems. The majority of the differences seen between the two were less than five percent. These were primarily a result of variances in cross sections between MCNP, cross section libraries used by other codes, and observed values. With this understanding, this code can now be used with confidence for burnup calculations in three-dimensional systems. It was designed for use in the Accelerator Transmutation of Waste project at Los Alamos National Laboratory but is also being applied to the analysis of isotopic production/destruction of transuranic actinides in a reactor system. The code has now been shown to sufficiently support these calculations.
Numerical identification of bacteria with a hand-held calculator as an alternative to code books.
Schindler, J.; Schindler, Z
1982-01-01
The Hewlett-Packard HP 41C hand-held calculator can be used for the numerical identification of bacteria. The dimensions of the identification matrix are limited to about 30 by 22; however, many groups of clinically important bacteria can be numerically identified by this method. Hand-held calculators can be used as an alternative to code books. At present, these calculators and additional tests can help solve identification problems in profiles not contained in code books.
Radiative transfer code: Application to the calculation of PAR
Indian Academy of Sciences (India)
Home; Journals; Journal of Earth System Science; Volume 109; Issue 4 ... The production of carbon in the ocean, the so-called primary production, depends on various physico- biological parameters: the biomass and nutrient amounts in oceans, the salinity and temperature of the water and the light available in the water ...
Energy Technology Data Exchange (ETDEWEB)
Hirayama, Hideo; Namito, Yoshihito; /KEK, Tsukuba; Bielajew, Alex F.; Wilderman, Scott J.; U., Michigan; Nelson, Walter R.; /SLAC
2005-12-20
In the nineteen years since EGS4 was released, it has been used in a wide variety of applications, particularly in medical physics, radiation measurement studies, and industrial development. Every new user and every new application bring new challenges for Monte Carlo code designers, and code refinements and bug fixes eventually result in a code that becomes difficult to maintain. Several of the code modifications represented significant advances in electron and photon transport physics, and required a more substantial invocation than code patching. Moreover, the arcane MORTRAN3[48] computer language of EGS4, was highest on the complaint list of the users of EGS4. The size of the EGS4 user base is difficult to measure, as there never existed a formal user registration process. However, some idea of the numbers may be gleaned from the number of EGS4 manuals that were produced and distributed at SLAC: almost three thousand. Consequently, the EGS5 project was undertaken. It was decided to employ the FORTRAN 77 compiler, yet include as much as possible, the structural beauty and power of MORTRAN3. This report consists of four chapters and several appendices. Chapter 1 is an introduction to EGS5 and to this report in general. We suggest that you read it. Chapter 2 is a major update of similar chapters in the old EGS4 report[126] (SLAC-265) and the old EGS3 report[61] (SLAC-210), in which all the details of the old physics (i.e., models which were carried over from EGS4) and the new physics are gathered together. The descriptions of the new physics are extensive, and not for the faint of heart. Detailed knowledge of the contents of Chapter 2 is not essential in order to use EGS, but sophisticated users should be aware of its contents. In particular, details of the restrictions on the range of applicability of EGS are dispersed throughout the chapter. First-time users of EGS should skip Chapter 2 and come back to it later if necessary. With the release of the EGS4 version
LOLA SYSTEM: A code block for nodal PWR simulation. Part. II - MELON-3, CONCON and CONAXI Codes
Energy Technology Data Exchange (ETDEWEB)
Aragones, J. M.; Ahnert, C.; Gomez Santamaria, J.; Rodriguez Olabarria, I.
1985-07-01
Description of the theory and users manual of the MELON-3, CONCON and CONAXI codes, which are part of the core calculation system by nodal theory in one group, called LOLA SYSTEM. These auxiliary codes, provide some of the input data for the main module SIMULA-3; these are, the reactivity correlations constants, the albe does and the transport factors. (Author) 7 refs.
Study on GEANT4 code applications to dose calculation using imaging data
Lee, Jeong Ok; Kang, Jeong Ku; Kim, Jhin Kee; Kwon, Hyeong Cheol; Kim, Jung Soo; Kim, Bu Gil; Jeong, Dong Hyeok
2015-07-01
The use of the GEANT4 code has increased in the medical field. Various studies have calculated the patient dose distributions by users the GEANT4 code with imaging data. In present study, Monte Carlo simulations based on DICOM data were performed to calculate the dose absorb in the patient's body. Various visualization tools are installed in the GEANT4 code to display the detector construction; however, the display of DICOM images is limited. In addition, to displaying the dose distributions on the imaging data of the patient is difficult. Recently, the gMocren code, a volume visualization tool for GEANT4 simulation, was developed and has been used in volume visualization of image files. In this study, the imaging based on the dose distributions absorbed in the patients was performed by using the gMocren code. Dosimetric evaluations with were carried out by using thermo luminescent dosimeter and film dosimetry to verify the calculated results.
ECCD calculations in ITER by means of the quasi-optical code
Bertelli, N.; Balakin, A. A.; Westerhof, E.; Buyanova, M. N.
2010-01-01
Electron cyclotron current drive (ECCD) calculations for the case of the ITER electron cyclotron resonant heating upper port launcher are presented making use of a quasi-optical (QO) code (Balakin et al 2008 Nucl. Fusion 48 065003). The QO code describes accurately the behaviour of the wave beam in
Energy Technology Data Exchange (ETDEWEB)
Strenge, D.L.; Peloquin, R.A.
1981-04-01
The computer code HADOC (Hanford Acute Dose Calculations) is described and instructions for its use are presented. The code calculates external dose from air submersion and inhalation doses following acute radionuclide releases. Atmospheric dispersion is calculated using the Hanford model with options to determine maximum conditions. Building wake effects and terrain variation may also be considered. Doses are calculated using dose conversion factor supplied in a data library. Doses are reported for one and fifty year dose commitment periods for the maximum individual and the regional population (within 50 miles). The fractional contribution to dose by radionuclide and exposure mode are also printed if requested.
Energy Technology Data Exchange (ETDEWEB)
Koo, Bon Seung; Lee, Kyung Hoon; Song, Jae Seung; Park, Sang Yoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2013-10-15
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.
Energy Technology Data Exchange (ETDEWEB)
Giuseppe Palmiotti
2015-05-01
In this work, the implementation of a collision history-based approach to sensitivity/perturbation calculations in the Monte Carlo code SERPENT is discussed. The proposed methods allow the calculation of the eects of nuclear data perturbation on several response functions: the eective multiplication factor, reaction rate ratios and bilinear ratios (e.g., eective kinetics parameters). SERPENT results are compared to ERANOS and TSUNAMI Generalized Perturbation Theory calculations for two fast metallic systems and for a PWR pin-cell benchmark. New methods for the calculation of sensitivities to angular scattering distributions are also presented, which adopts fully continuous (in energy and angle) Monte Carlo estimators.
Progress on China nuclear data processing code system
Liu, Ping; Wu, Xiaofei; Ge, Zhigang; Li, Songyang; Wu, Haicheng; Wen, Lili; Wang, Wenming; Zhang, Huanyu
2017-09-01
China is developing the nuclear data processing code Ruler, which can be used for producing multi-group cross sections and related quantities from evaluated nuclear data in the ENDF format [1]. The Ruler includes modules for reconstructing cross sections in all energy range, generating Doppler-broadened cross sections for given temperature, producing effective self-shielded cross sections in unresolved energy range, calculating scattering cross sections in thermal energy range, generating group cross sections and matrices, preparing WIMS-D format data files for the reactor physics code WIMS-D [2]. Programming language of the Ruler is Fortran-90. The Ruler is tested for 32-bit computers with Windows-XP and Linux operating systems. The verification of Ruler has been performed by comparison with calculation results obtained by the NJOY99 [3] processing code. The validation of Ruler has been performed by using WIMSD5B code.
Comparison of Calculated value by the Core Design ASTRA Code with Measured Data
Energy Technology Data Exchange (ETDEWEB)
Jung, Ji Eun; Yang, Sung Tae [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)
2012-05-15
ASTRA is a unique proprietary core design code being developed by KEPCO NF. This code is still under development, but it can nonetheless be analyzed and given feedback by a utility. This study will be used to develop a risk assessment procedure based on the ASTRA code. The calculated values of the ASTRA code and the measurement values are compared here using the data for Cycle 12 of Younggwang Nuclear Unit 3. The reactor core is composed of 177 fuel assemblies, consisting of a 16x16 array with 236 fuel rods and 5 guide tubes
Decay Power Calculation for Safety Analysis of Innovative Reactor Systems
Energy Technology Data Exchange (ETDEWEB)
Shwageraus, E.; Fridman, E. [Department of Nuclear Engineering, Ben-Gurion University of the Negev Beer-Sheva 84105 (Israel)
2008-07-01
In this work, we verified the decay heat calculation capabilities of BGCore computer code system developed recently at Ben-Gurion University. Decay power was calculated for a typical UO{sub 2} fuel in Pressurized Water Reactor environment using BGCore code and using procedure prescribed by the ANS/ANSI-2005 standard. Very good agreement between the two methods was obtained. Once BGCore calculation capabilities were verified, we calculated decay power as a function of time after shutdown for various reactors with innovative fuels, for which no standard procedure is currently available. Notable differences were observed for decay power of the advanced reactors as compared with conventional UO{sub 2} LWR. The observed differences suggest that the design of new reactors safety systems must be based on corresponding decay power curves for each individual case in order to assure the desired performance of such systems. (authors)
Energy Technology Data Exchange (ETDEWEB)
Cullen, D.E.
1994-08-01
The computer code WALKMAN performs electron single collision elastic scattering Monte Carlo calculations in spherical or planar geometry. It is intended as a research tool to obtain results that can be compared to the results of condensed history calculations. This code is designed to be self documenting, in the sense that the latest documentation is included as comment lines at the beginning of the code. Printed documentation, such as this document, is periodically published and consists mostly of a copy of the comment lines from the code. The user should be aware that the comment lines within the code are continually updated to reflect the most recent status of the code and these comments should always be considered to be the most recent documentation for the code and may supersede published documentation, such as this document. Therefore, the user is advised to always read the documentation within the actual code. The remainder of this report consists of example results and a listing of the documentation which appears at the beginning of the code.
Directory of Open Access Journals (Sweden)
Diego Ferraro
2011-01-01
Full Text Available Monte Carlo neutron transport codes are usually used to perform criticality calculations and to solve shielding problems due to their capability to model complex systems without major approximations. However, these codes demand high computational resources. The improvement in computer capabilities leads to several new applications of Monte Carlo neutron transport codes. An interesting one is to use this method to perform cell-level fuel assembly calculations in order to obtain few group constants to be used on core calculations. In the present work the VTT recently developed Serpent v.1.1.7 cell-oriented neutronic calculation code is used to perform cell calculations of a theoretical BWR lattice benchmark with burnable poisons, and the main results are compared to reported ones and with calculations performed with Condor v.2.61, the INVAP's neutronic collision probability cell code.
Verification of the Parallel C Code with Fortran Code for FDTD Calculations
2010-07-06
package for FDTD calculations of human exposure due to the electromagnetic fields ( EMF ) which was developed by Radio Frequency Radiation Branch, AFRL...SAR) of electromagnetic fields ( EMF ) 5a. CONTRACT NUMBER FA8655-09-1-3045 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 6. AUTHOR(S...of non-ionizing radiation Pohorskega bataljona 215 Ljubljana 1000 Slovenia 8. PERFORMING ORGANIZATION REPORT NUMBER N/A 9. SPONSORING
Low Spectral Efficiency Trellis Coded Modulation Systems
2006-09-01
2 bBW R= . The three alternative systems are all non- TCM systems and consist of QPSK with independent r=1/2 error correction coding on the in-phase...and quadrature components, with null-to-null bandwidth 2 bBW R= , 8-ary biorthogonal keying (8-BOK) with r=2/3 error correction coding with bandwidth...21 12 bBW R= and 16-BOK with r=3/4 error correction coding and with bandwidth 44 24 bBW R= . At the beginning of the analysis only the effect of
Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes
Energy Technology Data Exchange (ETDEWEB)
Baratta, A.J.
1997-07-01
To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts and engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.
Energy Technology Data Exchange (ETDEWEB)
Bordy, J.M.; Kodeli, I.; Menard, St.; Bouchet, J.L.; Renard, F.; Martin, E.; Blazy, L.; Voros, S.; Bochud, F.; Laedermann, J.P.; Beaugelin, K.; Makovicka, L.; Quiot, A.; Vermeersch, F.; Roche, H.; Perrin, M.C.; Laye, F.; Bardies, M.; Struelens, L.; Vanhavere, F.; Gschwind, R.; Fernandez, F.; Quesne, B.; Fritsch, P.; Lamart, St.; Crovisier, Ph.; Leservot, A.; Antoni, R.; Huet, Ch.; Thiam, Ch.; Donadille, L.; Monfort, M.; Diop, Ch.; Ricard, M
2006-07-01
The purpose of this conference was to describe the present state of computer codes dedicated to radiation transport or radiation source assessment or dosimetry. The presentations have been parted into 2 sessions: 1) methodology and 2) uses in industrial or medical or research domains. It appears that 2 different calculation strategies are prevailing, both are based on preliminary Monte-Carlo calculations with data storage. First, quick simulations made from a database of particle histories built though a previous Monte-Carlo simulation and secondly, a neuronal approach involving a learning platform generated through a previous Monte-Carlo simulation. This document gathers the slides of the presentations.
New beam-tracking simulation code using bulk-to-point calculation technique for space charge fields
Mizuno, A.
2016-02-01
A new two-dimensional beam-tracking simulation code for electron injectors using a bulk-to-point calculation technique for space charge fields was developed. The calculated space charge fields are produced not by a point charge but by a hollow cylinder that has a volume. Each tracked electron is a point charge. This bulk-to-point calculation technique for space charge fields is based on that used in the multiple beam envelope equations, which were developed by the author. The multiple beam envelope equations are a set of differential equations for investigating the beam dynamics of electron injectors and can be used to calculate bunched beam dynamics with high accuracy. However, there is one limitation. The bunched beam is assumed to be an ensemble of several segmentation pieces in both the transverse and longitudinal directions. In this bunch model, each longitudinal segmentation slice in a bunch must not warp; consequently, the accuracy of the calculated emittance is reduced in the case of a highly charged beam for calculations of a typical rf gun injector system. This limitation is related to the calculation model of longitudinal space charge fields. In the newly developed beam-tracking simulation code, the space charge field calculation scheme is upgraded and the limitation has been overcome. Therefore, the applicable range is extended while maintaining the high accuracy of emittance calculations. Simultaneously, the calculation time is markedly shortened because the emittance dependence on the segmentation number is extremely weak. In this paper, several examples of beam dynamics that cannot be calculated accurately using the multiple beam envelope equations are demonstrated using the new beam-tracking simulation code. The accuracy of the calculated emittance is also discussed.
Degenerate coding in neural systems.
Leonardo, Anthony
2005-11-01
When the dimensionality of a neural circuit is substantially larger than the dimensionality of the variable it encodes, many different degenerate network states can produce the same output. In this review I will discuss three different neural systems that are linked by this theme. The pyloric network of the lobster, the song control system of the zebra finch, and the odor encoding system of the locust, while different in design, all contain degeneracies between their internal parameters and the outputs they encode. Indeed, although the dynamics of song generation and odor identification are quite different, computationally, odor recognition can be thought of as running the song generation circuitry backwards. In both of these systems, degeneracy plays a vital role in mapping a sparse neural representation devoid of correlations onto external stimuli (odors or song structure) that are strongly correlated. I argue that degeneracy between input and output states is an inherent feature of many neural systems, which can be exploited as a fault-tolerant method of reliably learning, generating, and discriminating closely related patterns.
A Photovoltaic System Payback Calculator
Energy Technology Data Exchange (ETDEWEB)
Riley, Daniel M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Fleming, Jeffrey E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gallegos, Gerald R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
2016-06-01
The Roof Asset Management Program (RAMP) is a DOE NNSA initiative to manage roof repairs and replacement at NNSA facilities. In some cases, installation of a photovoltaic system on new roofs may be possible and desired for financial reasons and to meet federal renewable energy goals. One method to quantify the financial benefits of PV systems is the payback period, or the length of time required for a PV system to generate energy value equivalent to the system's cost. Sandia Laboratories created a simple spreadsheet-based solar energy valuation tool for use by RAMP personnel to quickly evaluate the estimated payback period of prospective or installed photovoltaic systems.
PSOLV: a code for calculating the potentials and densities in MFTF-B
Energy Technology Data Exchange (ETDEWEB)
Colborn, J.A.
1983-08-17
Code PSOLV solves for potential and densities at the cardinal points of MFTF-B. The code is equipped to handle both the throttle-coil and the axicell geometries. For the throttle-coil case, the potential at point MXO is input, while the potentials and densities at points MAI, b, and A are calculated. For the axicell case, the code must additionally solve for the potentials and densities at points X and MXO. PSOLV is intended primarily for use as a subroutine in TREQ, a code being developed by Rensink that calculates the densities and potentials at the cardinal points of MFTF-B as a function of time. TREQ is to be used for modeling start-up behavior.
Systems Improved Numerical Fluids Analysis Code
Costello, F. A.
1990-01-01
Systems Improved Numerical Fluids Analysis Code, SINFAC, consists of additional routines added to April, 1983, version of SINDA. Additional routines provide for mathematical modeling of active heat-transfer loops. Simulates steady-state and pseudo-transient operations of 16 different components of heat-transfer loops, including radiators, evaporators, condensers, mechanical pumps, reservoirs, and many types of valves and fittings. Program contains property-analysis routine used to compute thermodynamic properties of 20 different refrigerants. Source code written in FORTRAN 77.
NEPHTIS: Core depletion validation relying on 2D transport core calculations with the APOLLO2 code
Energy Technology Data Exchange (ETDEWEB)
Damian, F.; Raepsaet, X.; Groizard, M.; Poinot, C. [DEN/DM2S/SERMA/LCA, CEA Saclay, 91191 Gif-sur-Yvette Cedex (France)
2006-07-01
The CEA, in collaboration with EDF and AREVA-NP, is developing a core modelling tool called NEPHTIS, for Neutronic Process for HTGR Innovating Systems and dedicated at present day to the prismatic block-type HTGR (High Temperature Gas-Cooled Reactors). Due to the lack of usable HTGR experimental results, the confidence in this neutronic computational tool relies essentially on comparisons to reference or best-estimate calculations. In the present analysis, the Aleppo deterministic transport code has been selected as reference for validating core depletion simulations carried out within NEPHTIS. These reference calculations were performed on fully detailed 2D core configurations using the Method of Characteristics. The latter has been validated versus Monte Carlo method for different static core configurations [1], [2] and [3]. All the presented results come from an annular HTGR core loaded with uranium-based fuel (15% enrichment). During the core depletion validation, reactivity, reaction rates distributions and nuclei concentrations have been compared. In addition, the impact of various physical and geometrical parameters such as the core loading (one-through or batch-wise reloading) and the amount of burnable poison has been investigated during the validation phases. The results confirm that NEPHTIS is able to predict the core reactivity with uncertainties of {+-}350 pcm. At the end of the core irradiation, the U-235 consumption is calculated within {+-} 0, 7 % while the plutonium mass discharged from the core is calculated within {+-}1 %. As far as the core power distributions are concerned, small discrepancies ( and < 2.3 %) can be observed on the fuel block-averaged power distribution in the core. (authors)
ITP.FOR: A code to calculate thermal transients in High Level Waste Tanks
Energy Technology Data Exchange (ETDEWEB)
Kielpinski, A.L.
1992-10-01
A variety of processing operations for high level radioactive waste occur in the High Level Waste Tanks in the H-Area of the Savannah River Site. Thermal design constraints exist on these processes, principally to limit the amount of corrosion inhibitor which must be added to protect the tank and cooling coil materials. The required amount of corrosion inhibitor, which must subsequently be removed prior to trapping the waste in borosilicate glass, increases exponentially with temperature over a fairly narrow range (some tens of degrees Celsius). For this reason, there is a need to model the thermal-hydraulic processes occurring in the waste tanks. A FORTRAN computer code, called ITP.FOR, was written to provide a simple but reasonably accurate analysis tool for plant operation design. The code was specifically written to model Tank 48, in which the In-Tank Precipitation (ITP) process of precipitating radioactive cesium will be initiated. Although the ITP.FOR code was written as personal-use software for scoping design calculations for Tank 48, the current intent is to extend the code`s applicability to other H-Area waste tanks, and to certify the code in accordance with the NRTSC Quality Assurance requirements for critical-use software (1Q-34, 1991). Since the code`s capabilities have generated some interest to date, the present report is presented as interim documentation of the code`s mathematical models. This documentation will eventually be supplanted by the formal documentation of the expanded and benchmarked code.
A toolchain for the automatic generation of computer codes for correlated wavefunction calculations.
Krupička, Martin; Sivalingam, Kantharuban; Huntington, Lee; Auer, Alexander A; Neese, Frank
2017-06-05
In this work, the automated generator environment for ORCA (ORCA-AGE) is described. It is a powerful toolchain for the automatic implementation of wavefunction-based quantum chemical methods. ORCA-AGE consists of three main modules: (1) generation of "raw" equations from a second quantized Ansatz for the wavefunction, (2) factorization and optimization of equations, and (3) generation of actual computer code. We generate code for the ORCA package, making use of the powerful functionality for wavefunction-based correlation calculations that is already present in the code. The equation generation makes use of the most elementary commutation relations and hence is extremely general. Consequently, code can be generated for single reference as well as multireference approaches and spin-independent as well as spin-dependent operators. The performance of the generated code is demonstrated through comparison with efficient hand-optimized code for some well-understood standard configuration interaction and coupled cluster methods. In general, the speed of the generated code is no more than 30% slower than the hand-optimized code, thus allowing for routine application of canonical ab initio methods to molecules with about 500-1000 basis functions. Using the toolchain, complicated methods, especially those surpassing human ability for handling complexity, can be efficiently and reliably implemented in very short times. This enables the developer to shift the attention from debugging code to the physical content of the chosen wavefunction Ansatz. Automatic code generation also has the desirable property that any improvement in the toolchain immediately applies to all generated code. © 2017 Wiley Periodicals, Inc. © 2017 Wiley Periodicals, Inc.
Thermodynamic Calculations for Systems Biocatalysis
DEFF Research Database (Denmark)
Abu, Rohana; Gundersen, Maria T.; Woodley, John M.
2015-01-01
‘Systems Biocatalysis’ is a term describing multi-enzyme processes in vitro for the synthesis of chemical products. Unlike in-vivo systems, such an artificial metabolism can be controlled in a highly efficient way in order to achieve a sufficiently favourable conversion for a given target product...... the transamination of a pro-chiral ketone into a chiral amine (interesting in many pharmaceutical applications). Here, the products are often less energetically stable than the reactants, meaning that the reaction may be thermodynamically unfavourable. As in nature, such thermodynamically-challenged reactions can...... energy change, View the MathML source ΔGro′, of the overall cascade. The findings show that unfavourable reactions in the cascade can be improved by coupling to a favourable reaction giving more energetically stable products....
Performance calculation of the Compact Disc error correcting code ona memoryless channel
Driessen, L.H.M.E.; Vries, L.B.
2015-01-01
Recently N.V. PHILIPS of The Netherlands and SONY CORP. of Japan made a joint pnoposal for standardization of their COMPACT DISC DigitalAudio system. This standard as agreed upon, includes the choice ofan error correcting code called CIRC (Cross Interleave Reed SolomonCode),according to which the
PEGASUS: a preequilibrium and multi-step evaporation code for neutron cross section calculation
Energy Technology Data Exchange (ETDEWEB)
Nakagawa, Tsuneo; Sugi, Teruo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iijima, Shungo; Nishigori, Takeo
1999-06-01
The computer code PEGASUS was developed to calculate neutron-induced reaction cross sections on the basis of the closed form exciton model preequilibrium theory and the multi-step evaporation theory. The cross sections and emitted particle spectra are calculated for the compound elastic scattering, (n,{gamma}), (n,n`), (n,p), (n,{alpha}), (n,d), (n,t), (n,{sup 3}He), (n,2n), (n,n`p), (n,n`{alpha}), (n,n`d), (n,n`t), (n,2p) and (n,3n) reactions. The double differential cross sections of emitted particles are also calculated. The calculated results are written on a magnetic disk in the ENDF format. Parameter files and/or systematics formulas are provided for level densities, mass excess, radiation widths and inverse cross sections so that the input data to the code are made minimum. (author)
SYN3D: a single-channel, spatial flux synthesis code for diffusion theory calculations
Energy Technology Data Exchange (ETDEWEB)
Adams, C. H.
1976-07-01
This report is a user's manual for SYN3D, a computer code which uses single-channel, spatial flux synthesis to calculate approximate solutions to two- and three-dimensional, finite-difference, multigroup neutron diffusion theory equations. SYN3D is designed to run in conjunction with any one of several one- and two-dimensional, finite-difference codes (required to generate the synthesis expansion functions) currently being used in the fast reactor community. The report describes the theory and equations, the use of the code, and the implementation on the IBM 370/195 and CDC 7600 of the version of SYN3D available through the Argonne Code Center.
Energy Technology Data Exchange (ETDEWEB)
Urata, Kazuhiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment
2003-03-01
In design of the future fusion devises in which low activation ferritic steel is planned to use as the plasma facing material and/or the inserts for ripple reduction, the appreciation of the error field effect against the plasma as well as the optimization of ferritic plate arrangement to reduce the toroidal field ripple require calculation of magnetic field generated by ferritic steel. However iterative calculations concerning the non-linearity in B-H curve of ferritic steel disturbs high-speed calculation required as the design tool. In the strong toroidal magnetic field that is characteristic in the tokamak fusion devices, fully magnetic saturation of ferritic steel occurs. Hence a distribution of magnetic charges as magnetic field source is determined straightforward and any iteration calculation are unnecessary. Additionally objective ferritic steel geometry is limited to the thin plate and ferritic plates are installed along the toroidal magnetic field. Taking these special conditions into account, high-speed calculation code ''FEMAG'' has been developed. In this report, the formalization of 'FEMAG' code, how to use 'FEMAG', and the validity check of 'FEMAG' in comparison with a 3D FEM code, with the measurements of the magnetic field in JFT-2M are described. The presented examples are numerical results of design studies for JT-60 modification. (author)
TrackEtching - A Java based code for etched track profile calculations in SSNTDs
Muraleedhara Varier, K.; Sankar, V.; Gangadathan, M. P.
2017-09-01
A java code incorporating a user friendly GUI has been developed to calculate the parameters of chemically etched track profiles of ion-irradiated solid state nuclear track detectors. Huygen's construction of wavefronts based on secondary wavelets has been used to numerically calculate the etched track profile as a function of the etching time. Provision for normal incidence and oblique incidence on the detector surface has been incorporated. Results in typical cases are presented and compared with experimental data. Different expressions for the variation of track etch rate as a function of the ion energy have been utilized. The best set of values of the parameters in the expressions can be obtained by comparing with available experimental data. Critical angle for track development can also be calculated using the present code.
Linear calculations of edge current driven kink modes with BOUT++ code
Energy Technology Data Exchange (ETDEWEB)
Li, G. Q., E-mail: ligq@ipp.ac.cn; Xia, T. Y. [Institute of Plasma Physics, CAS, Hefei, Anhui 230031 (China); Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Xu, X. Q. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Snyder, P. B.; Turnbull, A. D. [General Atomics, San Diego, California 92186 (United States); Ma, C. H.; Xi, P. W. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); FSC, School of Physics, Peking University, Beijing 100871 (China)
2014-10-15
This work extends previous BOUT++ work to systematically study the impact of edge current density on edge localized modes, and to benchmark with the GATO and ELITE codes. Using the CORSICA code, a set of equilibria was generated with different edge current densities by keeping total current and pressure profile fixed. Based on these equilibria, the effects of the edge current density on the MHD instabilities were studied with the 3-field BOUT++ code. For the linear calculations, with increasing edge current density, the dominant modes are changed from intermediate-n and high-n ballooning modes to low-n kink modes, and the linear growth rate becomes smaller. The edge current provides stabilizing effects on ballooning modes due to the increase of local shear at the outer mid-plane with the edge current. For edge kink modes, however, the edge current does not always provide a destabilizing effect; with increasing edge current, the linear growth rate first increases, and then decreases. In benchmark calculations for BOUT++ against the linear results with the GATO and ELITE codes, the vacuum model has important effects on the edge kink mode calculations. By setting a realistic density profile and Spitzer resistivity profile in the vacuum region, the resistivity was found to have a destabilizing effect on both the kink mode and on the ballooning mode. With diamagnetic effects included, the intermediate-n and high-n ballooning modes can be totally stabilized for finite edge current density.
Linear calculations of edge current driven kink modes with BOUT++ code
Li, G. Q.; Xu, X. Q.; Snyder, P. B.; Turnbull, A. D.; Xia, T. Y.; Ma, C. H.; Xi, P. W.
2014-10-01
This work extends previous BOUT++ work to systematically study the impact of edge current density on edge localized modes, and to benchmark with the GATO and ELITE codes. Using the CORSICA code, a set of equilibria was generated with different edge current densities by keeping total current and pressure profile fixed. Based on these equilibria, the effects of the edge current density on the MHD instabilities were studied with the 3-field BOUT++ code. For the linear calculations, with increasing edge current density, the dominant modes are changed from intermediate-n and high-n ballooning modes to low-n kink modes, and the linear growth rate becomes smaller. The edge current provides stabilizing effects on ballooning modes due to the increase of local shear at the outer mid-plane with the edge current. For edge kink modes, however, the edge current does not always provide a destabilizing effect; with increasing edge current, the linear growth rate first increases, and then decreases. In benchmark calculations for BOUT++ against the linear results with the GATO and ELITE codes, the vacuum model has important effects on the edge kink mode calculations. By setting a realistic density profile and Spitzer resistivity profile in the vacuum region, the resistivity was found to have a destabilizing effect on both the kink mode and on the ballooning mode. With diamagnetic effects included, the intermediate-n and high-n ballooning modes can be totally stabilized for finite edge current density.
Calculation of an accident with delayed scram at NPP Greifswald using the coupled code DYN3D/ATHLET
Energy Technology Data Exchange (ETDEWEB)
Kliem, S.
1998-10-01
Complex computer codes modeling the whole reactor system including 3D neutron kinetics in combination with advanced thermohydraulic plant models become more and more important for the safety assessment of nuclear reactors. Transients or experiments with both neutron kinetic and thermalhydraulic data are needed for the validation of such coupled codes like DYN3D/ATHLET. First of all measured results from nuclear power plant (NPP) transients should be used, because the experimental thermalhydraulic facilities do not offer the possibility to model space-dependent neutron kinetic effects and research reactors with reliably measured 3D neutron kinetic data do not allow to study thermalhydraulic feedback effects. In this paper, an accident with delayed scram which occurred in 1989 at the NPP Greifswald is analyzed. Calculations of this accident were carried out with the goal to validate the coupled code DYN3D/ATHLET. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Eslinger, Paul W.; Aaberg, Rosanne L.; Lopresti, Charles A.; Miley, Terri B.; Nichols, William E.; Strenge, Dennis L.
2004-09-14
This document contains detailed user instructions for a suite of utility codes developed for Rev. 1 of the Systems Assessment Capability. The suite of computer codes for Rev. 1 of Systems Assessment Capability performs many functions.
Comparison of thick-target (alpha,n) yield calculation codes
Fernandes, Ana C.; Kling, Andreas; Vlaskin, Gennadiy N.
2017-09-01
Neutron production yields and energy distributions from (α,n) reactions in light elements were calculated using three different codes (SOURCES, NEDIS and USD) and compared with the existing experimental data in the 3.5-10 MeV alpha energy range. SOURCES and NEDIS display an agreement between calculated and measured yields in the decay series of 235U, 238U and 232Th within ±10% for most materials. The discrepancy increases with alpha energy but still an agreement of ±20% applies to all materials with reliable elemental production yields (the few exceptions are identified). The calculated neutron energy distributions describe the experimental data, with NEDIS retrieving very well the detailed features. USD generally underestimates the measured yields, in particular for compounds with heavy elements and/or at high alpha energies. The energy distributions exhibit sharp peaks that do not match the observations. These findings may be caused by a poor accounting of the alpha particle energy loss by the code. A big variability was found among the calculated neutron production yields for alphas from Sm decay; the lack of yield measurements for low ( 2 MeV) alphas does not allow to conclude on the codes' accuracy in this energy region.
Emergency Doses (ED) - Revision 3: A calculator code for environmental dose computations
Energy Technology Data Exchange (ETDEWEB)
Rittmann, P.D.
1990-12-01
The calculator program ED (Emergency Doses) was developed from several HP-41CV calculator programs documented in the report Seven Health Physics Calculator Programs for the HP-41CV, RHO-HS-ST-5P (Rittman 1984). The program was developed to enable estimates of offsite impacts more rapidly and reliably than was possible with the software available for emergency response at that time. The ED - Revision 3, documented in this report, revises the inhalation dose model to match that of ICRP 30, and adds the simple estimates for air concentration downwind from a chemical release. In addition, the method for calculating the Pasquill dispersion parameters was revised to match the GENII code within the limitations of a hand-held calculator (e.g., plume rise and building wake effects are not included). The summary report generator for printed output, which had been present in the code from the original version, was eliminated in Revision 3 to make room for the dispersion model, the chemical release portion, and the methods of looping back to an input menu until there is no further no change. This program runs on the Hewlett-Packard programmable calculators known as the HP-41CV and the HP-41CX. The documentation for ED - Revision 3 includes a guide for users, sample problems, detailed verification tests and results, model descriptions, code description (with program listing), and independent peer review. This software is intended to be used by individuals with some training in the use of air transport models. There are some user inputs that require intelligent application of the model to the actual conditions of the accident. The results calculated using ED - Revision 3 are only correct to the extent allowed by the mathematical models. 9 refs., 36 tabs.
ITP. FOR: A code to calculate thermal transients in High Level Waste Tanks
Energy Technology Data Exchange (ETDEWEB)
Kielpinski, A.L.
1992-10-01
A variety of processing operations for high level radioactive waste occur in the High Level Waste Tanks in the H-Area of the Savannah River Site. Thermal design constraints exist on these processes, principally to limit the amount of corrosion inhibitor which must be added to protect the tank and cooling coil materials. The required amount of corrosion inhibitor, which must subsequently be removed prior to trapping the waste in borosilicate glass, increases exponentially with temperature over a fairly narrow range (some tens of degrees Celsius). For this reason, there is a need to model the thermal-hydraulic processes occurring in the waste tanks. A FORTRAN computer code, called ITP.FOR, was written to provide a simple but reasonably accurate analysis tool for plant operation design. The code was specifically written to model Tank 48, in which the In-Tank Precipitation (ITP) process of precipitating radioactive cesium will be initiated. Although the ITP.FOR code was written as personal-use software for scoping design calculations for Tank 48, the current intent is to extend the code's applicability to other H-Area waste tanks, and to certify the code in accordance with the NRTSC Quality Assurance requirements for critical-use software (1Q-34, 1991). Since the code's capabilities have generated some interest to date, the present report is presented as interim documentation of the code's mathematical models. This documentation will eventually be supplanted by the formal documentation of the expanded and benchmarked code.
Desch, Steven; Lorenzo, Alejandro; Ko, Byeongkwan
2016-06-01
We present a computer code we have written for general release that calculates the interior structure and mass-radius relationships of solid exoplanets up to a few Earth masses. The basic algorithm is that of Seager et al. (2007), Zeng & Sasselov (2013) and Dorn et al. (2015): the code integrates the 1-D (spherical) equation of hydrostatic equilibrium to find pressure in shells of various depths assuming a gravitational acceleration, uses the bulk modulus of the materials as inputs to an equation of state to convert pressures into density and volume in each shell, recomputes the shell thicknesses and gravitational acceleration, and iterates the solution to convergence. Unlike most existing codes, we do not impose a particular mineralogy in each shell. Instead we adopt the approach of Dorn et al. (2015), in which we impose a stoichiometry in each shell; for rocky shells and the metal core the code calls the PerpleX code (Connolly et al. 2005) to compute the mineralogy and material properties appropriate to that shell’s stoichiometry, pressure and temperature. Unique attributes of the code are as follows. The mineralogy is complete in the Fe-Mg-Si-O system, including species like FeSi and FeO in the core. We also include FeS (VII) in the core. We have also included an approximate phase diagram for water ice to account for an icy mantle. We also include the effects of adiabatic temperature profiles and a temperature jump at the core-mantle boundary. Finally, we have created a user-friendly interface allowing the code to be downloaded and used as a teaching tool. Results of the code and a demonstration of its use will be presented at the meeting.
Wall-touching kink mode calculations with the M3D code
Energy Technology Data Exchange (ETDEWEB)
Breslau, J. A., E-mail: jbreslau@pppl.gov; Bhattacharjee, A. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08542 (United States)
2015-06-15
This paper seeks to address a controversy regarding the applicability of the 3D nonlinear extended MHD code M3D [W. Park et al., Phys. Plasmas 6, 1796 (1999)] and similar codes to calculations of the electromagnetic interaction of a disrupting tokamak plasma with the surrounding vessel structures. M3D is applied to a simple test problem involving an external kink mode in an ideal cylindrical plasma, used also by the Disruption Simulation Code (DSC) as a model case for illustrating the nature of transient vessel currents during a major disruption. While comparison of the results with those of the DSC is complicated by effects arising from the higher dimensionality and complexity of M3D, we verify that M3D is capable of reproducing both the correct saturation behavior of the free boundary kink and the “Hiro” currents arising when the kink interacts with a conducting tile surface interior to the ideal wall.
Tikhomirov, Georgy; Ternovikh, Mikhail; Smirnov, Anton; Saldikov, Ivan; Bahdanovich, Rynat; Gerasimov, Alexander
2017-09-01
System of test tasks is presented with the fast reactor BN-1200 with nitride fuel as prototype. The system of test tasks includes three test based on different geometric models. Model of fuel element in homogeneous and in heterogeneous form, model of fuel assembly in height-heterogeneous and full heterogeneous form, and modeling of the active core of BN-1200 reactor. Cross-verification of program codes was performed. Transition from simple geometry to more complex one allows to identify the causes of discrepancies in the results during the early stage of cross-verification of codes. This system of tests can be applied for certification of engineering programs based on the method of Monte Carlo to the calculation of full-scale models of the reactor core of the BN series. The developed tasks take into account the basic layout and structural features of the reactor BN-1200. They are intended for study of neutron-physical characteristics, estimation of influence of heterogeneous structure and influence of diffusion approximation. The development of system of test tasks allowed to perform independent testing of programs for calculation of neutron-physical characteristics: engineering programs JARFR and TRIGEX, and codes MCU, TDMCC, and MMK based on the method of Monte Carlo.
CHARADE: A characteristic code for calculating rate-dependent shock-wave response
Energy Technology Data Exchange (ETDEWEB)
Johnson, J.N.; Tonks, D.L.
1991-01-01
In this report we apply spatially one-dimensional methods and simple shock-tracking techniques to the solution of rate-dependent material response under flat-plate-impact conditions. This method of solution eliminates potential confusion of material dissipation with artificial dissipative effects inherent in finite-difference codes, and thus lends itself to accurate calculation of elastic-plastic deformation, shock-to-detonation transition in solid explosives, and shock-induced structural phase transformation. Equations are presented for rate-dependent thermoelastic-plastic deformation for (100) planar shock-wave propagation in materials of cubic symmetry (or higher). Specific numerical calculations are presented for polycrystalline copper using the mechanical threshold stress model of Follansbee and Kocks with transition to dislocation drag. A listing of the CHARADE (for characteristic rate dependence) code and sample input deck are given. 26 refs., 11 figs.
Oliveira, C
2001-01-01
A systematic study of isodose distributions and dose uniformity in sample carriers of the Portuguese Gamma Irradiation Facility was carried out using the MCNP code. The absorbed dose rate, gamma flux per energy interval and average gamma energy were calculated. For comparison purposes, boxes filled with air and 'dummy' boxes loaded with layers of folded and crumpled newspapers to achieve a given value of density were used. The magnitude of various contributions to the total photon spectra, including source-dependent factors, irradiator structures, sample material and other origins were also calculated.
First vapor explosion calculations performed with MC3D thermal-hydraulic code
Energy Technology Data Exchange (ETDEWEB)
Brayer, C.; Berthoud, G. [CEA Centre d`Etudes de Grenoble, 38 (France). Direction des Reacteurs Nucleaires
1998-01-01
This paper presents the first calculations performed with the `explosion` module of the multiphase computer code MC3D, which is devoted to the fine fragmentation and explosion phase of a fuel coolant interaction. A complete description of the physical laws included in this module is given. The fragmentation models, taking into account two fragmentation mechanisms, a thermal one and an hydrodynamic one, are also developed here. Results to some calculations to test the numerical behavior of MC3D and to test the explosion models in 1D or 2D are also presented. (author)
BetaShape: A new code for improved analytical calculations of beta spectra
Mougeot Xavier
2017-01-01
The new code BetaShape has been developed in order to improve the nuclear data related to beta decays. An analytical model was considered, except for the relativistic electron wave functions, for ensuring fast calculations. Output quantities are mean energies, log ft values and beta and neutrino spectra for single and multiple transitions. The uncertainties from the input parameters, read from an ENSDF file, are propagated. A database of experimental shape factors is included. A comparison ov...
Development of Monte Carlo decay gamma-ray transport calculation system
Energy Technology Data Exchange (ETDEWEB)
Sato, Satoshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Kawasaki, Nobuo [Fujitsu Ltd., Tokyo (Japan); Kume, Etsuo [Japan Atomic Energy Research Inst., Center for Promotion of Computational Science and Engineering, Tokai, Ibaraki (Japan)
2001-06-01
In the DT fusion reactor, it is critical concern to evaluate the decay gamma-ray biological dose rates after the reactor shutdown exactly. In order to evaluate the decay gamma-ray biological dose rates exactly, three dimensional Monte Carlo decay gamma-ray transport calculation system have been developed by connecting the three dimensional Monte Carlo particle transport calculation code and the induced activity calculation code. The developed calculation system consists of the following four functions. (1) The operational neutron flux distribution is calculated by the three dimensional Monte Carlo particle transport calculation code. (2) The induced activities are calculated by the induced activity calculation code. (3) The decay gamma-ray source distribution is obtained from the induced activities. (4) The decay gamma-rays are generated by using the decay gamma-ray source distribution, and the decay gamma-ray transport calculation is conducted by the three dimensional Monte Carlo particle transport calculation code. In order to reduce the calculation time drastically, a biasing system for the decay gamma-ray source distribution has been developed, and the function is also included in the present system. In this paper, the outline and the detail of the system, and the execution example are reported. The evaluation for the effect of the biasing system is also reported. (author)
A mean field theory of coded CDMA systems
Energy Technology Data Exchange (ETDEWEB)
Yano, Toru [Graduate School of Science and Technology, Keio University, Hiyoshi, Kohoku-ku, Yokohama-shi, Kanagawa 223-8522 (Japan); Tanaka, Toshiyuki [Graduate School of Informatics, Kyoto University, Yoshida Hon-machi, Sakyo-ku, Kyoto-shi, Kyoto 606-8501 (Japan); Saad, David [Neural Computing Research Group, Aston University, Birmingham B4 7ET (United Kingdom)], E-mail: yano@thx.appi.keio.ac.jp
2008-08-15
We present a mean field theory of code-division multiple-access (CDMA) systems with error-control coding. On the basis of the relation between the free energy and mutual information, we obtain an analytical expression of the maximum spectral efficiency of the coded CDMA system, from which a mean-field description of the coded CDMA system is provided in terms of a bank of scalar Gaussian channels whose variances in general vary at different code symbol positions. Regular low-density parity-check (LDPC)-coded CDMA systems are also discussed as an example of the coded CDMA systems.
Development of a prototype pin-by-pin fine mesh calculation code for BWR core analysis
Energy Technology Data Exchange (ETDEWEB)
Tada, Kenichi; Yamamoto, Akio; Yamane, Yoshihiro [Nagoya University, Nagoya (Japan); Kosaka, Shinya; Hirano, Gou [TEPCO SYSTEMS CORPORATION, Tokyo (Japan)
2008-07-01
A prototype core analysis code for BWR, SUBARU, which is based on the three-dimensional pin-by-pin fine-mesh calculation, is being developed. The SUBARU code has several features, e.g., incorporation of the SP3 transport theory, capability to treat the staggered meshes, and so on. In this paper, to estimate the prediction accuracy of this core analysis code, a hypothetical 2D ABWR core which is consisted by 8x8 low-enrichment UO{sub 2} fuel assembly, 9x9 high-enrichment UO{sub 2} fuel assembly, and 10x10 MOX fuel assembly is analyzed. To investigate the prediction accuracy, we compared the pin-wise fission rate distribution which was obtained by the cell-heterogeneous transport calculation by MOC. To evaluate the computational costs, a hypothetical 3D ABWR core is also used. These results suggest that SUBARU would have enough accuracy and reasonable calculation costs for the reference BWR core analysis when further investigation is taken into account. (authors)
An assessment of the CORCON-MOD3 code. Part 1: Thermal-hydraulic calculations
Energy Technology Data Exchange (ETDEWEB)
Strizhov, V.; Kanukova, V.; Vinogradova, T.; Askenov, E. [Russian Academy of Sciences, Moscow (Russian Federation). Inst. of Nuclear Safety; Nikulshin, V. [Russian Research Center, Moscow (Russian Federation). Kurchatov Inst.
1996-09-01
This report deals with the subject of CORCON-Mod3 code validation (thermal-hydraulic modeling capability only) based on MCCI (molten core concrete interaction) experiments conducted under different programs in the past decade. Thermal-hydraulic calculations (i.e., concrete ablation, melt temperature, melt energy, concrete temperature, and condensible and non-condensible gas generation) were performed with the code, and compared with the data from 15 experiments, conducted at different scales using both simulant (metallic and oxidic) and prototypic melt materials, using different concrete types, and with and without an overlying water pool. Sensitivity studies were performed in a few cases involving, for example, heat transfer from melt to concrete, condensed phase chemistry, etc. Further, special analysis was performed using the ACE L8 experimental data to illustrate the differences between the experimental and the reactor conditions, and to demonstrate that with proper corrections made to the code, the calculated results were in better agreement with the experimental data. Generally, in the case of dry cavity and metallic melts, CORCON-Mod3 thermal-hydraulic calculations were in good agreement with the test data. For oxidic melts in a dry cavity, uncertainties in heat transfer models played an important role for two melt configurations--a stratified geometry with segregated metal and oxide layers, and a heterogeneous mixture. Some discrepancies in the gas release data were noted in a few cases.
Energy Technology Data Exchange (ETDEWEB)
Pahn, T.; Jonkman, J.; Rolges, R.; Robertson, A.
2012-11-01
Physically measuring the dynamic responses of wind turbine support structures enables the calculation of the applied loads using an inverse procedure. In this process, inverse means deriving the inputs/forces from the outputs/responses. This paper presents results of a numerical verification of such an inverse load calculation. For this verification, the comprehensive simulation code FAST is used. FAST accounts for the coupled dynamics of wind inflow, aerodynamics, elasticity and turbine controls. Simulations are run using a 5-MW onshore wind turbine model with a tubular tower. Both the applied loads due to the instantaneous wind field and the resulting system responses are known from the simulations. Using the system responses as inputs to the inverse calculation, the applied loads are calculated, which in this case are the rotor thrust forces. These forces are compared to the rotor thrust forces known from the FAST simulations. The results of these comparisons are presented to assess the accuracy of the inverse calculation. To study the influences of turbine controls, load cases in normal operation between cut-in and rated wind speed, near rated wind speed and between rated and cut-out wind speed are chosen. The presented study shows that the inverse load calculation is capable of computing very good estimates of the rotor thrust. The accuracy of the inverse calculation does not depend on the control activity of the wind turbine.
Permutation coding technique for image recognition systems.
Kussul, Ernst M; Baidyk, Tatiana N; Wunsch, Donald C; Makeyev, Oleksandr; Martín, Anabel
2006-11-01
A feature extractor and neural classifier for image recognition systems are proposed. The proposed feature extractor is based on the concept of random local descriptors (RLDs). It is followed by the encoder that is based on the permutation coding technique that allows to take into account not only detected features but also the position of each feature on the image and to make the recognition process invariant to small displacements. The combination of RLDs and permutation coding permits us to obtain a sufficiently general description of the image to be recognized. The code generated by the encoder is used as an input data for the neural classifier. Different types of images were used to test the proposed image recognition system. It was tested in the handwritten digit recognition problem, the face recognition problem, and the microobject shape recognition problem. The results of testing are very promising. The error rate for the Modified National Institute of Standards and Technology (MNIST) database is 0.44% and for the Olivetti Research Laboratory (ORL) database it is 0.1%.
A three-dimensional transient calculation for the reactor model RAMONA using the COMMIX-2(V) code
Energy Technology Data Exchange (ETDEWEB)
Weinberg, D. (Kernforschungszentrum Karlsruhe GmbH (Germany). Inst. fuer Angewandte Thermo- und Fluiddynamik (IATF)); Frey, H.H. (Kernforschungszentrum Karlsruhe GmbH (Germany). Inst. fuer Angewandte Thermo- und Fluiddynamik (IATF)); Tschoeke, H. (Kernforschungszentrum Karlsruhe GmbH (Germany). Inst. fuer Angewandte Thermo- und Fluiddynamik (IATF))
1993-04-01
The safety graded decay heat removal system of the European Fast Reactor needs a high availability. This system operates entirely under natural convection. To guarantee a proper design, experiments are carried out to verify thermal-hydraulic computer codes able to predict precisely local temperature loadings of the components and the reactor tank in the transition region from nominal operation under forced convection to the decay heat removal operation. - With the COMMIX-2 (V) code three-dimensional transient calculations have been performed to simulate experiments in the 360 deg. reactor model RAMONA, scaled 1:20 to the reality with water as simulant fluid for sodium. The computed average and local temperatures as well as the velocity distributions show a good agreement with the experimental results. Further efforts are necessary to reduce the computation time. (orig.)
Calculating the n-point correlation function with general and efficient python code
Genier, Fred; Bellis, Matthew
2018-01-01
There are multiple approaches to understanding the evolution of large-scale structure in our universe and with it the role of baryonic matter, dark matter, and dark energy at different points in history. One approach is to calculate the n-point correlation function estimator for galaxy distributions, sometimes choosing a particular type of galaxy, such as luminous red galaxies. The standard way to calculate these estimators is with pair counts (for the 2-point correlation function) and with triplet counts (for the 3-point correlation function). These are O(n2) and O(n3) problems, respectively and with the number of galaxies that will be characterized in future surveys, having efficient and general code will be of increasing importance. Here we show a proof-of-principle approach to the 2-point correlation function that relies on pre-calculating galaxy locations in coarse “voxels”, thereby reducing the total number of necessary calculations. The code is written in python, making it easily accessible and extensible and is open-sourced to the community. Basic results and performance tests using SDSS/BOSS data will be shown and we discuss the application of this approach to the 3-point correlation function.
Comparison of thick-target (alpha,n yield calculation codes
Directory of Open Access Journals (Sweden)
Fernandes Ana C.
2017-01-01
Full Text Available Neutron production yields and energy distributions from (α,n reactions in light elements were calculated using three different codes (SOURCES, NEDIS and USD and compared with the existing experimental data in the 3.5-10 MeV alpha energy range. SOURCES and NEDIS display an agreement between calculated and measured yields in the decay series of 235U, 238U and 232Th within ±10% for most materials. The discrepancy increases with alpha energy but still an agreement of ±20% applies to all materials with reliable elemental production yields (the few exceptions are identified. The calculated neutron energy distributions describe the experimental data, with NEDIS retrieving very well the detailed features. USD generally underestimates the measured yields, in particular for compounds with heavy elements and/or at high alpha energies. The energy distributions exhibit sharp peaks that do not match the observations. These findings may be caused by a poor accounting of the alpha particle energy loss by the code. A big variability was found among the calculated neutron production yields for alphas from Sm decay; the lack of yield measurements for low (~2 MeV alphas does not allow to conclude on the codes’ accuracy in this energy region.
Work Function Calculation For Hafnium- Barium System
Directory of Open Access Journals (Sweden)
K.A. Tursunmetov
2015-08-01
Full Text Available The adsorption process of barium atoms on hafnium is considered. A structural model of the system is presented and on the basis of calculation of interaction between ions dipole system the dependence of the work function on the coating.
Calculations to an IAHR-benchmark test using the CFD-code CFX-4
Energy Technology Data Exchange (ETDEWEB)
Krepper, E.
1998-10-01
The calculation concerns a test, which was defined as a benchmark for 3-D codes by the working group of advanced nuclear reactor types of IAHR (International Association of Hydraulic Research). The test is well documented and detailed measuring results are available. The test aims at the investigation of phenomena, which are important for heat removal at natural circulation conditions in a nuclear reactor. The task for the calculation was the modelling of the forced flow field of a single phase incompressible fluid with consideration of heat transfer and influence of gravity. These phenomena are typical also for other industrial processes. The importance of correct modelling of these phenomena also for other applications is a motivation for performing these calculations. (orig.)
Santoyo, E.; Garcia, A.; Espinosa, G.; Hernandez, I.; Santoyo, S.
2000-03-01
The development and application of the computer code STATIC_TEMP, a useful tool for calculating static formation temperatures from actual bottomhole temperature data logged in geothermal wells is described. STATIC_TEMP is based on five analytical methods which are the most frequently used in the geothermal industry. Conductive and convective heat flow models (radial, spherical/radial and cylindrical/radial) were selected. The computer code is a useful tool that can be reliable used in situ to determine static formation temperatures before or during the completion stages of geothermal wells (drilling and cementing). Shut-in time and bottomhole temperature measurements logged during well completion activities are required as input data. Output results can include up to seven computations of the static formation temperature by each wellbore temperature data set analysed. STATIC_TEMP was written in Fortran-77 Microsoft language for MS-DOS environment using structured programming techniques. It runs on most IBM compatible personal computers. The source code and its computational architecture as well as the input and output files are described in detail. Validation and application examples on the use of this computer code with wellbore temperature data (obtained from specialised literature) and with actual bottomhole temperature data (taken from completion operations of some geothermal wells) are also presented.
BetaShape: A new code for improved analytical calculations of beta spectra
Mougeot, Xavier
2017-09-01
The new code BetaShape has been developed in order to improve the nuclear data related to beta decays. An analytical model was considered, except for the relativistic electron wave functions, for ensuring fast calculations. Output quantities are mean energies, log ft values and beta and neutrino spectra for single and multiple transitions. The uncertainties from the input parameters, read from an ENSDF file, are propagated. A database of experimental shape factors is included. A comparison over the entire ENSDF database with the standard code currently used in nuclear data evaluations shows consistent results for the vast majority of the transitions and highlights the improvements that can be expected with the use of BetaShape.
BetaShape: A new code for improved analytical calculations of beta spectra
Directory of Open Access Journals (Sweden)
Mougeot Xavier
2017-01-01
Full Text Available The new code BetaShape has been developed in order to improve the nuclear data related to beta decays. An analytical model was considered, except for the relativistic electron wave functions, for ensuring fast calculations. Output quantities are mean energies, log ft values and beta and neutrino spectra for single and multiple transitions. The uncertainties from the input parameters, read from an ENSDF file, are propagated. A database of experimental shape factors is included. A comparison over the entire ENSDF database with the standard code currently used in nuclear data evaluations shows consistent results for the vast majority of the transitions and highlights the improvements that can be expected with the use of BetaShape.
Progress on accelerated calculation of 3D MHD equilibrium with the PIES code
Raburn, Daniel; Reiman, Allan; Monticello, Donald
2016-10-01
Continuing progress has been made in accelerating the 3D MHD equilibrium code, PIES, using an external numerical wrapper. The PIES code (Princeton Iterative Equilibrium Solver) is capable of calculating 3D MHD equilibria with islands. The numerical wrapper has been demonstrated to greatly improve the rate of convergence in numerous cases corresponding to equilibria in the TFTR device where magnetic islands are present; the numerical wrapper makes use of a Jacobian-free Newton-Krylov solver along with adaptive preconditioning and a sophisticated subspace-restricted Levenberg backtracking algorithm. The wrapper has recently been improved by automation which combines the preexisting backtracking algorithm with insights gained from the stability of the Picard algorithm traditionally used with PIES. Improved progress logging and stopping criteria have also been incorporated in to the numerical wrapper.
Energy Technology Data Exchange (ETDEWEB)
Pritchett, J.W.
1980-11-01
The essential features of the reservoir codes CHARGR and MUSHRM are described. Solutions obtained for the problem set posed by DOE are presented. CHARGR was used for all six problems; MUSHRM was used for one. These problems are: the 1-D Avdonin solution, the 1-D well test analysis, 2-D flow to a well in fracture/block media, expanding two-phase system with drainage, flow in a 2-D areal reservoir, and flow in a 3-D reservoir. Results for the last problem using both codes are compared. (MHR)
Energy Technology Data Exchange (ETDEWEB)
Wilson, W. B. (William B.); Perry, R. T. (Robert T.); Shores, E. F. (Erik F.); Charlton, W. S. (William S.); Parish, Theodore A.; Estes, G. P. (Guy P.); Brown, T. H. (Thomas H.); Arthur, Edward D. (Edward Dana),; Bozoian, Michael; England, T. R.; Madland, D. G.; Stewart, J. E. (James E.)
2002-01-01
SOURCES 4C is a computer code that determines neutron production rates and spectra from ({alpha},n) reactions, spontaneous fission, and delayed neutron emission due to radionuclide decay. The code is capable of calculating ({alpha},n) source rates and spectra in four types of problems: homogeneous media (i.e., an intimate mixture of a-emitting source material and low-Z target material), two-region interface problems (i.e., a slab of {alpha}-emitting source material in contact with a slab of low-Z target material), three-region interface problems (i.e., a thin slab of low-Z target material sandwiched between {alpha}-emitting source material and low-Z target material), and ({alpha},n) reactions induced by a monoenergetic beam of {alpha}-particles incident on a slab of target material. Spontaneous fission spectra are calculated with evaluated half-life, spontaneous fission branching, and Watt spectrum parameters for 44 actinides. The ({alpha},n) spectra are calculated using an assumed isotropic angular distribution in the center-of-mass system with a library of 107 nuclide decay {alpha}-particle spectra, 24 sets of measured and/or evaluated ({alpha},n) cross sections and product nuclide level branching fractions, and functional {alpha}-particle stopping cross sections for Z < 106. The delayed neutron spectra are taken from an evaluated library of 105 precursors. The code provides the magnitude and spectra, if desired, of the resultant neutron source in addition to an analysis of the'contributions by each nuclide in the problem. LASTCALL, a graphical user interface, is included in the code package.
Criticality calculation of non-ordinary systems
Energy Technology Data Exchange (ETDEWEB)
Kalugin, A. V., E-mail: Kalugin-AV@nrcki.ru; Tebin, V. V. [National Research Centre Kurchatov Institute (Russian Federation)
2016-12-15
The specific features of calculation of the effective multiplication factor using the Monte Carlo method for weakly coupled and non-asymptotic multiplying systems are discussed. Particular examples are considered and practical recommendations on detection and Monte Carlo calculation of systems typical in numerical substantiation of nuclear safety for VVER fuel management problems are given. In particular, the problems of the choice of parameters for the batch mode and the method for normalization of the neutron batch, as well as finding and interpretation of the eigenvalue spectrum for the integral fission matrix, are discussed.
Criticality coefficient calculation for a small PWR using Monte Carlo Transport Code
Energy Technology Data Exchange (ETDEWEB)
Trombetta, Debora M.; Su, Jian, E-mail: dtrombetta@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Chirayath, Sunil S., E-mail: sunilsc@tamu.edu [Department of Nuclear Engineering and Nuclear Security Science and Policy Institute, Texas A and M University, TX (United States)
2015-07-01
Computational models of reactors are increasingly used to predict nuclear reactor physics parameters responsible for reactivity changes which could lead to accidents and losses. In this work, preliminary results for criticality coefficient calculation using the Monte Carlo transport code MCNPX were presented for a small PWR. The computational modeling developed consists of the core with fuel elements, radial reflectors, and control rods inside a pressure vessel. Three different geometries were simulated, a single fuel pin, a fuel assembly and the core, with the aim to compare the criticality coefficients among themselves.The criticality coefficients calculated were: Doppler Temperature Coefficient, Coolant Temperature Coefficient, Coolant Void Coefficient, Power Coefficient, and Control Rod Worth. The coefficient values calculated by the MCNP code were compared with literature results, showing good agreement with reference data, which validate the computational model developed and allow it to be used to perform more complex studies. Criticality Coefficient values for the three simulations done had little discrepancy for almost all coefficients investigated, the only exception was the Power Coefficient. Preliminary results presented show that simple modelling as a fuel assembly can describe changes at almost all the criticality coefficients, avoiding the need of a complex core simulation. (author)
Calculations of the giant-dipole-resonance photoneutrons using a coupled EGS4-morse code
Energy Technology Data Exchange (ETDEWEB)
Liu, J.C.; Nelson, W.R.; Kase, K.R.; Mao, X.S.
1995-10-01
The production and transport of the photoneutrons from the giant-dipoleresonance reaction have been implemented in a coupled EGS4-MORSE code. The total neutron yield (including both the direct neutron and evaporation neutron components) is calculated by folding the photoneutron yield cross sections with the photon track length distribution in the target. Empirical algorithms based on the measurements have been developed to estimate the fraction and energy of the direct neutron component for each photon. The statistical theory in the EVAP4 code, incorporated as a MORSE subroutine, is used to determine the energies of the evaporation neutrons. These represent major improvements over other calculations that assumed no direct neutrons, a constant fraction of direct neutrons, monoenergetic direct neutron, or a constant nuclear temperature for the evaporation neutrons. It was also assumed that the slow neutrons (< 2.5 MeV) are emitted isotropically and the fast neutrons are emitted anisotropically in the form of 1+Csin{sup 2}{theta}, which have a peak emission at 900. Comparisons between the calculated and the measured photoneutron results (spectra of the direct, evaporation and total neutrons; nuclear temperatures; direct neutron fractions) for materials of lead, tungsten, tantalum and copper have been made. The results show that the empirical algorithms, albeit simple, can produce reasonable results over the interested photon energy range.
Energy Technology Data Exchange (ETDEWEB)
Xhonneux, Andre, E-mail: a.xhonneux@fz-juelich.de [Forschungszentrum Jülich, 52425 Jülich (Germany); Institute for Reactor Safety and Reactor Technology RWTH-Aachen, 52064 Aachen (Germany); Allelein, Hans-Josef [Forschungszentrum Jülich, 52425 Jülich (Germany); Institute for Reactor Safety and Reactor Technology RWTH-Aachen, 52064 Aachen (Germany)
2014-05-01
The computer codes FRESCO-I, FRESCO-II, PANAMA and SPATRA developed at Forschungszentrum Jülich in Germany in the early 1980s are essential tools to predict the fission product release from spherical fuel elements and the TRISO fuel performance, respectively, under given normal or accidental conditions. These codes are able to calculate a conservative estimation of the source term, i.e. quantity and duration of radionuclide release. Recently, these codes have been reversed engineered, modernized (FORTRAN 95/2003) and combined to form a consistent code named STACY (Source Term Analysis Code System). STACY will later become a module of the V/HTR Code Package (HCP). In addition, further improvements have been implemented to enable more detailed calculations. For example the distinct temperature profile along the pebble radius is now taken into account and coated particle failure rates can be calculated under normal operating conditions. In addition, the absolute fission product release of an V/HTR pebble bed core can be calculated by using the newly developed burnup code Topological Nuclide Transformation (TNT) replacing the former rudimentary approach. As a new functionality, spatially resolved fission product release calculations for normal operating conditions as well as accident conditions can be performed. In case of a full-core calculation, a large number of individual pebbles which follow a random path through the reactor core can be simulated. The history of the individual pebble is recorded, too. Main input data such as spatially resolved neutron fluxes and fluid dynamics data are provided by the VSOP code. Capabilities of the FRESCO-I and SPATRA code which allow for the simulation of the redistribution of fission products within the primary circuit and the deposition of fission products on graphitic and metallic surfaces are also available in STACY. In this paper, details of the STACY model and first results for its application to the 200 MW(th) HTR
Concatenated coding system with iterated sequential inner decoding
DEFF Research Database (Denmark)
Jensen, Ole Riis; Paaske, Erik
1995-01-01
We describe a concatenated coding system with iterated sequential inner decoding. The system uses convolutional codes of very long constraint length and operates on iterations between an inner Fano decoder and an outer Reed-Solomon decoder......We describe a concatenated coding system with iterated sequential inner decoding. The system uses convolutional codes of very long constraint length and operates on iterations between an inner Fano decoder and an outer Reed-Solomon decoder...
Energy Technology Data Exchange (ETDEWEB)
Mairani, A [University of Pavia, Department of Nuclear and Theoretical Physics, and INFN, via Bassi 6, 27100 Pavia (Italy); Brons, S; Parodi, K [Heidelberg Ion Beam Therapy Center and Department of Radiation Oncology, Im Neuenheimer Feld 450, 69120 Heidelberg (Germany); Cerutti, F; Ferrari, A; Sommerer, F [CERN, 1211 Geneva 23 (Switzerland); Fasso, A [SLAC National Accelerator Laboratory, 2575 Sand Hill Road, Menlo Park, CA 94025 (United States); Kraemer, M; Scholz, M, E-mail: Andrea.Mairani@mi.infn.i [GSI Biophysik, Planck-Str. 1, D-64291 Darmstadt (Germany)
2010-08-07
Clinical Monte Carlo (MC) calculations for carbon ion therapy have to provide absorbed and RBE-weighted dose. The latter is defined as the product of the dose and the relative biological effectiveness (RBE). At the GSI Helmholtzzentrum fuer Schwerionenforschung as well as at the Heidelberg Ion Therapy Center (HIT), the RBE values are calculated according to the local effect model (LEM). In this paper, we describe the approach followed for coupling the FLUKA MC code with the LEM and its application to dose and RBE-weighted dose calculations for a superimposition of two opposed {sup 12}C ion fields as applied in therapeutic irradiations. The obtained results are compared with the available experimental data of CHO (Chinese hamster ovary) cell survival and the outcomes of the GSI analytical treatment planning code TRiP98. Some discrepancies have been observed between the analytical and MC calculations of absorbed physical dose profiles, which can be explained by the differences between the laterally integrated depth-dose distributions in water used as input basic data in TRiP98 and the FLUKA recalculated ones. On the other hand, taking into account the differences in the physical beam modeling, the FLUKA-based biological calculations of the CHO cell survival profiles are found in good agreement with the experimental data as well with the TRiP98 predictions. The developed approach that combines the MC transport/interaction capability with the same biological model as in the treatment planning system (TPS) will be used at HIT to support validation/improvement of both dose and RBE-weighted dose calculations performed by the analytical TPS.
Mandrekas, John
2004-08-01
GTNEUT is a two-dimensional code for the calculation of the transport of neutral particles in fusion plasmas. It is based on the Transmission and Escape Probabilities (TEP) method and can be considered a computationally efficient alternative to traditional Monte Carlo methods. The code has been benchmarked extensively against Monte Carlo and has been used to model the distribution of neutrals in fusion experiments. Program summaryTitle of program: GTNEUT Catalogue identifier: ADTX Program obtainable from: CPC Program Library, Queen's University of Belfast, N. Ireland Program summary URL:http://cpc.cs.qub.ac.uk/summaries/ADTX Computer for which the program is designed and others on which it has been tested: The program was developed on a SUN Ultra 10 workstation and has been tested on other Unix workstations and PCs. Operating systems or monitors under which the program has been tested: Solaris 8, 9, HP-UX 11i, Linux Red Hat v8.0, Windows NT/2000/XP. Programming language used: Fortran 77 Memory required to execute with typical data: 6 219 388 bytes No. of bits in a word: 32 No. of processors used: 1 Has the code been vectorized or parallelized?: No No. of bytes in distributed program, including test data, etc.: 300 709 No. of lines in distributed program, including test data, etc.: 17 365 Distribution format: compressed tar gzip file Keywords: Neutral transport in plasmas, Escape probability methods Nature of physical problem: This code calculates the transport of neutral particles in thermonuclear plasmas in two-dimensional geometric configurations. Method of solution: The code is based on the Transmission and Escape Probability (TEP) methodology [1], which is part of the family of integral transport methods for neutral particles and neutrons. The resulting linear system of equations is solved by standard direct linear system solvers (sparse and non-sparse versions are included). Restrictions on the complexity of the problem: The current version of the code can
Validation of VHTRC calculation benchmark of critical experiment using the MCB code
Directory of Open Access Journals (Sweden)
Stanisz Przemysław
2016-01-01
Full Text Available The calculation benchmark problem Very High Temperature Reactor Critical (VHTR a pin-in-block type core critical assembly has been investigated with the Monte Carlo Burnup (MCB code in order to validate the latest version of Nuclear Data Library based on ENDF format. Executed benchmark has been made on the basis of VHTR benchmark available from the International Handbook of Evaluated Reactor Physics Benchmark Experiments. This benchmark is useful for verifying the discrepancies in keff values between various libraries and experimental values. This allows to improve accuracy of the neutron transport calculations that may help in designing the high performance commercial VHTRs. Almost all safety parameters depend on the accuracy of neutron transport calculation results that, in turn depend on the accuracy of nuclear data libraries. Thus, evaluation of the libraries applicability to VHTR modelling is one of the important subjects. We compared the numerical experiment results with experimental measurements using two versions of available nuclear data (ENDF-B-VII.1 and JEFF-3.2 prepared for required temperatures. Calculations have been performed with the MCB code which allows to obtain very precise representation of complex VHTR geometry, including the double heterogeneity of a fuel element. In this paper, together with impact of nuclear data, we discuss also the impact of different lattice modelling inside the fuel pins. The discrepancies of keff have been successfully observed and show good agreement with each other and with the experimental data within the 1 σ range of the experimental uncertainty. Because some propagated discrepancies observed, we proposed appropriate corrections in experimental constants which can improve the reactivity coefficient dependency. Obtained results confirm the accuracy of the new Nuclear Data Libraries.
Development and testing of a Monte Carlo code system for analysis of ionization chamber responses
Energy Technology Data Exchange (ETDEWEB)
Johnson, J.O.; Gabriel, T.A.
1986-01-01
To predict the perturbation of interactions between radiation and material by the presence of a detector, a differential Monte Carlo computer code system entitled MICAP was developed and tested. This code system determines the neutron, photon, and total response of an ionization chamber to mixed field radiation environments. To demonstrate the ability of MICAP in calculating an ionization chamber response function, a comparison was made to 05S, an established Monte Carlo code extensively used to accurately calibrate liquid organic scintillators. Both code systems modeled an organic scintillator with a parallel beam of monoenergetic neutrons incident on the scintillator. (LEW)
Energy Technology Data Exchange (ETDEWEB)
Ratnam, Challa [Physics Department, New Science College, Ameerpet, Hyderabad (India); Rao, Vadlamudi Lakshmana [Physics Department, New Science College, Ameerpet, Hyderabad (India); Goud, Sivagouni Lachaa [Department of Physics, Osmania University, Hyderabad (India)
2006-10-07
In the present paper, and a series of papers to follow, the Fourier analytical properties of multiple annuli coded aperture (MACA) and complementary multiple annuli coded aperture (CMACA) systems are investigated. First, the transmission function for MACA and CMACA is derived using Fourier methods and, based on the Fresnel-Kirchoff diffraction theory, the formulae for the point spread function are formulated. The PSF maxima and minima are calculated for both the MACA and CMACA systems. The dependence of these properties on the number of zones is studied and reported in this paper.
Ratnam, Challa; Lakshmana Rao, Vadlamudi; Lachaa Goud, Sivagouni
2006-10-01
In the present paper, and a series of papers to follow, the Fourier analytical properties of multiple annuli coded aperture (MACA) and complementary multiple annuli coded aperture (CMACA) systems are investigated. First, the transmission function for MACA and CMACA is derived using Fourier methods and, based on the Fresnel-Kirchoff diffraction theory, the formulae for the point spread function are formulated. The PSF maxima and minima are calculated for both the MACA and CMACA systems. The dependence of these properties on the number of zones is studied and reported in this paper.
A GPU-based Monte Carlo dose calculation code for photon transport in a voxel phantom
Energy Technology Data Exchange (ETDEWEB)
Bellezzo, M.; Do Nascimento, E.; Yoriyaz, H., E-mail: mbellezzo@gmail.br [Instituto de Pesquisas Energeticas e Nucleares / CNEN, Av. Lineu Prestes 2242, Cidade Universitaria, 05508-000 Sao Paulo (Brazil)
2014-08-15
As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo method has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this paper, we present the CUBMC code, a GPU-based Mc photon transport algorithm for dose calculation under the Compute Unified Device Architecture platform. The simulation of physical events is based on the algorithm used in Penelope, and the cross section table used is the one generated by the Material routine, als present in Penelope code. Photons are transported in voxel-based geometries with different compositions. To demonstrate the capabilities of the algorithm developed in the present work four 128 x 128 x 128 voxel phantoms have been considered. One of them is composed by a homogeneous water-based media, the second is composed by bone, the third is composed by lung and the fourth is composed by a heterogeneous bone and vacuum geometry. Simulations were done considering a 6 MeV monoenergetic photon point source. There are two distinct approaches that were used for transport simulation. The first of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon stop in the frontier will be considered depending on the material changing across the photon travel line. Dose calculations using these methods are compared for validation with Penelope and MCNP5 codes. Speed-up factors are compared using a NVidia GTX 560-Ti GPU card against a 2.27 GHz Intel Xeon CPU processor. (Author)
Development of code system for analysis of skyshine dose rate
Energy Technology Data Exchange (ETDEWEB)
Yamamoto, Hidemasa; Mishima, Tsuyoshi; Nakae, Nobuo (Power Reactor and Nuclear Fuel Development Corp., Tokai, Ibaraki (Japan). Tokai Works)
1989-09-01
A code system which can easily and accurately analyze skyshine dose rate of neutron and gamma rays has been developed by combining analysis codes with cross-section libraries. The code system is composed of four kinds of modules and driver-routine and the modules are separately used for each objective. The code system is also equiped with atomic densities for major materials and the function of automatically meshing and so on. The code system has been verified by using the data of available benchmark problems which have already been reported. In this paper, the system developed here is described and the results of verification is reported. (author).
Energy Technology Data Exchange (ETDEWEB)
Madland, D.G.; Arthur, E.D.; Estes, G.P.; Stewart, J.E.; Bozoian, M.; Perry, R.T.; Parish, T.A.; Brown, T.H.; England, T.R.; Wilson, W.B.; Charlton, W.S.
1999-09-01
SOURCES 4A is a computer code that determines neutron production rates and spectra from ({alpha},n) reactions, spontaneous fission, and delayed neutron emission due to the decay of radionuclides. The code is capable of calculating ({alpha},n) source rates and spectra in four types of problems: homogeneous media (i.e., a mixture of {alpha}-emitting source material and low-Z target material), two-region interface problems (i.e., a slab of {alpha}-emitting source material in contact with a slab of low-Z target material), three-region interface problems (i.e., a thin slab of low-Z target material sandwiched between {alpha}-emitting source material and low-Z target material), and ({alpha},n) reactions induced by a monoenergetic beam of {alpha}-particles incident on a slab of target material. Spontaneous fission spectra are calculated with evaluated half-life, spontaneous fission branching, and Watt spectrum parameters for 43 actinides. The ({alpha},n) spectra are calculated using an assumed isotropic angular distribution in the center-of-mass system with a library of 89 nuclide decay {alpha}-particle spectra, 24 sets of measured and/or evaluated ({alpha},n) cross sections and product nuclide level branching fractions, and functional {alpha}-particle stopping cross sections for Z < 106. The delayed neutron spectra are taken from an evaluated library of 105 precursors. The code outputs the magnitude and spectra of the resultant neutron source. It also provides an analysis of the contributions to that source by each nuclide in the problem.
A simple numerical coding system for clinical electrocardiography
Robles de Medina, E.O.; Meijler, F.L.
1974-01-01
A simple numerical coding system for clinical electrocardiography has been developed. This system enables the storage in coded form of the ECG analysis. The code stored on a digital magnetic tape can be used for a computer print-out of the analysis, while the information can be retrieved at any time
Temperature Calculations in the Coastal Modeling System
2017-04-01
System by Honghai Li and Mitchell E. Brown PURPOSE: This Coastal and Hydraulics Engineering Technical Note (CHETN) describes procedures to calculate...strong tidal signals and sufficient wind energy to provide the vertical mixing. Also, the assumption of sufficient energy to mix over the water...of the Corrotoman River is predominated by tidal process with occasional passages of meteorological events. Tide and wind provide sufficient energy
Joint design of QC-LDPC codes for coded cooperation system with joint iterative decoding
Zhang, Shunwai; Yang, Fengfan; Tang, Lei; Ejaz, Saqib; Luo, Lin; Maharaj, B. T.
2016-03-01
In this paper, we investigate joint design of quasi-cyclic low-density-parity-check (QC-LDPC) codes for coded cooperation system with joint iterative decoding in the destination. First, QC-LDPC codes based on the base matrix and exponent matrix are introduced, and then we describe two types of girth-4 cycles in QC-LDPC codes employed by the source and relay. In the equivalent parity-check matrix corresponding to the jointly designed QC-LDPC codes employed by the source and relay, all girth-4 cycles including both type I and type II are cancelled. Theoretical analysis and numerical simulations show that the jointly designed QC-LDPC coded cooperation well combines cooperation gain and channel coding gain, and outperforms the coded non-cooperation under the same conditions. Furthermore, the bit error rate performance of the coded cooperation employing jointly designed QC-LDPC codes is better than those of random LDPC codes and separately designed QC-LDPC codes over AWGN channels.
On Analyzing LDPC Codes over Multiantenna MC-CDMA System
Directory of Open Access Journals (Sweden)
S. Suresh Kumar
2014-01-01
Full Text Available Multiantenna multicarrier code-division multiple access (MC-CDMA technique has been attracting much attention for designing future broadband wireless systems. In addition, low-density parity-check (LDPC code, a promising near-optimal error correction code, is also being widely considered in next generation communication systems. In this paper, we propose a simple method to construct a regular quasicyclic low-density parity-check (QC-LDPC code to improve the transmission performance over the precoded MC-CDMA system with limited feedback. Simulation results show that the coding gain of the proposed QC-LDPC codes is larger than that of the Reed-Solomon codes, and the performance of the multiantenna MC-CDMA system can be greatly improved by these QC-LDPC codes when the data rate is high.
Monte Carlo dose calculation algorithm on a distributed system
Chauvie, Stéphane; Dominoni, Matteo; Marini, Piergiorgio; Stasi, Michele; Pia, Maria Grazia; Scielzo, Giuseppe
2003-09-01
The main goal of modern radiotherapy, such as 3D conformal radiotherapy and intensity-modulated radiotherapy is to deliver a high dose to the target volume sparing the surrounding healthy tissue. The accuracy of dose calculation in a treatment planning system is therefore a critical issue. Among many algorithms developed over the last years, those based on Monte Carlo proven to be very promising in terms of accuracy. The most severe obstacle in application to clinical practice is the high time necessary for calculations. We have studied a high performance network of Personal Computer as a realistic alternative to a high-costs dedicated parallel hardware to be used routinely as instruments of evaluation of treatment plans. We set-up a Beowulf Cluster, configured with 4 nodes connected with low-cost network and installed MC code Geant4 to describe our irradiation facility. The MC, once parallelised, was run on the Beowulf Cluster. The first run of the full simulation showed that the time required for calculation decreased linearly increasing the number of distributed processes. The good scalability trend allows both statistically significant accuracy and good time performances. The scalability of the Beowulf Cluster system offers a new instrument for dose calculation that could be applied in clinical practice. These would be a good support particularly in high challenging prescription that needs good calculation accuracy in zones of high dose gradient and great dishomogeneities.
Analysis of the KUCA MEU experiments using the ANL code system
Energy Technology Data Exchange (ETDEWEB)
Shiroya, S.; Hayashi, M.; Kanda, K.; Shibata, T.; Woodruff, W.L.; Matos, J.E.
1982-01-01
This paper provides some preliminary results on the analysis of the KUCA critical experiments using the ANL code system. Since this system was employed in the earlier neutronics calculations for the KUHFR, it is important to assess its capabilities for the KUHFR. The KUHFR has a unique core configuration which is difficult to model precisely with current diffusion theory codes. This paper also provides some results from a finite-element diffusion code (2D-FEM-KUR), which was developed in a cooperative research program between KURRI and JAERI. This code provides the capability for mockup of a complex core configuration as the KUHFR. Using the same group constants generated by the EPRI-CELL code, the results of the 2D-FEM-KUR code are compared with the finite difference diffusion code (DIF3D(2D) which is mainly employed in this analysis.
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.)
Bensadoun, H
2001-02-01
The clinical coding system recently instituted in France, the PMSI (Projet de Médicalisation du Système d'Information), has become an unavoidable element in funding allocations for short-term private and public hospitalization centers. Surgeons must take into serious consideration this controversial medicoeconomic instrument. Coding is a dire time-consuming task but, like the hospitalization or surgery report, is an essential part of the discharge procedure. Coding can in the long run be used to establish pricing by pathology. Surgeons should learn the rules and the logic behind this coding system: which, not being based on a medical rationale, may be somewhat difficult to understand. Choosing the right main diagnosis and the comobidity Items is crucial. Quality homogeneous coding is essential if one expects the health authorities to make good use of the system. Our medical societies have a role to play in promoting and harmonizing the coding technique.
HPLWR equilibrium core design with the KARATE code system
Energy Technology Data Exchange (ETDEWEB)
Maraczy, Cs.; Hegyi, Gy.; Hordosy, G.; Temesvari, E. [KFKI Atomic Energy Research Inst., Hungarian Academy of Sciences, Budapest (Hungary)
2011-07-01
The High Performance Light Water Reactor (HPLWR) is the European version of the various supercritical water cooled reactor proposals. The paper presents the activity of KFKI-AEKI in the field of neutronic core design within the framework of the 'HPLWR Phase 2' FP-6 and the Hungarian 'NUKENERG' projects. As the coolant density along the axial direction shows remarkable change, coupled neutronic- thermohydraulic calculations are essential which take into account the heating of moderator in the special water rods of the assemblies. A parametrized diffusion cross section library was prepared for the HPLWR assembly with the MULTICELL neutronic transport code. The parametrized cross sections are used by the KARATE program system, which was verified for supercritical conditions by comparative Monte Carlo calculations. To design the HPLWR equilibrium core preliminary loadings were assessed, which contain insulated assemblies with Gd burnable absorbers. The fuel assemblies have radial and axial enrichment zoning to reduce hot spots. (author)
Development of sump model for containment hydrogen distribution calculations using CFD code
Energy Technology Data Exchange (ETDEWEB)
Ravva, Srinivasa Rao, E-mail: srini@aerb.gov.in [Indian Institute of Technology-Bombay, Mumbai (India); Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai (India); Iyer, Kannan N. [Indian Institute of Technology-Bombay, Mumbai (India); Gaikwad, A.J. [Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai (India)
2015-12-15
Highlights: • Sump evaporation model was implemented in FLUENT using three different approaches. • Validated the implemented sump evaporation models against TOSQAN facility. • It was found that predictions are in good agreement with the data. • Diffusion based model would be able to predict both condensation and evaporation. - Abstract: Computational Fluid Dynamics (CFD) simulations are necessary for obtaining accurate predictions and local behaviour for carrying out containment hydrogen distribution studies. However, commercially available CFD codes do not have all necessary models for carrying out hydrogen distribution analysis. One such model is sump or suppression pool evaporation model. The water in the sump may evaporate during the accident progression and affect the mixture concentrations in the containment. Hence, it is imperative to study the sump evaporation and its effect. Sump evaporation is modelled using three different approaches in the present work. The first approach deals with the calculation of evaporation flow rate and sump liquid temperature and supplying these quantities through user defined functions as boundary conditions. In this approach, the mean values of the domain are used. In the second approach, the mass, momentum, energy and species sources arise due to the sump evaporation are added to the domain through user defined functions. Cell values adjacent to the sump interface are used in this. Heat transfer between gas and liquid is calculated automatically by the code itself. However, in these two approaches, the evaporation rate was computed using an experimental correlation. In the third approach, the evaporation rate is directly estimated using diffusion approximation. The performance of these three models is compared with the sump behaviour experiment conducted in TOSQAN facility.Classification: K. Thermal hydraulics.
Next generation Zero-Code control system UI
CERN. Geneva
2017-01-01
Developing ergonomic user interfaces for control systems is challenging, especially during machine upgrade and commissioning where several small changes may suddenly be required. Zero-code systems, such as *Inspector*, provide agile features for creating and maintaining control system interfaces. More so, these next generation Zero-code systems bring simplicity and uniformity and brake the boundaries between Users and Developers. In this talk we present *Inspector*, a CERN made Zero-code application development system, and we introduce the major differences and advantages of using Zero-code control systems to develop operational UI.
Zhang, Bintuan; Dang, Bingrong; Wang, Zhuanzi; Wei, Wei; Li, Wenjian
2013-10-01
The skin tissue-equivalent slab reported in the International Commission on Radiological Protection (ICRP) Publication 116 to calculate the localised skin dose conversion coefficients (LSDCCs) was adopted into the Monte Carlo transport code Geant4. The Geant4 code was then utilised for computation of LSDCCs due to a circular parallel beam of monoenergetic electrons, protons and alpha particles electrons and alpha particles are found to be in good agreement with the results using the MCNPX code of ICRP 116 data. The present work thus validates the LSDCC values for both electrons and alpha particles using the Geant4 code.
Communication Systems Simulator with Error Correcting Codes Using MATLAB
Gomez, C.; Gonzalez, J. E.; Pardo, J. M.
2003-01-01
In this work, the characteristics of a simulator for channel coding techniques used in communication systems, are described. This software has been designed for engineering students in order to facilitate the understanding of how the error correcting codes work. To help students understand easily the concepts related to these kinds of codes, a…
Monte Carlo code Serpent calculation of the parameters of the stationary nuclear fission wave
Directory of Open Access Journals (Sweden)
V. M. Khotyayintsev
2017-12-01
Full Text Available n this work, propagation of the stationary nuclear fission wave was simulated for series of fixed power values using Monte Carlo code Serpent. The wave moved in the axial direction in 5 m long cylindrical core of fast reactor with pure 238U raw fuel. Stationary wave mode arises some period later after the wave ignition and lasts sufficiently long to determine kef with high enough accuracy. The velocity characteristic of the reactor was determined as the dependence of the wave velocity on the neutron multiplication factor. As we have recently shown within a one-group diffusion description, the velocity characteristic is two-valued due to the effect of concentration mechanisms, while thermal feedback affects it only quantitatively. The shape and parameters of the velocity characteristic critically affect feasibility of the reactor design since stationary wave solutions of the lower branch are unstable and do not correspond to any real waves in self-regulated reactor, like CANDLE. In this work calculations were performed without taking into account thermal feedback. They confirm that theoretical dependence correctly describes the shape of the velocity characteristic calculated using the results of the Serpent modeling.
Litsarev, Mikhail S.
2013-02-01
A description of the DEPOSIT computer code is presented. The code is intended to calculate total and m-fold electron-loss cross-sections (m is the number of ionized electrons) and the energy T(b) deposited to the projectile (positive or negative ion) during a collision with a neutral atom at low and intermediate collision energies as a function of the impact parameter b. The deposited energy is calculated as a 3D integral over the projectile coordinate space in the classical energy-deposition model. Examples of the calculated deposited energies, ionization probabilities and electron-loss cross-sections are given as well as the description of the input and output data. Program summaryProgram title: DEPOSIT Catalogue identifier: AENP_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AENP_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: GNU General Public License version 3 No. of lines in distributed program, including test data, etc.: 8726 No. of bytes in distributed program, including test data, etc.: 126650 Distribution format: tar.gz Programming language: C++. Computer: Any computer that can run C++ compiler. Operating system: Any operating system that can run C++. Has the code been vectorised or parallelized?: An MPI version is included in the distribution. Classification: 2.4, 2.6, 4.10, 4.11. Nature of problem: For a given impact parameter b to calculate the deposited energy T(b) as a 3D integral over a coordinate space, and ionization probabilities Pm(b). For a given energy to calculate the total and m-fold electron-loss cross-sections using T(b) values. Solution method: Direct calculation of the 3D integral T(b). The one-dimensional quadrature formula of the highest accuracy based upon the nodes of the Yacobi polynomials for the cosθ=x∈[-1,1] angular variable is applied. The Simpson rule for the φ∈[0,2π] angular variable is used. The Newton-Cotes pattern of the seventh order
Energy Technology Data Exchange (ETDEWEB)
Marguet, S.D. [Electricite de France (EDF), 92 - Clamart (France)
1997-12-31
Among all parameters, the fuel temperature has a significant influence on the reactivity of the core, because of the Doppler effect on cross-sections. Most neutronic codes use a straightforward method to calculate an average fuel temperature used in their specific feed-back models. For instance, EDF`s neutronic code COCCINELLE uses the Rowland`s formula using the temperatures of the center and the surface of the pellet. COCCINELLE is coupled to the 3D thermal-hydraulic code THYC with calculates TDoppler with is standard thermal model. In order to improve the accuracy of such calculations, we have developed the coupling of our two latest codes in thermal-hydraulics (THYC) and thermo-mechanics (CYRANO3). THYC calculates two-phase flows in pipes or rod bundles and is used for transient calculations such as steam-line break, boron dilution accidents, DNB predictions, steam generator and condenser studies. CYRANO3 calculates most of the phenomena that take place in the fuel such as: 1) heat transfer induced by nuclear power; 2) thermal expansion of the fuel and the cladding; 3) release of gaseous fission`s products; 4) mechanical interaction between the pellet and the cladding. These two codes are now qualified in their own field and the coupling, using Parallel Virtual Machine (PVM) libraries customized in an home-made-easy-to-use package called CALCIUM, has been validated on `low` configurations (no thermal expansion, constant thermal characteristics) and used on accidental transients such as rod ejection and loss of coolant accident. (K.A.) 7 refs.
IBAR: Interacting boson model calculations for large system sizes
Casperson, R. J.
2012-04-01
Scaling the system size of the interacting boson model-1 (IBM-1) into the realm of hundreds of bosons has many interesting applications in the field of nuclear structure, most notably quantum phase transitions in nuclei. We introduce IBAR, a new software package for calculating the eigenvalues and eigenvectors of the IBM-1 Hamiltonian, for large numbers of bosons. Energies and wavefunctions of the nuclear states, as well as transition strengths between them, are calculated using these values. Numerical errors in the recursive calculation of reduced matrix elements of the d-boson creation operator are reduced by using an arbitrary precision mathematical library. This software has been tested for up to 1000 bosons using comparisons to analytic expressions. Comparisons have also been made to the code PHINT for smaller system sizes. Catalogue identifier: AELI_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AELI_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: GNU General Public License version 3 No. of lines in distributed program, including test data, etc.: 28 734 No. of bytes in distributed program, including test data, etc.: 4 104 467 Distribution format: tar.gz Programming language: C++ Computer: Any computer system with a C++ compiler Operating system: Tested under Linux RAM: 150 MB for 1000 boson calculations with angular momenta of up to L=4 Classification: 17.18, 17.20 External routines: ARPACK (http://www.caam.rice.edu/software/ARPACK/) Nature of problem: Construction and diagonalization of large Hamiltonian matrices, using reduced matrix elements of the d-boson creation operator. Solution method: Reduced matrix elements of the d-boson creation operator have been stored in data files at machine precision, after being recursively calculated with higher than machine precision. The Hamiltonian matrix is calculated and diagonalized, and the requested transition strengths are calculated
Energy Technology Data Exchange (ETDEWEB)
Voronova, O.A.; Grebennikov, A.N.; Zvenigorodskaya, O.A. [Russian Federal Nuclear Center, Arzamas (Russian Federation)] [and others
1996-09-01
The paper briefly presents the method for numerical computation of 3-D group neutron diffusion equation implemented in KORAT 3D code within SATURN complex and oriented to 3-D stationary and nonstationary calculations in nuclear reactor physics including those on up-to-date distributed-memory multiprocessors. The computational results are given obtained with the above mentioned code for multidimensional test reactor problems as well as the results of numerical efficiency studies for the code parallelization on multiprocessors Cray T3D and SP-2. Then the setup is given for 3-D dynamic problem simulating a hypothetical mode of WPBER reactor. The computational results are presented for the above problem which includes the neutron-nuclear processes, thermohydraulic processes and the feedback. The computations used two program complexes: KORAT 3D + multichannel two-velocity two-temperature thermohydraulics code RATEG and READY complex. (author)
Methods and computer codes for nuclear systems calculations
Indian Academy of Sciences (India)
Pramana – Journal of Physics. Current Issue : Vol. 89, Issue 2 · Current Issue Volume 89 | Issue 2. August 2017. Home · Volumes & Issues · Special Issues · Forthcoming Articles · Search · Editorial Board · Information for Authors · Subscription ...
Energy Technology Data Exchange (ETDEWEB)
Poletiko, C.; Hueber, C. [Inst. de Protection et de Surete Nucleaire, C.E. Cadarache, St. Paul-lez-Durance (France); Fabre, B. [CISI, C.E. Cadarache, St. Paul-lez-Durance (France)
1996-12-01
In case of severe nuclear accident, radioactive material may be released into the environment. Among the fission products involved, are the very volatile iodine isotopes. However, the chemical forms are not well known due to the presence of different species in the containment with which iodine may rapidly react to form aerosols, molecular iodine, hydroiodic acid and iodo-organics. Tentative explanations of different mechanisms were performed through benchscale tests. A series of tests has been performed at AEA Harwell (GB) to study parameters such as pH, dose rate, concentration, gas flow rate, temperature in relation to molecular iodine production, under dynamic conditions. Another set of tests has been performed in AECL Whiteshell (CA) to study the behaviour of painted coupons, standing in gas phase or liquid phase or both, with iodine compounds under radiation. The purpose of our paper is to synthesize the data and compare the results to the IODE code calculation. Some parameters of the code were studied to fit the experimental result the best. A law, concerning the reverse reaction of iodide radiolytic oxidation, has been proposed versus: pH, concentrations and gas flow-rate. This law does not apply for dose rate variations. For the study of painted coupons, it has been pointed out that molecular iodine tends to be adsorbed or chemically absorbed on the surface in gas phase, but the mechanism should be more sophisticated in the aqueous phase. The iodo-organics present in liquid phase tend to be partly or totally destroyed by oxidation under radiation (depending upon the dose delivered). These points are discussed. (author) 18 figs., 3 tabs., 15 refs.
Modern Nuclear Data Evaluation with the TALYS Code System
Energy Technology Data Exchange (ETDEWEB)
Koning, A.J., E-mail: koning@nrg.eu [Nuclear Research and Consultancy Group NRG, P.O. Box, 1755 ZG Petten (Netherlands); Rochman, D. [Nuclear Research and Consultancy Group NRG, P.O. Box, 1755 ZG Petten (Netherlands)
2012-12-15
This paper presents a general overview of nuclear data evaluation and its applications as developed at NRG, Petten. Based on concepts such as robustness, reproducibility and automation, modern calculation tools are exploited to produce original nuclear data libraries that meet the current demands on quality and completeness. This requires a system which comprises differential measurements, theory development, nuclear model codes, resonance analysis, evaluation, ENDF formatting, data processing and integral validation in one integrated approach. Software, built around the TALYS code, will be presented in which all these essential nuclear data components are seamlessly integrated. Besides the quality of the basic data and its extensive format testing, a second goal lies in the diversity of processing for different type of users. The implications of this scheme are unprecedented. The most important are: 1. Complete ENDF-6 nuclear data files, in the form of the TENDL library, including covariance matrices, for many isotopes, particles, energies, reaction channels and derived quantities. All isotopic data files are mutually consistent and are supposed to rival those of the major world libraries. 2. More exact uncertainty propagation from basic nuclear physics to applied (reactor) calculations based on a Monte Carlo approach: 'Total' Monte Carlo (TMC), using random nuclear data libraries. 3. Automatic optimization in the form of systematic feedback from integral measurements back to the basic data. This method of work also opens a new way of approaching the analysis of nuclear applications, with consequences in both applied nuclear physics and safety of nuclear installations, and several examples are given here. This applied experience and feedback is integrated in a final step to improve the quality of the nuclear data, to change the users vision and finally to orchestrate their integration into simulation codes.
Shielding calculations for industrial 5/7.5MeV electron accelerators using the MCNP Monte Carlo Code
Peri, Eyal; Orion, Itzhak
2017-09-01
High energy X-rays from accelerators are used to irradiate food ingredients to prevent growth and development of unwanted biological organisms in food, and by that extend the shelf life of the products. The production of X-rays is done by accelerating 5 MeV electrons and bombarding them into a heavy target (high Z). Since 2004, the FDA has approved using 7.5 MeV energy, providing higher production rates with lower treatments costs. In this study we calculated all the essential data needed for a straightforward concrete shielding design of typical food accelerator rooms. The following evaluation is done using the MCNP Monte Carlo code system: (1) Angular dependence (0-180°) of photon dose rate for 5 MeV and 7.5 MeV electron beams bombarding iron, aluminum, gold, tantalum, and tungsten targets. (2) Angular dependence (0-180°) spectral distribution simulations of bremsstrahlung for gold, tantalum, and tungsten bombarded by 5 MeV and 7.5 MeV electron beams. (3) Concrete attenuation calculations in several photon emission angles for the 5 MeV and 7.5 MeV electron beams bombarding a tantalum target. Based on the simulation, we calculated the expected increase in dose rate for facilities intending to increase the energy from 5 MeV to 7.5 MeV, and the concrete width needed to be added in order to keep the existing dose rate unchanged.
Coded diffraction system in X-ray crystallography using a boolean phase coded aperture approximation
Pinilla, Samuel; Poveda, Juan; Arguello, Henry
2018-03-01
Phase retrieval is a problem present in many applications such as optics, astronomical imaging, computational biology and X-ray crystallography. Recent work has shown that the phase can be better recovered when the acquisition architecture includes a coded aperture, which modulates the signal before diffraction, such that the underlying signal is recovered from coded diffraction patterns. Moreover, this type of modulation effect, before the diffraction operation, can be obtained using a phase coded aperture, just after the sample under study. However, a practical implementation of a phase coded aperture in an X-ray application is not feasible, because it is computationally modeled as a matrix with complex entries which requires changing the phase of the diffracted beams. In fact, changing the phase implies finding a material that allows to deviate the direction of an X-ray beam, which can considerably increase the implementation costs. Hence, this paper describes a low cost coded X-ray diffraction system based on block-unblock coded apertures that enables phase reconstruction. The proposed system approximates the phase coded aperture with a block-unblock coded aperture by using the detour-phase method. Moreover, the SAXS/WAXS X-ray crystallography software was used to simulate the diffraction patterns of a real crystal structure called Rhombic Dodecahedron. Additionally, several simulations were carried out to analyze the performance of block-unblock approximations in recovering the phase, using the simulated diffraction patterns. Furthermore, the quality of the reconstructions was measured in terms of the Peak Signal to Noise Ratio (PSNR). Results show that the performance of the block-unblock phase coded apertures approximation decreases at most 12.5% compared with the phase coded apertures. Moreover, the quality of the reconstructions using the boolean approximations is up to 2.5 dB of PSNR less with respect to the phase coded aperture reconstructions.
Noncoherent Spectral Optical CDMA System Using 1D Active Weight Two-Code Keying Codes
Directory of Open Access Journals (Sweden)
Bih-Chyun Yeh
2016-01-01
Full Text Available We propose a new family of one-dimensional (1D active weight two-code keying (TCK in spectral amplitude coding (SAC optical code division multiple access (OCDMA networks. We use encoding and decoding transfer functions to operate the 1D active weight TCK. The proposed structure includes an optical line terminal (OLT and optical network units (ONUs to produce the encoding and decoding codes of the proposed OLT and ONUs, respectively. The proposed ONU uses the modified cross-correlation to remove interferences from other simultaneous users, that is, the multiuser interference (MUI. When the phase-induced intensity noise (PIIN is the most important noise, the modified cross-correlation suppresses the PIIN. In the numerical results, we find that the bit error rate (BER for the proposed system using the 1D active weight TCK codes outperforms that for two other systems using the 1D M-Seq codes and 1D balanced incomplete block design (BIBD codes. The effective source power for the proposed system can achieve −10 dBm, which has less power than that for the other systems.
Selection of a computer code for Hanford low-level waste engineered-system performance assessment
Energy Technology Data Exchange (ETDEWEB)
McGrail, B.P.; Mahoney, L.A.
1995-10-01
Planned performance assessments for the proposed disposal of low-level waste (LLW) glass produced from remediation of wastes stored in underground tanks at Hanford, Washington will require calculations of radionuclide release rates from the subsurface disposal facility. These calculations will be done with the aid of computer codes. Currently available computer codes were ranked in terms of the feature sets implemented in the code that match a set of physical, chemical, numerical, and functional capabilities needed to assess release rates from the engineered system. The needed capabilities were identified from an analysis of the important physical and chemical process expected to affect LLW glass corrosion and the mobility of radionuclides. The highest ranked computer code was found to be the ARES-CT code developed at PNL for the US Department of Energy for evaluation of and land disposal sites.
Energy Technology Data Exchange (ETDEWEB)
Mavroulakis, A.; Trombe, A. [INSA - Genie Civl, Laboratoire d`Etudes Thermiques et Mecaniques, 31 - Toulouse (France)
1996-12-31
This paper presents the main processes which allow to determine and to take into account in terms of form factors, a scene seen from an emitter and projected onto a receiver. The elements that compose the emitter have a triangular shape while no subdivision is made on the receiver. The analytical method used for the calculation of the form factors of one element in front of a polygonal receiver is briefly presented. Two cell configurations are presented, the second one having not convex facets with no prerequisite subdivision. The sums of form factors from one given emitter are less than 0.01 away from the unit value. For each configuration, the influence of obstacles is encoded as change rates of individual form factors. Finally, in order to illustrate the interest of these form factor calculations, an example of computerized simulation applied to a complex cavity is presented. (J.S.) 6 refs.
Eisenbach, Markus; Larkin, Jeff; Lutjens, Justin; Rennich, Steven; Rogers, James H.
2017-02-01
The Locally Self-consistent Multiple Scattering (LSMS) code solves the first principles Density Functional theory Kohn-Sham equation for a wide range of materials with a special focus on metals, alloys and metallic nano-structures. It has traditionally exhibited near perfect scalability on massively parallel high performance computer architectures. We present our efforts to exploit GPUs to accelerate the LSMS code to enable first principles calculations of O(100,000) atoms and statistical physics sampling of finite temperature properties. We reimplement the scattering matrix calculation for GPUs with a block matrix inversion algorithm that only uses accelerator memory. Using the Cray XK7 system Titan at the Oak Ridge Leadership Computing Facility we achieve a sustained performance of 14.5PFlop/s and a speedup of 8.6 compared to the CPU only code.
Energy Technology Data Exchange (ETDEWEB)
Novokhatski, Alexander; /SLAC
2005-12-01
The problem of electromagnetic interaction of a beam and accelerator elements is very important for linear colliders, electron-positron factories, and free electron lasers. Precise calculation of wake fields is required for beam dynamics study in these machines. We describe a method which allows computation of wake fields of the very short bunches. Computer code NOVO was developed based on this method. This method is free of unphysical solutions like ''self-acceleration'' of a bunch head, which is common to well known wake field codes. Code NOVO was used for the wake fields study for many accelerator projects all over the world.
Calculations of reactor-accident consequences, Version 2. CRAC2: computer code user's guide
Energy Technology Data Exchange (ETDEWEB)
Ritchie, L.T.; Johnson, J.D.; Blond, R.M.
1983-02-01
The CRAC2 computer code is a revision of the Calculation of Reactor Accident Consequences computer code, CRAC, developed for the Reactor Safety Study. The CRAC2 computer code incorporates significant modeling improvements in the areas of weather sequence sampling and emergency response, and refinements to the plume rise, atmospheric dispersion, and wet deposition models. New output capabilities have also been added. This guide is to facilitate the informed and intelligent use of CRAC2. It includes descriptions of the input data, the output results, the file structures, control information, and five sample problems.
Energy Technology Data Exchange (ETDEWEB)
Berna, G. A; Bohn, M. P.; Rausch, W. N.; Williford, R. E.; Lanning, D. D.
1981-01-01
FRAPCON-2 is a FORTRAN IV computer code that calculates the steady state response of light Mater reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, deformation, and tai lure histories of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (a) heat conduction through the fuel and cladding, (b) cladding elastic and plastic deformation, (c) fuel-cladding mechanical interaction, (d) fission gas release, (e} fuel rod internal gas pressure, (f) heat transfer between fuel and cladding, (g) cladding oxidation, and (h) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat transfer correlations. FRAPCON-2 is programmed for use on the CDC Cyber 175 and 176 computers. The FRAPCON-2 code Is designed to generate initial conditions for transient fuel rod analysis by either the FRAP-T6 computer code or the thermal-hydraulic code, RELAP4/MOD7 Version 2.
RELAP5/MOD3 code manual: Code structure, system models, and solution methods. Volume 1
Energy Technology Data Exchange (ETDEWEB)
NONE
1995-08-01
The RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents, and operational transients, such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling, approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. RELAP5/MOD3 code documentation is divided into seven volumes: Volume I provides modeling theory and associated numerical schemes.
Energy Technology Data Exchange (ETDEWEB)
Sugino, Kazuteru [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center
1998-07-01
As a tool to perform a fast reactor core calculations with high accuracy, NSHEX the nodal transport calculation code for three-dimensional hexagonal-Z geometry is under development. To improve the practical applicability of NSHEX, for instance, in its application to safety analysis and commercial reactor core design studies, we investigated the basic theory used in it, improved the program performance, and evaluated its applicability to the analysis of commercial reactor cores. The current studies show the following: (1) An improvement in the treatment of radial leakage in the radial nodal coupling equation bettered calculational convergence for safety analysis calculation, so the applicability of NSHEX to safety analysis was improved. (2) As a result of comparison of results from NSHEX and the standard core calculation code, it was confirmed that there was consistency between them. (3) According to the evaluation of the effect due to the difference of calculational condition, it was found that the calculation under appropriate nodal expansion orders and Sn orders correspond to the one under most detailed condition. However further investigation is required to reduce the uncertainty in calculational results due to the treatment of high order flux moments. (4) A whole core version of NSHEX enabling calculation for any FBR core geometry has been developed, this improved general applicability of NSHEX. (5) An investigation of the applicability of the rebalance method to acceleration clarified that this improved calculational convergence and it was effective. (J.P.N.)
Rebuilding for Array Codes in Distributed Storage Systems
Wang, Zhiying; Bruck, Jehoshua
2010-01-01
In distributed storage systems that use coding, the issue of minimizing the communication required to rebuild a storage node after a failure arises. We consider the problem of repairing an erased node in a distributed storage system that uses an EVENODD code. EVENODD codes are maximum distance separable (MDS) array codes that are used to protect against erasures, and only require XOR operations for encoding and decoding. We show that when there are two redundancy nodes, to rebuild one erased systematic node, only 3/4 of the information needs to be transmitted. Interestingly, in many cases, the required disk I/O is also minimized.
Energy Technology Data Exchange (ETDEWEB)
Chung, Young Jong; Kim, Soo Hyoung; Kim, See Darl (and others)
2008-10-15
The TASS/SMR code has been developed with domestic technologies for the safety analysis of the SMART plant which is an integral type pressurized water reactor. It can be applied to the analysis of design basis accidents including non-LOCA (loss of coolant accident) and LOCA of the SMART plant. The TASS/SMR code can be applied to any plant regardless of the structural characteristics of a reactor since the code solves the same governing equations for both the primary and secondary system. The code has been developed to meet the requirements of the safety analysis code. This report describes the overall structure of the TASS/SMR, input processing, and the processes of a steady state and transient calculations. In addition, basic differential equations, finite difference equations, state relationships, and constitutive models are described in the report. First, the conservation equations, a discretization process for numerical analysis, search method for state relationship are described. Then, a core power model, heat transfer models, physical models for various components, and control and trip models are explained.
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.
Ferrero, Mauro; Rérat, Michel; Orlando, Roberto; Dovesi, Roberto
2008-07-15
The Coupled Perturbed Hartree-Fock (CPHF) scheme has been implemented in the CRYSTAL06 program, that uses a gaussian type basis set, for systems periodic in 1D (polymers), 2D (slabs), 3D (crystals) and, as a limiting case, 0D (molecules), which enables comparison with molecular codes. CPHF is applied to the calculation of the polarizability alpha of LiF in different aggregation states: finite and infinite chains, slabs, and cubic crystal. Correctness of the computational scheme for the various dimensionalities and its numerical efficiency are confirmed by the correct trend of alpha: alpha for a finite linear chain containing N LiF units with large N tends to the value for the infinite chain, N parallel chains give the slab value when N is sufficiently large, and N superimposed slabs tend to the bulk value. CPHF results compare well with those obtained with a saw-tooth potential approach, previously implemented in CRYSTAL. High numerical accuracy can easily be achieved at relatively low cost, with the same kind of dependence on the computational parameters as for the SCF cycle. Overall, the cost of one component of the dielectric tensor is roughly the same as for the SCF cycle, and it is dominated by the calculation of two-electron four-center integrals. (c) 2008 Wiley Periodicals, Inc.
Energy Technology Data Exchange (ETDEWEB)
Tavares, Desirée Yael de Sena; Silva, Adilson Costa da; Lima, Zelmo Rodrigues de, E-mail: zelmolima@yahoo.com.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)
2017-07-01
This work proposes to implement the cell calculation considering the fuel burning using the WIMSD-5B code. The cell calculation procedure allows to determine the nuclear parameters present in the multi-group neutron diffusion equation and for this purpose the neutron transport theory is used in a problem with dimensional reduction, but in contrast is considered a large number of groups associated with the neutron spectrum. There are a variety of reactor physics codes that determine the nuclear parameters by solving the neutron transport equation applied to an equivalent cell representing a fuel element. The WIMSD-5B code is a deterministic code that solves the transport equation using collision probability method. The simulation of fuel burning in the cell calculation took into account different nuclear data libraries. The WIMSD-5B code supports several nuclear data libraries and in the present work the following libraries were used: IAEA, ENDFB-VII.1, JENDL3.2, JEFF3.1 and JEF2.2, all formatted for 69 energy groups. (author)
Energy Technology Data Exchange (ETDEWEB)
Ihara, Hitoshi; Katakura, Jun-ichi; Nakagawa, Tsuneo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1995-11-01
In a nuclear reactor radioactive nuclides are generated and depleted with burning up of nuclear fuel. The radioactive nuclides, emitting {gamma} ray and {beta} ray, play role of radioactive source of decay heat in a reactor and radiation exposure. In safety evaluation of nuclear reactor and nuclear fuel cycle, it is needed to estimate the number of nuclides generated in nuclear fuel under various burn-up condition of many kinds of nuclear fuel used in a nuclear reactor. FPGS90 is a code calculating the number of nuclides, decay heat and spectrum of emitted {gamma} ray from fission products produced in a nuclear fuel under the various kinds of burn-up condition. The nuclear data library used in FPGS90 code is the library `JNDC Nuclear Data Library of Fission Products - second version -`, which is compiled by working group of Japanese Nuclear Data Committee for evaluating decay heat in a reactor. The code has a function of processing a so-called evaluated nuclear data file such as ENDF/B, JENDL, ENSDF and so on. It also has a function of making figures of calculated results. Using FPGS90 code it is possible to do all works from making library, calculating nuclide generation and decay heat through making figures of the calculated results. (author).
Energy Technology Data Exchange (ETDEWEB)
Rey, H. K.; Park, S. K.; Kim, D. W.; Kang, S. C.; Jung, H. K.; Park, S. K. [Korea Electric Power Research Institute, Taejon (Korea, Republic of)
2000-10-01
Comparisons of the calculation of the differential pressure through four MOV's between Flowmaster code and SFM model has been presented. The Flowmaster and SFM model basically use 1-D steady-state equation, but for the transient analysis, the Flowmaster uses Joukowsky equation considering the effect of fluid velocity variation and wave speed, while, the SFM model uses quasi-steady equation including fluid inertia effect due to pipe inertia. The maximum differential pressures in opening stroke are almost the same between Flowmaster and SFM model, because the two code have the same steady-state equation. For closing stroke, however, the maximum differential pressure is somewhat different, the Flowmaster code shows higher large estimation than SFM code.
Energy Technology Data Exchange (ETDEWEB)
Endo, Akira; Kim, Eunjoo; Yamaguchi, Yasuhiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2001-10-01
A Monte Carlo code SCINFUL has been utilized for calculating response functions of organic scintillators for high-energy neutron spectroscopy. However, the applicability of SCINFUL is limited to the calculations for cylindrical NE213 and NE110 scintillators. In the present study, SCINFUL-CG was developed by introducing a geometry specifying function and high-energy neutron cross section data into SCINFUL. The geometry package MARS-CG, the extended version of the CG (Combinatorial Geometry), was programmed into SCINFUL-CG to express various geometries of detectors. Neutron spectra in the regions specified by the CG can be evaluated by the track length estimator. The cross section data of silicon, oxygen and aluminum for neutron transport calculation were incorporated up to 100 MeV using the data of LA150 library. Validity of SCINFUL-CG was examined by comparing calculated results with those by SCINFUL and MCNP and experimental data measured using high-energy neutron fields. SCINFUL-CG can be used for the calculations of the response functions and neutron spectra in the organic scintillators in various shapes. The computer code will be applicable to the designs of high-energy neutron spectrometers and neutron monitors using the organic scintillators. The present report describes the new features of SCINFUL-CG and explains how to use the code. (author)
Coding in the mammalian gustatory system.
Carleton, Alan; Accolla, Riccardo; Simon, Sidney A
2010-07-01
To understand gustatory physiology and associated dysfunctions it is important to know how oral taste stimuli are encoded both in the periphery and in taste-related brain centres. The identification of distinct taste receptors, together with electrophysiological recordings and behavioral assessments in response to taste stimuli, suggest that information about distinct taste modalities (e.g. sweet versus bitter) are transmitted from the periphery to the brain via segregated pathways. By contrast, gustatory neurons throughout the brain are more broadly tuned, indicating that ensembles of neurons encode taste qualities. Recent evidence reviewed here suggests that the coding of gustatory stimuli is not immutable, but is dependant on a variety of factors including appetite-regulating molecules and associative learning. Copyright (c) 2010 Elsevier Ltd. All rights reserved.
Energy Technology Data Exchange (ETDEWEB)
Lawrence, G.; Barnard, C.; Viswanathan, V.K.
1986-01-01
Historically, wave optics computer codes have been paraxial in nature. Folded systems could be modeled by ''unfolding'' the optical system. Calculation of optical aberrations is, in general, left for the analyst to do with off-line codes. While such paraxial codes were adequate for the simpler systems being studied 10 years ago, current problems such as phased arrays, ring resonators, coupled resonators, and grazing incidents optics require a major advance in analytical capability. This paper describes extension of the physical optics codes GLAD and GLAD V to include a global coordinate system and exact ray aberration calculations. The global coordinate system allows components to be positioned and rotated arbitrarily. Exact aberrations are calculated for components in aligned or misaligned configurations by using ray tracing to compute optical path differences and diffraction propagation. Optical path lengths between components and beam rotations in complex mirror systems are calculated accurately so that coherent interactions in phased arrays and coupled devices may be treated correctly.
Sihver, L.; Mancusi, D.; Niita, K.; Sato, T.; Townsend, L.; Farmer, C.; Pinsky, L.; Ferrari, A.; Cerutti, F.; Gomes, I.
Particles and heavy ions are used in various fields of nuclear physics, medical physics, and material science, and their interactions with different media, including human tissue and critical organs, have therefore carefully been investigated both experimentally and theoretically since the 1930s. However, heavy-ion transport includes many complex processes and measurements for all possible systems, including critical organs, would be impractical or too expensive; e.g. direct measurements of dose equivalents to critical organs in humans cannot be performed. A reliable and accurate particle and heavy-ion transport code is therefore an essential tool in the design study of accelerator facilities as well as for other various applications. Recently, new applications have also arisen within transmutation and reactor science, space and medicine, especially radiotherapy, and several accelerator facilities are operating or planned for construction. Accurate knowledge of the physics of interaction of particles and heavy ions is also necessary for estimating radiation damage to equipment used on space vehicles, to calculate the transport of the heavy ions in the galactic cosmic ray (GCR) through the interstellar medium, and the evolution of the heavier elements after the Big Bang. Concerns about the biological effect of space radiation and space dosimetry are increasing rapidly due to the perspective of long-duration astronaut missions, both in relation to the International Space Station and to manned interplanetary missions in near future. Radiation protection studies for crews of international flights at high altitude have also received considerable attention in recent years. There is therefore a need to develop accurate and reliable particle and heavy-ion transport codes. To be able to calculate complex geometries, including production and transport of protons, neutrons, and alpha particles, 3-dimensional transport using Monte Carlo (MC) technique must be used. Today
Mohammed-Azizi, B.; Medjadi, D. E.
2007-05-01
We present a new version of the computer program which solves the Schrödinger equation of the stationary states for an average nuclear potential of Woods-Saxon type. In this work, we take specifically into account triaxial (i.e. ellipsoidal) nuclear surfaces. The deformation is specified by the usual Bohr parameters. The calculations are carried out in two stages. In the first, one calculates the representative matrix of the Hamiltonian in the Cartesian oscillator basis. In the second stage one diagonalizes this matrix with the help of subroutines of the EISPACK library. This new version calculates all the eigenvalues up to a given cutoff energy, and gives the components of the corresponding eigenfunctions. For a more convenient handling, these results are stored simultaneously in the computer memory, and on a files. Program summaryTitle of program:Triaxial2007 Catalogue identifier:ADSK_v2_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/ADSK_v2_0 Program obtainable from: CPC Program Library, Queen's University of Belfast, N. Ireland Summary of revision:One input file instead two. Reduced number of input parameters. Storage of eigenvalues and eigenvectors in memory in a very simple way which makes the code very convenient to the user. Reasons for the new version: More convenient handling of the eigenvectors Catalogue number old version: ADSK Catalogue number new version:ADSK_v2_0 Journal: Computer Physics Commun. 156 (2004) 241-282 Licensing provisions: none Computer: PC Pentium 4, 2600 MHz Hard disk: 40 Gb RAM: 256 Mb Swap file: 4 Gb Operating system: WINDOWS XP Software used: Compaq Visual FORTRAN (with full optimizations in the settings project options) Programming language used:Fortran 77/90 (double precision) Number of bits in a word: 32 No. of lines in distributed program, including test data, etc.:4058 No. of bytes in distributed program, including test data, etc.:75 590 Distribution format:tar.gz Nature of the problem: The single particle energies
Energy Technology Data Exchange (ETDEWEB)
Habib, B.; Poumarede, B.; Tola, F.; Barthe, J. [CEA, LIST, Dept Technol Capteur et Signal, F-91191 Gif Sur Yvette, (France)
2010-07-01
The aim of the present study is to demonstrate the potential of accelerated dose calculations, using the fast Monte Carlo (MC) code referred to as PENFAST, rather than the conventional MC code PENELOPE, without losing accuracy in the computed dose. For this purpose, experimental measurements of dose distributions in homogeneous and inhomogeneous phantoms were compared with simulated results using both PENELOPE and PENFAST. The simulations and experiments were performed using a Saturne 43 linac operated at 12 MV (photons), and at 18 MeV (electrons). Pre-calculated phase space files (PSFs) were used as input data to both the PENELOPE and PENFAST dose simulations. Since depth-dose and dose profile comparisons between simulations and measurements in water were found to be in good agreement (within {+-} 1% to 1 mm), the PSF calculation is considered to have been validated. In addition, measured dose distributions were compared to simulated results in a set of clinically relevant, inhomogeneous phantoms, consisting of lung and bone heterogeneities in a water tank. In general, the PENFAST results agree to within a 1% to 1 mm difference with those produced by PENELOPE, and to within a 2% to 2 mm difference with measured values. Our study thus provides a pre-clinical validation of the PENFAST code. It also demonstrates that PENFAST provides accurate results for both photon and electron beams, equivalent to those obtained with PENELOPE. CPU time comparisons between both MC codes show that PENFAST is generally about 9-21 times faster than PENELOPE. (authors)
Energy Technology Data Exchange (ETDEWEB)
Gilles, D
2005-07-01
This report is devoted to illustrate the power of a Monte Carlo (MC) simulation code to study the thermodynamical properties of a plasma, composed of classical point particles at thermodynamical equilibrium. Such simulations can help us to manage successfully the challenge of taking into account 'exactly' all classical correlations between particles due to density effects, unlike analytical or semi-analytical approaches, often restricted to low dense plasmas. MC simulations results allow to cover, for laser or astrophysical applications, a wide range of thermodynamical conditions from more dense (and correlated) to less dense ones (where potentials are long ranged type). Therefore Yukawa potentials, with a Thomas-Fermi temperature- and density-dependent screening length, are used to describe the effective ion-ion potentials. In this report we present two MC codes ('PDE' and 'PUCE') and applications performed with these codes in different fields (spectroscopy, opacity, equation of state). Some examples of them are discussed and illustrated at the end of the report. (author)
14 CFR Sec. 1-4 - System of accounts coding.
2010-01-01
... General Accounting Provisions Sec. 1-4 System of accounts coding. (a) A four digit control number is... sequentially, within blocks, designating basic balance sheet classifications. The first two digits of the four...
ARC Code TI: Optimal Alarm System Design and Implementation
National Aeronautics and Space Administration — An optimal alarm system can robustly predict a level-crossing event that is specified over a fixed prediction horizon. The code contained in this packages provides...
Calculs de doses générées par les rayonnements ionisants principes physiques et codes de calcul
Vivier, Alain
2016-01-01
Cet ouvrage et les codes associés s’adressent aux utilisateurs de sources de rayonnements ionisants : techniciens, ingénieurs de sécurité, personnes compétentes en radioprotection, mais aussi médecins, chercheurs, concepteurs, décideurs… Les contraintes croissantes liées à la radioprotection rendent indispensables l’utilisation de codes de calcul permettant d’évaluer les débits de doses générées par ces sources et la façon dont on peut s’en protéger au mieux. De nombreux codes existent, dont certains restent des références incontournables, mais ils sont relativement complexes à mettre en oeuvre et restent en général réservés aux bureaux d’études. En outre, ces codes sont souvent des « boîtes noires » qui ne permettent pas de comprendre la physique sous-jacente. L’objectif de cet ouvrage est double : - Exposer les principes physiques permettant de comprendre les phénomènes à l’oeuvre lorsque la matière est irradiée par des rayonnements ionisants. Il devient al...
Self-Shielding Treatment to Perform Cell Calculation for Seed Furl In Th/U Pwr Using Dragon Code
Directory of Open Access Journals (Sweden)
Ahmed Amin El Said Abd El Hameed
2015-08-01
Full Text Available Time and precision of the results are the most important factors in any code used for nuclear calculations. Despite of the high accuracy of Monte Carlo codes, MCNP and Serpent, in many cases their relatively long computational time leads to difficulties in using any of them as the main calculation code. Usually, Monte Carlo codes are used only to benchmark the results. The deterministic codes, which are usually used in nuclear reactor’s calculations, have limited precision, due to the approximations in the methods used to solve the multi-group transport equation. Self- Shielding treatment, an algorithm that produces an average cross-section defined over the complete energy domain of the neutrons in a nuclear reactor, is responsible for the biggest error in any deterministic codes. There are mainly two resonance self-shielding models commonly applied: models based on equivalence and dilution and models based on subgroup approach. The fundamental problem with any self-shielding method is that it treats any isotope as there are no other isotopes with resonance present in the reactor. The most practical way to solve this problem is to use multi-energy groups (50-200 that are chosen in a way that allows us to use all major resonances without self-shielding. In this paper, we perform cell calculations, for a fresh seed fuel pin which is used in thorium/uranium reactors, by solving 172 energy group transport equation using the deterministic DRAGON code, for the two types of self-shielding models (equivalence and dilution models and subgroup models Using WIMS-D5 and DRAGON data libraries. The results are then tested by comparing it with the stochastic MCNP5 code. We also tested the sensitivity of the results to a specific change in self-shielding method implemented, for example the effect of applying Livolant-Jeanpierre Normalization scheme and Rimman Integration improvement on the equivalence and dilution method, and the effect of using Ribbon
Code-modulated interferometric imaging system using phased arrays
Chauhan, Vikas; Greene, Kevin; Floyd, Brian
2016-05-01
Millimeter-wave (mm-wave) imaging provides compelling capabilities for security screening, navigation, and bio- medical applications. Traditional scanned or focal-plane mm-wave imagers are bulky and costly. In contrast, phased-array hardware developed for mass-market wireless communications and automotive radar promise to be extremely low cost. In this work, we present techniques which can allow low-cost phased-array receivers to be reconfigured or re-purposed as interferometric imagers, removing the need for custom hardware and thereby reducing cost. Since traditional phased arrays power combine incoming signals prior to digitization, orthogonal code-modulation is applied to each incoming signal using phase shifters within each front-end and two-bit codes. These code-modulated signals can then be combined and processed coherently through a shared hardware path. Once digitized, visibility functions can be recovered through squaring and code-demultiplexing operations. Pro- vided that codes are selected such that the product of two orthogonal codes is a third unique and orthogonal code, it is possible to demultiplex complex visibility functions directly. As such, the proposed system modulates incoming signals but demodulates desired correlations. In this work, we present the operation of the system, a validation of its operation using behavioral models of a traditional phased array, and a benchmarking of the code-modulated interferometer against traditional interferometer and focal-plane arrays.
Multiple Description Coding for Closed Loop Systems over Erasure Channels
DEFF Research Database (Denmark)
Østergaard, Jan; Quevedo, Daniel
2013-01-01
dropouts and delays, we transmit quantized control vectors containing current control values for the decoder as well as future predicted control values. Second, we utilize multiple description coding based on forward error correction codes to further aid in the robustness towards packet erasures......In this paper, we consider robust source coding in closed-loop systems. In particular, we consider a (possibly) unstable LTI system, which is to be stabilized via a network. The network has random delays and erasures on the data-rate limited (digital) forward channel between the encoder (controller...
Mazurier, J; Gouriou, J; Chauvenet, B; Barthe, J
2001-06-01
The BNM-LNHB (formerly BNM-LPRI, the French national standard laboratory for ionizing radiation) is equipped with a SATURNE 43 linear accelerator (GE Medical Systems) dedicated to establishing national references of absorbed dose to water for high-energy photon and electron beams. These standards are derived from a dose measurement with a graphite calorimeter and a transfer procedure to water using Fricke dosimeters. This method has already been used to obtain the reference of absorbed dose to water for cobalt-60 beams. The correction factors rising from the perturbations generated by the dosimeters were determined by Monte Carlo calculations. To meet these applications, the Monte Carlo code PENELOPE was used and user codes were specially developed. The first step consisted of simulating the electron and photon showers produced by primary electrons within the accelerator head to determine the characteristics of the resulting photon beams and absorbed dose distributions in a water phantom. These preliminary computations were described in a previous paper. The second step, described in this paper, deals with the calculation of the perturbation correction factors of the graphite calorimeter and of Fricke dosimeters. To point out possible systematic biases, these correction factors were calculated with another Monte Carlo code, EGS4, widely used for years in the field of dose metrology applications. Comparison of the results showed no significant bias. When they were possible, experimental verifications confirmed the calculated values.
Validation Calculations for the Application of MARS Code to the Safety Analysis of Research Reactors
Energy Technology Data Exchange (ETDEWEB)
Park, Cheol; Kim, H.; Chae, H. T.; Lim, I. C
2006-10-15
In order to investigate the applicability of MARS code to the accident analysis of the HANARO and other RRs, the following test data were simulated. Test data of the HANARO design and operation, Test data of flow instability and void fraction from published documents, IAEA RR transient data in TECDOC-643, Brazilian IEA-R1 experimental data. For the simulation of the HANARO data with finned rod type fuels at low pressure and low temperature conditions, MARS code, developed for the transient analysis of power reactors, was modified. Its prediction capability was assessed against the experimental data for the HANARO. From the assessment results, it can be said that the modified MARS code could be used for analyzing the thermal hydraulic transient of the HANARO. Some other simulations such as flow instability test and reactor transients were also done for the application of MARS code to RRs with plate type fuels. In the simulation for these cases, no modification was made. The results of simulated cases show that the MARS code can be used to the transient analysis of RRs with careful considerations. In particular, it seems that an improvement on a void model may be necessary for dealing with the phenomena in high void conditions.
Energy Technology Data Exchange (ETDEWEB)
Neymotin, L. [Brookhaven National Lab., Upton, NY (United States)
1994-04-01
Over the past several years, the OECD/NEA and CEC sponsored an international program intercomparing a group of six probabilistic consequence assessment (PCA) codes designed to simulate health and economic consequences of radioactive releases into atmosphere of radioactive materials following severe accidents at nuclear power plants (NPPs): ARANO (Finland), CONDOR (UK), COSYMA (CEC), LENA (Sweden), MACCS (USA), and OSCAAR (Japan). In parallel with this effort, two separate groups performed similar calculations using the MACCS and COSYMA codes. Results produced in the MACCS Users Group (Greece, Italy, Spain, and USA) calculations and their comparison are contained in the present report. Version 1.5.11.1 of the MACCS code was used for the calculations. Good agreement between the results produced in the four participating calculations has been reached, with the exception of the results related to the ingestion pathway dose predictions. The main reason for the scatter in those particular results is attributed to the lack of a straightforward implementation of the specifications for agricultural production and counter-measures criteria provided for the exercise. A significantly smaller scatter in predictions of other consequences was successfully explained by differences in meteorological files and weather sampling, grids, rain distance intervals, dispersion model options, and population distributions.
ATHENA code manual. Volume 1. Code structure, system models, and solution methods
Energy Technology Data Exchange (ETDEWEB)
Carlson, K.E.; Roth, P.A.; Ransom, V.H.
1986-09-01
The ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) code has been developed to perform transient simulation of the thermal hydraulic systems which may be found in fusion reactors, space reactors, and other advanced systems. A generic modeling approach is utilized which permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of a complete facility. Several working fluids are available to be used in one or more interacting loops. Different loops may have different fluids with thermal connections between loops. The modeling theory and associated numerical schemes are documented in Volume I in order to acquaint the user with the modeling base and thus aid effective use of the code. The second volume contains detailed instructions for input data preparation.
Energy Technology Data Exchange (ETDEWEB)
Reshak, Ali H., E-mail: maalidph@yahoo.co.uk [School of Complex Systems, FFWP, CENAKVA - South Bohemia University CB, Nove Hrady 37333 (Czech Republic); School of Material Engineering, Malaysia University of Perlis, P.O. Box 77, d/a Pejabat Pos Besar, 01007 Kangar, Perlis (Malaysia); Jamal, Morteza [School of Complex Systems, FFWP, CENAKVA - South Bohemia University CB, Nove Hrady 37333 (Czech Republic)
2012-12-05
Highlights: Black-Right-Pointing-Pointer A new package for calculating elastic constants of orthorhombic structure is released. Black-Right-Pointing-Pointer The package called ortho-elastic. Black-Right-Pointing-Pointer It is compatible with [FP-(L)APW+lo] method implemented in WIEN2k code. Black-Right-Pointing-Pointer Several orthorhombic structure compounds were used to test the new package. Black-Right-Pointing-Pointer Elastic constants calculated using this package show good agreement with experiment. - Abstract: A new package for calculating the elastic constants of orthorhombic structure is released. The package called ortho-elastic. The formalism of calculating the ortho-elastic constants is described in details. The package is compatible with the highly accurate all-electron full-potential (linearized) augmented plane-wave plus local orbital [FP-(L)APW+lo] method implemented in WIEN2k code. Several orthorhombic structure compounds were used to test the new package. We found that the calculated elastic constants using the new package show better agreement with the available experimental data than the previous theoretical results used different methods. In this package the second-order derivative E{sup Double-Prime }({epsilon}) of polynomial fit E=E({epsilon}) of energy vs strains at zero strain ({epsilon}=0), used to calculate the orthorhombic elastic constants.
Design and implementation of encrypted and decrypted file system based on USBKey and hardware code
Wu, Kehe; Zhang, Yakun; Cui, Wenchao; Jiang, Ting
2017-05-01
To protect the privacy of sensitive data, an encrypted and decrypted file system based on USBKey and hardware code is designed and implemented in this paper. This system uses USBKey and hardware code to authenticate a user. We use random key to encrypt file with symmetric encryption algorithm and USBKey to encrypt random key with asymmetric encryption algorithm. At the same time, we use the MD5 algorithm to calculate the hash of file to verify its integrity. Experiment results show that large files can be encrypted and decrypted in a very short time. The system has high efficiency and ensures the security of documents.
Fuel burnup calculation of Ghana MNSR using ORIGEN2 and REBUS3 codes.
Abrefah, R G; Nyarko, B J B; Fletcher, J J; Akaho, E H K
2013-10-01
Ghana Research Reactor-1 core is to be converted from HEU fuel to LEU fuel in the near future and managing the spent nuclear fuel is very important. A fuel depletion analysis of the GHARR-1 core was performed using ORIGEN2 and REBUS3 codes to estimate the isotopic inventory at end-of-cycle in order to help in the design of an appropriate spent fuel cask. The results obtained for both codes were consistent for U-235 burnup weight percent and Pu-239 build up as a result of burnup. Copyright © 2013 Elsevier Ltd. All rights reserved.
V.S.O.P. (99/05) computer code system
Energy Technology Data Exchange (ETDEWEB)
Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Scherer, W.
2005-11-01
V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to HTRs and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. V.S.O.P.(99 / 05) represents the further development of V.S.O.P. (99). Compared to its precursor, the code system has been improved in many details. Major improvements and extensions have been included concerning the neutron spectrum calculation, the 3-d neutron diffusion options, and the thermal hydraulic section with respect to 'multi-pass'-fuelled pebblebed cores. This latest code version was developed and tested under the WINDOWS-XP - operating system. The storage requirement for the executables and the basic libraries associated with the code amounts to about 15 MB. Another 5 MB are required - if desired - for storage of the source code ({approx}65000 Fortran statements). (orig.)
Energy Technology Data Exchange (ETDEWEB)
Jung, Andreas; Sarkadi, Peter; Schaffrath, Andreas [TUEV NORD SysTec GmbH und Co. KG, Hamburg (Germany)
2010-05-15
Before a numerical code (e.g. ATHLET) is used for simulation of physical phenomena being new or unknown for the code and/or the user, the user ensures the applicability of the code and his own experience of handling with it by means of a so-called validation. Parametric studies with the code are executed for that matter and the results have to be compared with verified experimental data. Corresponding reference values are available in terms of so-called single-effect-tests (e.g. ECTHOR). In this work the system-code ATHLET Mod. 2.1 Cycle A is validated by post test calculation of the ECTHOR experiment due to the above named aspects. With the ECTHOR-tests the clearing of a water-filled model of a loop seal by means of an air-stream was investigated including momentum exchange at the phase interface under adiabatic and atmospheric conditions. The post test calculations show that the analytical results meet the experimental data within the reproducibility of the experiments. Further findings of the parametric studies are: - The experimental results obtained with the system water-air (ECTHOR) can be assigned to a water-steam-system, if the densities of the phases are equal in both cases. - The initial water level in the loop seal has no influence on the results as long as the gas mass flow is increased moderately. - The loop seal is appropriately nodalized if the mean length of the control volumes accords approx. 1.5 tim es the hydraulic pipe diameter. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Jung, Andreas; Sarkadi, Peter; Schaffrath, Andreas [TUEV NORD SysTec GmbH und Co. KG, Hamburg (Germany)
2010-06-15
Before a numerical code (e.g. ATHLET) is used for simulation of physical phenomena being new or unknown for the code and/or the user, the user ensures the applicability of the code and his own experience of handling with it by means of a so-called validation. Parametric studies with the code are executed for that matter und the results have to be compared with verified experimental data. Corresponding reference values are available in terms of so-called single-effect-tests (e.g. ECTHOR). In this work the system-code ATHLET Mod. 2.1 Cycle A is validated by post test calculation of the ECTHOR experiment due to the above named aspects. With the ECTHOR-tests the clearing of a water-filled model of a loop seal by means of an air-stream was investigated including momentum exchange at the phase interface under adiabatic and atmospheric conditions. The post test calculations show that the analytical results meet the experimental data within the reproducibility of the experiments. Further findings of the parametric studies are: - The experimental results obtained with the system water-air (ECTHOR) can be assigned to a water-steam-system, if the densities of the phases are equal in both cases. - The initial water level in the loop seal has no influence on the results as long as the gas mass flow is increased moderately. - The loop seal is appropriately nodalized if the mean length of the control volumes accords approx. 1.5 times the hydraulic pipe diameter. (orig.)
Directory of Open Access Journals (Sweden)
Khedr Ahmed
2005-01-01
Full Text Available The safety evaluation of nuclear power and re search reactors is a very important step before their construction and during their operation. This evaluation based on the best estimate calculations requires qualified codes qualified users, and qualified nodalizations. The effect of code users on the RELAP5 results during the analysis of loss of flow transient in MTR research reactors is presented in this pa per. To clarify this effect, two nodalizations for research reactor different in the simulation of the open water surface boundary conditions of the reactor pool have been used. Very different results are obtained with few choices for code users. The core natural circulation flow with the be ginning of core boiling doesn't stop but in creases. The in creasing in the natural circulation flow shifts out the boiling from the core and the clad temperature decreases be low the local saturation temperature.
Source Code Vulnerabilities in IoT Software Systems
Directory of Open Access Journals (Sweden)
Saleh Mohamed Alnaeli
2017-08-01
Full Text Available An empirical study that examines the usage of known vulnerable statements in software systems developed in C/C++ and used for IoT is presented. The study is conducted on 18 open source systems comprised of millions of lines of code and containing thousands of files. Static analysis methods are applied to each system to determine the number of unsafe commands (e.g., strcpy, strcmp, and strlen that are well-known among research communities to cause potential risks and security concerns, thereby decreasing a system’s robustness and quality. These unsafe statements are banned by many companies (e.g., Microsoft. The use of these commands should be avoided from the start when writing code and should be removed from legacy code over time as recommended by new C/C++ language standards. Each system is analyzed and the distribution of the known unsafe commands is presented. Historical trends in the usage of the unsafe commands of 7 of the systems are presented to show how the studied systems evolved over time with respect to the vulnerable code. The results show that the most prevalent unsafe command used for most systems is memcpy, followed by strlen. These results can be used to help train software developers on secure coding practices so that they can write higher quality software systems.
Ferrero, Mauro; Rérat, Michel; Kirtman, Bernard; Dovesi, Roberto
2008-12-01
A computational scheme for the evaluation of the static first (β) and second (γ) hyperpolarizability tensors of systems periodic in 1D (polymers), 2D (slabs), 3D (crystals), and, as a limiting case, 0D (molecules) has been implemented, within the coupled perturbed Hartree-Fock framework (CPHF), in the CRYSTAL code, which uses a Gaussian type basis set. This generalizes to 2D and 3D the work by Bishop et al. (J. Chem. Phys. 114, 7633 (2001)). CPHF is applied for β and γ (the polarizability tensor α is also reported for completeness) of LiF in different aggregation states: finite and infinite chains, slabs, and cubic crystal. Correctness of the computational scheme and its numerical efficiency are documented by the trend of β and γ for increasing dimensionality: for a finite linear chain containing N LiF units, the hyperpolarizability tends to the infinite chain value at large N, N parallel chains give the slab value when N is sufficiently large, and N superimposed slabs tend to the bulk value. High numerical accuracy can be achieved at relatively low cost, with a dependence on the computational parameters similar to that observed for field-free self-consistent field (SCF) calculations.
Composite system reliability evaluation by stochastic calculation of system operation
Energy Technology Data Exchange (ETDEWEB)
Haubrick, H.-J.; Hinz, H.-J.; Landeck, E. [Dept. of Power Systems and Power Economics (Germany)
1994-12-31
This report describes a new developed probabilistic approach for steady-state composite system reliability evaluation and its exemplary application to a bulk power test system. The new computer program called PHOENIX takes into consideration transmission limitations, outages of lines and power stations and, as a central element, a highly sophisticated model to the dispatcher performing remedial actions after disturbances. The kernel of the new method is a procedure for optimal power flow calculation that has been specially adapted for the use in reliability evaluations under the above mentioned conditions. (author) 11 refs., 8 figs., 1 tab.
Koeman, T.; Offermans, N.S.M.; Christopher-De Vries, Y.; Slottje, P.; Brandt, P.A. van den; Goldbohm, R.A.; Kromhout, H.; Vermeulen, R.
2013-01-01
Background: In epidemiological studies, occupational exposure estimates are often assigned through linkage of job histories to job-exposure matrices (JEMs). However, available JEMs may have a coding system incompatible with the coding system used to code the job histories, necessitating a
14 CFR 234.8 - Calculation of on-time performance codes.
2010-01-01
... reportable flight, except those scheduled to operate three times or less during a month. In addition, each reporting carrier shall assign an on-time performance code to each of its single plane one-stop or multi-stop flights, or portion thereof, that the carrier holds out to the public through a CRS, the last...
Directory of Open Access Journals (Sweden)
G.N. Manturov
2016-09-01
Development of the integral unified nuclear data support system and its implementation in the calculation codes will ensure not only the unification of the procedure for nuclear data preparation, which will allow enhancing reliability of their verification, but, as well, will enhance accuracy and reliability of calculation prediction of all the most important characteristics of the reactors under design, will ensure their licensing compliance, competitiveness and independence from foreign products.
A user's manual for MASH 1. 0: A Monte Carlo Adjoint Shielding Code System
Energy Technology Data Exchange (ETDEWEB)
Johnson, J.O. (ed.)
1992-03-01
The Monte Carlo Adjoint Shielding Code System, MASH, calculates neutron and gamma-ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air-over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system include the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. MASH is the successor to the Vehicle Code System (VCS) initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the dose importance'' of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response a a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user's manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem (input data and selected output edits) for each code.
A user`s manual for MASH 1.0: A Monte Carlo Adjoint Shielding Code System
Energy Technology Data Exchange (ETDEWEB)
Johnson, J.O. [ed.
1992-03-01
The Monte Carlo Adjoint Shielding Code System, MASH, calculates neutron and gamma-ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air-over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system include the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. MASH is the successor to the Vehicle Code System (VCS) initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the ``dose importance`` of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response a a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user`s manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem (input data and selected output edits) for each code.
Nonterminals and codings in defining variations of OL-systems
DEFF Research Database (Denmark)
Skyum, Sven
1974-01-01
The use of nonterminals versus the use of codings in variations of OL-systems is studied. It is shown that the use of nonterminals produces a comparatively low generative capacity in deterministic systems while it produces a comparatively high generative capacity in nondeterministic systems...
AUS98 - The 1998 version of the AUS modular neutronic code system
Energy Technology Data Exchange (ETDEWEB)
Robinson, G.S.; Harrington, B.V
1998-07-01
AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multi-dimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results. Calculations of multi-region resonance shielding, coupled neutron and photon transport, energy deposition, fission product inventory and neutron diffusion are combined within the one code system. The major changes from the previous AUS publications are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM main-frame computers to UNIX workstations This report gives details of all system aspects of AUS and all modules except the POW3D multi-dimensional diffusion module refs., tabs.
SCALE: A modular code system for performing standardized computer analyses for licensing evaluation
Energy Technology Data Exchange (ETDEWEB)
NONE
1997-03-01
This Manual represents Revision 5 of the user documentation for the modular code system referred to as SCALE. The history of the SCALE code system dates back to 1969 when the current Computational Physics and Engineering Division at Oak Ridge National Laboratory (ORNL) began providing the transportation package certification staff at the U.S. Atomic Energy Commission with computational support in the use of the new KENO code for performing criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976 the certification staff relied on the ORNL staff to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analyses of transportation packages. However, the certification staff learned that, with only occasional use of the codes, it was difficult to become proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the move of the certification staff to the U.S. Nuclear Regulatory Commission (NRC), the NRC staff proposed the development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system was born. This manual covers an array of modules written for the SCALE package, consisting of drivers, system libraries, cross section and materials properties libraries, input/output routines, storage modules, and help files.
Energy Technology Data Exchange (ETDEWEB)
Slater, C.O.
1992-01-01
The DRC2 code, which couples MASH or MASHX adjoint leakages with DORT 2-D discrete ordinates forward directional fluences, is described. The forward fluences are allowed to vary both axially and radially over the coupling surface, as opposed to the strictly axial variation allowed by the predecessor DRC code. Input instructions are presented along with descriptions and results from several sample problems. Results from the sample problems are used to compare DRC2 with DRC, DRC2 with DORT, and DRC2 with itself for the case of x-y dependence versus no x-y dependence of the forward fluence. The test problems demonstrate that for small systems DRC and DRC2 give essentially the same results. Some significant differences are noted for larger systems. Additionally, DRC2 results with no x-y dependence of the forward directional fluences are practically the same as those calculated by DRC.
Improvement of JRR-4 core management code system
Energy Technology Data Exchange (ETDEWEB)
Izumo, H.; Watanabe, S.; Nagatomi, H.; Hori, N. [Department of Research Reactor, Tokai Research Establishment, Japan Atomic Energy Institute, Tokai, Ibaraki (Japan)
2000-10-01
In the modification of JRR-4, the fuel was changed from 93% high enrichment uranium aluminized fuel to 20% low enriched uranium silicide fuel in conformity with the framework of reduced enrichment program on JAERI research reactors. As changing of this, JRR-4 core management code system which estimates excess reactivity of core, fuel burn-up and so on, was improved too. It had been difficult for users to operate the former code system because its input-output form was text-form. But, in the new code system (COMMAS-JRR), users are able to operate the code system without using difficult text-form input. The estimation results of excess reactivity of JRR-4 LEU fuel core were showed very good agreements with the measured value. It is the strong points of this new code system to be operated simply by using the windows form pictures act on a personal workstation equip with the graphical-user-interface (GUI), and to estimate accurately the specific characteristics of the LEU core. (author)
Computational system for activity calculation of radiopharmaceuticals
African Journals Online (AJOL)
... this is specially practised in big countries like Brazil where the distance from one state to other is bigger than one country compared to others in continents like Europe. The purpose of this paper is to describe a computational system developed to evaluate the dose of radiopharmaceuticals during the production until the ...
A User's Manual for MASH V1.5 - A Monte Carlo Adjoint Shielding Code System
Energy Technology Data Exchange (ETDEWEB)
C. O. Slater; J. M. Barnes; J. O. Johnson; J.D. Drischler
1998-10-01
The Monte Carlo ~djoint ~ielding Code System, MASH, calculates neutron and gamma- ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air- over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system includes the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. The current version, MASH v 1.5, is the successor to the original MASH v 1.0 code system initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the "dose importance" of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response as a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user's manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem.
Habib, B; Poumarede, B; Tola, F; Barthe, J
2010-01-01
The aim of the present study is to demonstrate the potential of accelerated dose calculations, using the fast Monte Carlo (MC) code referred to as PENFAST, rather than the conventional MC code PENELOPE, without losing accuracy in the computed dose. For this purpose, experimental measurements of dose distributions in homogeneous and inhomogeneous phantoms were compared with simulated results using both PENELOPE and PENFAST. The simulations and experiments were performed using a Saturne 43 linac operated at 12 MV (photons), and at 18 MeV (electrons). Pre-calculated phase space files (PSFs) were used as input data to both the PENELOPE and PENFAST dose simulations. Since depth-dose and dose profile comparisons between simulations and measurements in water were found to be in good agreement (within +/-1% to 1 mm), the PSF calculation is considered to have been validated. In addition, measured dose distributions were compared to simulated results in a set of clinically relevant, inhomogeneous phantoms, consisting of lung and bone heterogeneities in a water tank. In general, the PENFAST results agree to within a 1% to 1 mm difference with those produced by PENELOPE, and to within a 2% to 2 mm difference with measured values. Our study thus provides a pre-clinical validation of the PENFAST code. It also demonstrates that PENFAST provides accurate results for both photon and electron beams, equivalent to those obtained with PENELOPE. CPU time comparisons between both MC codes show that PENFAST is generally about 9-21 times faster than PENELOPE. Copyright 2009 Associazione Italiana di Fisica Medica. Published by Elsevier Ltd. All rights reserved.
3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors
Energy Technology Data Exchange (ETDEWEB)
Langenbuch, S.; Velkov, K. [GRS, Garching (Germany); Lizorkin, M. [Kurchatov-Institute, Moscow (Russian Federation)] [and others
1997-07-01
This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.
Mairani, A; Kraemer, M; Sommerer, F; Parodi, K; Scholz, M; Cerutti, F; Ferrari, A; Fasso, A
2010-01-01
Clinical Monte Carlo (MC) calculations for carbon ion therapy have to provide absorbed and RBE-weighted dose. The latter is defined as the product of the dose and the relative biological effectiveness (RBE). At the GSI Helmholtzzentrum fur Schwerionenforschung as well as at the Heidelberg Ion Therapy Center (HIT), the RBE values are calculated according to the local effect model (LEM). In this paper, we describe the approach followed for coupling the FLUKA MC code with the LEM and its application to dose and RBE-weighted dose calculations for a superimposition of two opposed C-12 ion fields as applied in therapeutic irradiations. The obtained results are compared with the available experimental data of CHO (Chinese hamster ovary) cell survival and the outcomes of the GSI analytical treatment planning code TRiP98. Some discrepancies have been observed between the analytical and MC calculations of absorbed physical dose profiles, which can be explained by the differences between the laterally integrated depth-d...
García-Jerez, Antonio; Piña-Flores, José; Sánchez-Sesma, Francisco J.; Luzón, Francisco; Perton, Mathieu
2016-12-01
During a quarter of a century, the main characteristics of the horizontal-to-vertical spectral ratio of ambient noise HVSRN have been extensively used for site effect assessment. In spite of the uncertainties about the optimum theoretical model to describe these observations, over the last decade several schemes for inversion of the full HVSRN curve for near surface surveying have been developed. In this work, a computer code for forward calculation of H/V spectra based on the diffuse field assumption (DFA) is presented and tested. It takes advantage of the recently stated connection between the HVSRN and the elastodynamic Green's function which arises from the ambient noise interferometry theory. The algorithm allows for (1) a natural calculation of the Green's functions imaginary parts by using suitable contour integrals in the complex wavenumber plane, and (2) separate calculation of the contributions of Rayleigh, Love, P-SV and SH waves as well. The stability of the algorithm at high frequencies is preserved by means of an adaptation of the Wang's orthonormalization method to the calculation of dispersion curves, surface-waves medium responses and contributions of body waves. This code has been combined with a variety of inversion methods to make up a powerful tool for passive seismic surveying.
Joint fractional Fourier analysis of wavefront-coding systems.
Barwick, Shane; Finnigan, Jerome S
2009-01-15
The analysis of wavefront-coding systems is explored via the joint fractional Fourier signal representation (JFF) of the pupil function. The properties of the JFF of the pupil function are presented and are shown to be revealing with regard to the system response to defocus. Numerical examples that illustrate the properties are given.
Programme Code for Projecting of WDM Fiber Optic Sensor Systems
Directory of Open Access Journals (Sweden)
R. Probstner
1993-04-01
Full Text Available Wavelength division multiplex (WDM offers a potentially powerful technique for use within optical fibre sensor systems. The paper deals with short description of methodology and a programme code for WDM fiber optic sensor system projecting with use of CAD.
Development of platform to compare different wall heat transfer packages for system analysis codes
Energy Technology Data Exchange (ETDEWEB)
Kim, Min-Gil; Lee, Won Woong; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of); Shin, Sung Gil [Hanyang University, Seoul (Korea, Republic of)
2016-10-15
System thermal hydraulic (STH) analysis code is used for analyzing and evaluating the safety of a designed nuclear system. The system thermal hydraulic analysis code typically solves mass, momentum and energy conservation equations for multiple phases with sets of selected empirical constitutive equations to close the problem. Several STH codes are utilized in academia, industry and regulators, such as MARS-KS, SPACE, RELAP5, COBRA-TF, TRACE, and so on. Each system thermal hydraulic code consists of different sets of governing equations and correlations. However, the packages and sets of correlations of each code are not compared quantitatively yet. Wall heat transfer mode transition maps of SPACE and MARS-KS have a little difference for the transition from wall nucleate heat transfer mode to wall film heat transfer mode. Both codes have the same heat transfer packages and correlations in most region except for wall film heat transfer mode. Most of heat transfer coefficients calculated for the range of selected variables of SPACE are the same with those of MARS-KS. For the intervals between 500K and 540K of wall temperature, MARS-KS selects the wall film heat transfer mode and Bromley correlation but SPACE select the wall nucleate heat transfer mode and Chen correlation. This is because the transition from nucleate boiling to film boiling of MARS-KS is earlier than SPACE. More detailed analysis of the heat transfer package and flow regime package will be followed in the near future.
Simplified method for calculating SNCR system efficiency
Directory of Open Access Journals (Sweden)
Pronobis Marek
2017-01-01
Full Text Available SNCR (Selective Non-Catalytic Reduction technology is aimed at reducing NOx emissions. SNCR efficiency is appropriately high only for the reaction temperature range called ‘the SNCR temperature window’. It is a narrow temperature range defined in various ways in the literature, which makes it difficult to evaluate the DeNOx system’s efficiency. Therefore, this study attempts to approximate the relationship between SNCR system efficiency and the flue gas temperature. The approximation was performed on the basis of literature data and verified using data from an experiment. Measurements were performed in a Polish boiler with a maximum continuous rating of 230 t/h. The verified, evaluated function could be used to forecast efficiency of SNCR systems in existing units that use urea or ammonia as a reagent. The approximation results are polynomial functions that depend on flue gas temperature, which fit the literature data with the coefficient of determination R2 = 0.83-0.86. Therefore, these equations could be used by the designer or operator of the boiler for preliminary determination of current SNCR system efficiency.
Space charge calculations for sub-three-dimensional particle-in-cell code
Directory of Open Access Journals (Sweden)
Leonid G. Vorobiev
2000-11-01
Full Text Available A novel approach for modeling high-current, charged particle beams in a three-dimensional manner is introduced. While the integration of beam motion equations is done as in completely 3D particle-in-cell codes, the space charge forces are found by an approximately self-consistent inclusion of the transverse and longitudinal fields. The algorithm is dramatically faster than fully 3D algorithms with computational times comparable to 2D field solvers. In addition, a sparser spatial grid and fewer required macroparticles provide significantly reduced memory demands. The proposed sub-3D technique has been verified with good agreement with other independent algorithms.
State of the art of aerolastic codes for wind turbine calculations
Energy Technology Data Exchange (ETDEWEB)
Maribo Pedersen, B. [ed.
1996-09-01
The technological development of modern wind turbines has been dependent on the parallel development of the computational skills of the designers. The combination of the calculation of the flow field around the wind turbine rotor - both far field and near field - and the calculation of the response of the wind turbine structure to the resulting, non-stationary air loads, also known as aero-elastic calculations have now reached a reasonable degree of maturity. At this expert meeting two main points may be clarified. To what level of accuracy can we now determine the behaviour of the different elements of a wind turbine, i.e. how well are we able to compute deflections, fluctuating loads and power output. Which are the main outstanding areas upon which our next research efforts should be focused. (EG)
Energy Technology Data Exchange (ETDEWEB)
Bangart, R.L.; Bell, L.G.; Boegli, J.S.; Burke, W.C.; Lee, J.Y.; Minns, J.L.; Stoddart, P.G.; Weller, R.A.; Collins, J.T.
1978-12-01
The calculational procedures described in the report reflect current NRC staff practice. The methods described will be used in the evaluation of applications for construction permits and operating licenses docketed after January 1, 1979, until this NUREG is revised as a result of additional staff review. The BWR-GALE (Boiling Water Reactor Gaseous and Liquid Effluents) Code is a computerized mathematical model for calculating the release of radioactive material in gaseous and liquid effluents from boiling water reactors (BWRs). The calculations are based on data generated from operating reactors, field tests, laboratory tests, and plant-specific design considerations incorporated to reduce the quantity of radioactive materials that may be released to the environment.
1975-01-01
A system is presented which processes FORTRAN based software systems to surface potential problems before they become execution malfunctions. The system complements the diagnostic capabilities of compilers, loaders, and execution monitors rather than duplicating these functions. Also, it emphasizes frequent sources of FORTRAN problems which require inordinate manual effort to identify. The principle value of the system is extracting small sections of unusual code from the bulk of normal sequences. Code structures likely to cause immediate or future problems are brought to the user's attention. These messages stimulate timely corrective action of solid errors and promote identification of 'tricky' code. Corrective action may require recoding or simply extending software documentation to explain the unusual technique.
Energy Technology Data Exchange (ETDEWEB)
Santoyo, E. [Universidad Nacional Autonoma de Mexico, Centro de Investigacion en Energia, Temixco (Mexico); Garcia, A.; Santoyo, S. [Unidad Geotermia, Inst. de Investigaciones Electricas, Temixco (Mexico); Espinosa, G. [Universidad Autonoma Metropolitana, Co. Vicentina (Mexico); Hernandez, I. [ITESM, Centro de Sistemas de Manufactura, Monterrey (Mexico)
2000-07-01
The development and application of the computer code STATIC{sub T}EMP, a useful tool for calculating static formation temperatures from actual bottomhole temperature data logged in geothermal wells is described. STATIC{sub T}EMP is based on five analytical methods which are the most frequently used in the geothermal industry. Conductive and convective heat flow models (radial, spherical/radial and cylindrical/radial) were selected. The computer code is a useful tool that can be reliably used in situ to determine static formation temperatures before or during the completion stages of geothermal wells (drilling and cementing). Shut-in time and bottomhole temperature measurements logged during well completion activities are required as input data. Output results can include up to seven computations of the static formation temperature by each wellbore temperature data set analysed. STATIC{sub T}EMP was written in Fortran-77 Microsoft language for MS-DOS environment using structured programming techniques. It runs on most IBM compatible personal computers. The source code and its computational architecture as well as the input and output files are described in detail. Validation and application examples on the use of this computer code with wellbore temperature data (obtained from specialised literature) and with actual bottomhole temperature data (taken from completion operations of some geothermal wells) are also presented. (Author)
On the performance of block codes. [in space communication systems
Helgert, H. J.
1973-01-01
It is important to define and evaluate measures which incorporate most, if not all, of the quantities affecting the overall system reliability. For the simple types of block codes normally employed in space communication systems, the complexity of the encoder and decoder is of little consequence, since the use of integrated circuit technology allows the construction of the basic components in an inexpensive fashion. The complexity is essentially independent of the particular code-decoder used. The processing speed is generally a function of the type of logic used and the technology in the construction of the integrated circuits.
Simulation of water hammer phenomena using the system code ATHLET
Energy Technology Data Exchange (ETDEWEB)
Bratfisch, Christoph; Koch, Marco K. [Bochum Univ. (Germany). Reactor Simulation and Safety Group
2017-07-15
Water Hammer Phenomena can endanger the integrity of structures leading to a possible failure of pipes in nuclear power plants as well as in many industrial applications. These phenomena can arise in nuclear power plants in the course of transients and accidents induced by the start-up of auxiliary feed water systems or emergency core cooling systems in combination with rapid acting valves and pumps. To contribute to further development and validation of the code ATHLET (Analysis of Thermalhydraulics of Leaks and Transients), an experiment performed in the test facility Pilot Plant Pipework (PPP) at Fraunhofer UMSICHT is simulated using the code version ATHLET 3.0A.
Calculation of low-cycle fatigue in accordance with the national standard and strength codes
Kontorovich, T. S.; Radin, Yu. A.
2017-08-01
Over the most recent 15 years, the Russian power industry has largely relied on imported equipment manufactured in compliance with foreign standards and procedures. This inevitably necessitates their harmonization with the regulatory documents of the Russian Federation, which include calculations of strength, low cycle fatigue, and assessment of the equipment service life. An important regulatory document providing the engineering foundation for cyclic strength and life assessment for high-load components of the boiler and steamline of a water/steam circuit is RD 10-249-98:2000: Standard Method of Strength Estimation in Stationary Boilers and Steam and Water Piping. In January 2015, the National Standard of the Russian Federation 12952-3:2001 was introduced regulating the issues of design and calculation of the pressure parts of water-tube boilers and auxiliary installations. Thus, there appeared to be two documents simultaneously valid in the same energy field and using different methods for calculating the low-cycle fatigue strength, which leads to different results. In this connection, the current situation can lead to incorrect ideas about the cyclic strength and the service life of high-temperature boiler parts. The article shows that the results of calculations performed in accordance with GOST R 55682.3-2013/EN 12952-3: 2001 are less conservative than the results of the standard RD 10-249-98. Since the calculation of the expected service life of boiler parts should use GOST R 55682.3-2013/EN 12952-3: 2001, it becomes necessary to establish the applicability scope of each of the above documents.
ITER first wall cooling system simulation with the ATHENA code
Energy Technology Data Exchange (ETDEWEB)
Van Hove, W.; Komen, E.; Bodart, A. [Belgatom SA, Brussels (Belgium)] [and others
1997-12-31
This paper presents the simulation of the first wall/shield blanket (FW/SB) cooling system of the ITER reactor by means of the ATHENA code for a number of operational transients and design basis accidents. The ATHENA model used in this study represents one FW/SB cooling system. The major components of the Primary Heat Transfer System (PHTS) and all 15 different types of FW/SB modules are simulated explicitly. In toroidal direction however, identical components are lumped together. The operational transients are analyzed to support the conceptual design of the FW/SB modules and the PHTS components and to identify the requirements on the control systems. The results show that the system does not experience unacceptable conditions and that the proposed control systems are feasible and effective. The design basis accident analyses are performed to show compliance with the safety criteria for these accidents. The resulting mass and energy releases are used in confinement codes to evaluate the radiological releases to the environment (not reported here). The analyses are part of NSRR-1. The code shortcomings observed in these analyses justify further refinement of the code models and correlations, in order to avoid over-conservative results. (author)
Prostate dose calculations for permanent implants using the MCNPX code and the Voxels phantom MAX
Energy Technology Data Exchange (ETDEWEB)
Reis Junior, Juraci Passos dos; Silva, Ademir Xavier da, E-mail: jjunior@con.ufrj.b, E-mail: Ademir@con.ufrj.b [Coordenacao dos Programas de Pos-Graduacao de Engenharia (COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear; Facure, Alessandro N.S., E-mail: facure@cnen.gov.b [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)
2010-07-01
This paper presents the modeling of 80, 88 and 100 of {sup 125}I seeds, punctual and volumetric inserted into the phantom spherical volume representing the prostate and prostate phantom voxels MAX. Starting values of minimum and maximum activity, 0.27 mCi and 0.38 mCi, respectively, were simulated in the Monte Carlo code MCNPX in order to determine whether the final dose, according to the integration of the equation of decay at time t = 0 to t = {infinity} corresponds to the default value set by the AAPM 64 which is 144 Gy. The results showed that consider sources results in doses exceeding the percentage discrepancy of the default value of 200%, while volumetric consider sources result in doses close to 144 Gy. (author)
EBR-II Static Neutronic Calculations by PHISICS / MCNP6 codes
Energy Technology Data Exchange (ETDEWEB)
Paolo Balestra; Carlo Parisi; Andrea Alfonsi
2016-02-01
The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) on the Shutdown Heat Removal Tests (SHRT) performed in the '80s at the Experimental fast Breeder Reactor EBR-II, USA. The scope of the CRP is to improve and validate the simulation tools for the study and the design of the liquid metal cooled fast reactors. Moreover, training of the next generation of fast reactor analysts is being also considered the other scope of the CRP. In this framework, a static neutronic model was developed, using state-of-the art neutron transport codes like SCALE/PHISICS (deterministic solution) and MCNP6 (stochastic solution). Comparison between both solutions is briefly illustrated in this summary.
Directory of Open Access Journals (Sweden)
Recep Ali Dedecan
2013-08-01
Full Text Available The goal of this paper is to review the validity of seismic assessment procedure given in the Turkish Earthquake Code by comparing the assessment results with real structures from Eastern Turkey, where the 2011 Van earthquake occurred. To test the analysis methods for a typically suitable structure, the cultural center building at Erciş with 3 stories, is selected. In order to compare the results of the three different analysis techniques, for an identical earthquake, the ground motion used in analysis was characterized by equivalent elastic earthquake spectra, which were developed from available time history at the nearest construction site. It was found that the damage predictions by using the by Turkish Earthquake Code procedures point out the different level of damages. But, it is concluded that nonlinear time history analysis calculated the best estimation of the damage observed in the site.
Mairani, A; Valente, M; Battistoni, G; Botta, F; Pedroli, G; Ferrari, A; Cremonesi, M; Di Dia, A; Ferrari, M; Fasso, A
2011-01-01
Purpose: The calculation of patient-specific dose distribution can be achieved by Monte Carlo simulations or by analytical methods. In this study, FLUKA Monte Carlo code has been considered for use in nuclear medicine dosimetry. Up to now, FLUKA has mainly been dedicated to other fields, namely high energy physics, radiation protection, and hadrontherapy. When first employing a Monte Carlo code for nuclear medicine dosimetry, its results concerning electron transport at energies typical of nuclear medicine applications need to be verified. This is commonly achieved by means of calculation of a representative parameter and comparison with reference data. Dose point kernel (DPK), quantifying the energy deposition all around a point isotropic source, is often the one. Methods: FLUKA DPKS have been calculated in both water and compact bone for monoenergetic electrons (10-3 MeV) and for beta emitting isotopes commonly used for therapy ((89)Sr, (90)Y, (131)I, (153)Sm, (177)Lu, (186)Re, and (188)Re). Point isotropic...
Energy Technology Data Exchange (ETDEWEB)
D' Acierno, J.; Hermelee, A.; Fredrickson, C.P.; Van Valkenburg, K.
1979-11-01
The coding methodology for creating facility ID's and energy codes from information existing in EIA data systems currently being mapped into the EEMIS data structure is presented. A comprehensive approach is taken to facilitate implementation of EEMIS. A summary of EIA data sources which will be a part of the final system is presented in a table showing the intersection of 19 EIA data systems with the EEMIS data structure. The methodology for establishing ID codes for EIA sources and the corresponding EEMIS facilities in this table is presented. Detailed energy code translations from EIA source systems to the EEMIS energy codes are provided in order to clarify the transfer of energy data from many EIA systems which use different coding schemes. 28 tables.
Analysis of an XADS Target with the System Code TRACE
Energy Technology Data Exchange (ETDEWEB)
Jaeger, Wadim; Sanchez Espinoza, Victor H. [Forschungszentrum Karlsruhe GmbH, Institute for Reactor Safety, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Feng, Bo [Massachusetts Institute of Technology, 77 Massachusetts Avenue, NW12-219, Cambridge, MA 02139 (United States)
2008-07-01
Accelerator-driven systems (ADS) present an option to reduce the radioactive waste of the nuclear industry. The experimental Accelerator-Driven System (XADS) has been designed to investigate the feasibility of using ADS on an industrial scale to burn minor actinides. The target section lies in the middle of the subcritical core and is bombarded by a proton beam to produce spallation neutrons. The thermal energy produced from this reaction requires a heat removal system for the target section. The target is cooled by liquid lead-bismuth-eutectics (LBE) in the primary system which in turn transfers the heat via a heat exchanger (HX) to the secondary coolant, Diphyl THT (DTHT), a synthetic diathermic fluid. Since this design is still in development, a detailed investigation of the system is necessary to evaluate the behavior during normal and transient operations. Due to the lack of experimental facilities and data for ADS, the analyses are mostly done using thermal hydraulic codes. In addition to evaluating the thermal hydraulics of the XADS, this paper also benchmarks a new code developed by the NRC, TRACE, against other established codes. The events used in this study are beam power switch-on/off transients and a loss of heat sink accident. The obtained results from TRACE were in good agreement with the results of various other codes. (authors)
Neutron and gamma ray transport calculations in shielding system
Energy Technology Data Exchange (ETDEWEB)
Masukawa, Fumihiro; Sakamoto, Hiroki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1998-03-01
In the shields for radiation in nuclear facilities, the penetrating holes of various kinds and irregular shapes are made for the reasons of operation, control and others. These penetrating holes and gaps are filled with air or the substances with relatively small shielding performance, and radiation flows out through them, which is called streaming. As the calculation techniques for the shielding design or analysis related to the streaming problem, there are the calculations by simplified evaluation, transport calculation and Monte Carlo method. In this report, the example of calculation by Monte Carlo method which is represented by MCNP code is discussed. A number of variance reduction techniques which seem effective for the analysis of streaming problem were tried. As to the investigation of the applicability of MCNP code to streaming analysis, the object of analysis which are the concrete walls without hole and with horizontal hole, oblique hole and bent oblique hole, the analysis procedure, the composition of concrete, and the conversion coefficient of dose equivalent, and the results of analysis are reported. As for variance reduction technique, cell importance was adopted. (K.I.)
An interactive computer code for calculation of gas-phase chemical equilibrium (EQLBRM)
Pratt, B. S.; Pratt, D. T.
1984-01-01
A user friendly, menu driven, interactive computer program known as EQLBRM which calculates the adiabatic equilibrium temperature and product composition resulting from the combustion of hydrocarbon fuels with air, at specified constant pressure and enthalpy is discussed. The program is developed primarily as an instructional tool to be run on small computers to allow the user to economically and efficiency explore the effects of varying fuel type, air/fuel ratio, inlet air and/or fuel temperature, and operating pressure on the performance of continuous combustion devices such as gas turbine combustors, Stirling engine burners, and power generation furnaces.
Derivation and implementation of the three-region problem in the SOURCES code system
Energy Technology Data Exchange (ETDEWEB)
Charlton, William S.; Perry, Robert T.; Estes, Guy P. [Los Alamos National Laboratory, NM (United States); Parish, Theodore A. [Dept. of Nuclear Engineering, Texas A and M Univ., College Station, TX (United States)
2000-03-01
The SOURCES code system was designed to calculate neutron production rates and spectra in materials due to the decay of radionuclides [specifically from ({alpha},n) reactions, spontaneous fission, and delayed neutron emission]. The current version (SOURCES-3A) is capable of calculating ({alpha},n) source rates and spectra for three types of problems: homogeneous materials, interface problems, and beam problems. Recent interest in ({alpha},n) sources has prompted the development of a fourth scenario: the three-region problem. To allow SOURCES to confront this problem, the relevant equations defining the {alpha}-particle source rates and spectra at each interface and the neutron source rates and spectra per unit area of interface were derived. These equations (in discretized form) were added as a new subroutine to the SOURCES code system (dubbed SOURCES-4A). The new code system was tested by analyzing the results for a simple three-region problem in two limits: with an optically thin 'intermediate region' and with an optically thick 'intermediate region.' To further validate the code system, SOURCES-4A will be experimentally benchmarked as measured data becomes available. (author)
Two-Factor Authentication System based on QR-Codes
Directory of Open Access Journals (Sweden)
Andrey Yunusovich Iskhakov
2014-09-01
Full Text Available The opportunity of two-factor authentication usage in the control systems and access management on the basis of Quick Response codes with one-time passwords is analyzed in the work. The mobile application is proposed to use as a software token.
An EGS4 user code developed for design and optimization of gamma-ray detection systems
Oishi, T; Yoshida, M; Sugita, T
2003-01-01
An EGS4 user code is developed to design and optimize gamma-ray detection systems for various types of radiation sources. The code is fundamentally based on the PRESTA-CG, which is the user code introducing the PRESTA algorithm and a combinatorial geometry method. The additional and existing functions are integrated in the present code, and the handling is simplified. The latest techniques related to the calculation of material cross section are also incorporated in the code for the accurate simulation. The main additional functions are classified into two parts of the definition of radiation sources and the photon tracing. The former includes the functions on the simple handling of source geometry and the use of redefined radiation source. In the latter functions, it is possible to estimate the simultaneous events among plural detectors and the trace of photon in the interested regions. The developed user code is applied to detection systems in order to demonstrate its availability. As the result, it is foun...
THE CALCULATION OF MULTICOMPONENT ROD SYSTEMS BY DECOMPOSITION METHOD
Directory of Open Access Journals (Sweden)
O. S. Raspopov
2008-12-01
Full Text Available For decomposition of two- and three-dimensional rod systems the effective algorithm of dividing a system into blocks and the appropriate technique of coding for states of each from subsystems are developed. It is shown that the structure of multi-dimensional models can be set by means of space matrices based on the research of topological properties of the system graph.
Using a rapidly identifiable access code system in the OR.
Kastner, D G; Weingarten, L
1987-01-01
To ensure that this system works, each staff member makes an entry into the computer or on the master chart when an item is taken from the supply area. He or she is also expected to check for outdated instrumentation and proper placement on the shelf. The coding system has increased the staff's organization and productivity. It has been successful because it uses a numerical system instead of a memory-based system, and because all instrumentation are categorized and stored according to specialty. The simplicity of the system that allows for quicker access to instrumentation also makes it inexpensive to implement.
Fusion PIC code performance analysis on the Cori KNL system
Energy Technology Data Exchange (ETDEWEB)
Koskela, Tuomas S. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). National Energy Research Scientific Computing Center (NERSC); Deslippe, Jack [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). National Energy Research Scientific Computing Center (NERSC); Friesen, Brian [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). National Energy Research Scientific Computing Center (NERSC); Raman, Karthic [INTEL Corp. (United States)
2017-05-25
We study the attainable performance of Particle-In-Cell codes on the Cori KNL system by analyzing a miniature particle push application based on the fusion PIC code XGC1. We start from the most basic building blocks of a PIC code and build up the complexity to identify the kernels that cost the most in performance and focus optimization efforts there. Particle push kernels operate at high AI and are not likely to be memory bandwidth or even cache bandwidth bound on KNL. Therefore, we see only minor benefits from the high bandwidth memory available on KNL, and achieving good vectorization is shown to be the most beneficial optimization path with theoretical yield of up to 8x speedup on KNL. In practice we are able to obtain up to a 4x gain from vectorization due to limitations set by the data layout and memory latency.
What's in a code? Towards a formal account of the relation of ontologies and coding systems.
Rector, Alan L
2007-01-01
Terminologies are increasingly based on "ontologies" developed in description logics and related languages such as the new Web Ontology Language, OWL. The use of description logic has been expected to reduce ambiguity and make it easier determine logical equivalence, deal with negation, and specify EHRs. However, this promise has not been fully realised: in part because early description logics were relatively inexpressive, in part, because the relation between coding systems, EHRs, and ontologies expressed in description logics has not been fully understood. This paper presents a unifying approach using the expressive formalisms available in the latest version of OWL, OWL 1.1.
Energy Technology Data Exchange (ETDEWEB)
Siccama, N.B.; Koning, H
1998-04-01
The International Atomic Energy Agency (IAEA) Co-ordinated Research Programme (CRP) on `Heat Transport and Afterheat Removal for Gas Cooled Reactors under Accident Conditions` has organised benchmark analyses to support verification and validation of analytical tools used by the participants to predict the thermal behaviour of advanced gas cooled reactors during accidents. One of thew benchmark analyses concerns the code-to-code analysis of the Gas Turbine Modular Helium Reactor (GT-MHR) plutonium burner accidents. The GT-MHR is a passive safe, helium cooled, graphite moderated, advanced reactor system with a thermal power of 600 MW that is based on existing technology. The GT-MHR can also be fuelled with plutonium. If the main helium cooling and the auxiliary shut-down cooling systems fail or become unavailable, the core afterheat is removed by radiation and convection inside the reactor vessel and the reactor cavity to the Reactor Cavity Cooling System (RCCS). The objective of the RCCS is to serve as an ultimate heat sink, ensuring the thermal integrity of the core, vessel and critical equipment within the reactor cavity for the entire spectrum of postulated accident sequences. This paper describes the heat transport inside the reactor core to the RCCS. For this purpose, the heat transfer mechanisms as well as the flow patterns inside the core, the reactor pressure vessel, and the cavity have been calculated by the Computational Fluid Dynamics (CFD) code CFX-F3D. The behaviour of the RCCS itself is not described. One calculation considers the full power operation, while two calculations consider Loss Of Forced Convection (LOFC) accidents, one at pressurised conditions and the other depressurised conditions. The heat transfer from the reactor vessel to the environment under normal operation conditions is 2.64 MW. The highest temperature in the core is 1222K, and the average core temperature is 1075K. The highest reactor vessel temperature is 679K. The highest
Neutronic and thermal-hydraulic calculations for the AP-1000 NPP with the MCNP6 and SERPENT codes
Energy Technology Data Exchange (ETDEWEB)
Stefani, Giovanni Laranjo; Maiorino, Jose R.; Santos, Thiago A., E-mail: giovanni.laranjo@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br, E-mail: thiago.santos@ufabc.edu.br [Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais; Rossi, Pedro R., E-mail: pedro.russorossi@gmail.com [FERMIUM - Tecnologia Nuclear, Sao Paulo, SP (Brazil)
2015-07-01
The AP-1000 is an evolutionary PWR reactor designed as an evolution of the AP-600 project. The reactor is already pre-licensed by NRC, and is considered to have achieved high standards of safety, possible short construction time and good economic competitiveness. The core is a 17x17 typical assembly using Zirlo as cladding, 3 different enrichment regions, and is controlled by boron, control banks, and burnable poison. The expected fuel final burnup is 62 MWD/ton U and a cycle of 18 months. In this paper we present results for neutronic and thermal-hydraulic calculations for the AP-1000. We use the MCNP6 and SERPENT codes to calculate the first cycle of operation. The calculated parameters are K{sub eff} at BOL and EOL and its variation with burnup and neutron flux, and reactivity coefficients. The production of transuranic elements such as Pu-239 and Pu-241, and burning fuel are calculated over time. In the work a complete reactor was burned for 450 days with no control elements, boron or burnable poison were considered, these results were compared with data provided by the Westinghouse. The results are compared with those reported in the literature. A simple thermal hydraulic analysis allows verification of thermal limits such as fuel and cladding temperatures, and MDNB. (author)
(n,xnγ) cross sections on actinides versus reaction code calculations
Kerveno, Maëlle; Bacquias, Antoine; Belloni, Francesca; Borcea, Catalin; Capote, Roberto; Dessagne, Philippe; Dupuis, Marc; Henning, Greg; Hilaire, Stéphane; Kawano, Toshihiko; Nankov, Nicolas; Negret, Alexandru; Nyman, Markus; Party, Eliot; Plompen, Arjan; Romain, Pascal; Rouki, Charoula; Rudolf, Gérard; Stanoiu, Mihai
2017-09-01
The experimental setup GRAPhEME (GeRmanium array for Actinides PrEcise MEasurements) has been used at GELINA (EC-JRC, Geel, Belgium) to perform (n,xn γ) cross sections measurements. GRAPhEME has been especially designed to take into account the specific difficulties generated by the use of actinides samples. This work takes place in the context of new nuclear data measurements for nuclear reactor applications. Considering the very tight accuracy requested for new experimental data, special care has been paid to quantify as accurately as possible all the uncertainties from the instruments and the analysis procedure. From the precise (n,xn γ) cross sections produced with GRAPhEME, the use of model calculations is required to obtain (n,xn) cross sections. Beyond the measurements, extensive work on theoretical models is necessary to achieve a better evaluation of the (n,xn) processes. In this paper, we will discuss the final step of the 238U data analysis and present some recent results obtained on 232Th compared to TALYS modellings. A new measurement campaign on 233U has started recently, a first assessment of the recorded data will be presented.
Calculation of Industrial Enterprise Ventilation System by Network Integral Method
Directory of Open Access Journals (Sweden)
Mihienkova Evgeniya I.
2016-01-01
Full Text Available This paper describe a ventilation system calculation of the technology building industrial enterprise. On the basis of the calculation model for the enterprise offered technical decision of ventilation systems, subject to a compliance exchange multiplicity, purification efficiency, decontamination from the work area; provided the required volume of gas extraction from process equipment according to the sanitary standards and environmental requirements. Produced selection of ventilation equipment parameters, solved the problem of the air exchange balancing between ventilation systems to prevent the emergence of parasitic flows between the rooms building. SigmaNet software package was used for the implement the calculation.
The number processing and calculation system: evidence from cognitive neuropsychology.
Salguero-Alcañiz, M P; Alameda-Bailén, J R
2015-04-01
Cognitive neuropsychology focuses on the concepts of dissociation and double dissociation. The performance of number processing and calculation tasks by patients with acquired brain injury can be used to characterise the way in which the healthy cognitive system manipulates number symbols and quantities. The objective of this study is to determine the components of the numerical processing and calculation system. Participants consisted of 6 patients with acquired brain injuries in different cerebral localisations. We used Batería de evaluación del procesamiento numérico y el cálculo, a battery assessing number processing and calculation. Data was analysed using the difference in proportions test. Quantitative numerical knowledge is independent from number transcoding, qualitative numerical knowledge, and calculation. Recodification is independent from qualitative numerical knowledge and calculation. Quantitative numerical knowledge and calculation are also independent functions. The number processing and calculation system comprises at least 4 components that operate independently: quantitative numerical knowledge, number transcoding, qualitative numerical knowledge, and calculation. Therefore, each one may be damaged selectively without affecting the functioning of another. According to the main models of number processing and calculation, each component has different characteristics and cerebral localisations. Copyright © 2013 Sociedad Española de Neurología. Published by Elsevier Espana. All rights reserved.
Chierici, A.; Chirco, L.; Da Vià, R.; Manservisi, S.; Scardovelli, R.
2017-11-01
Nowadays the rapidly-increasing computational power allows scientists and engineers to perform numerical simulations of complex systems that can involve many scales and several different physical phenomena. In order to perform such simulations, two main strategies can be adopted: one may develop a new numerical code where all the physical phenomena of interest are modelled or one may couple existing validated codes. With the latter option, the creation of a huge and complex numerical code is avoided but efficient methods for data exchange are required since the performance of the simulation is highly influenced by its coupling techniques. In this work we propose a new algorithm that can be used for volume and/or boundary coupling purposes for both multiscale and multiphysics numerical simulations. The proposed algorithm is used for a multiscale simulation involving several CFD domains and monodimensional loops. We adopt the overlapping domain strategy, so the entire flow domain is simulated with the system code. We correct the system code solution by matching averaged inlet and outlet fields located at the boundaries of the CFD domains that overlap parts of the monodimensional loop. In particular we correct pressure losses and enthalpy values with source-sink terms that are imposed in the system code equations. The 1D-CFD coupling is a defective one since the CFD code requires point-wise values on the coupling interfaces and the system code provides only averaged quantities. In particular we impose, as inlet boundary conditions for the CFD domains, the mass flux and the mean enthalpy that are calculated by the system code. With this method the mass balance is preserved at every time step of the simulation. The coupling between consecutive CFD domains is not a defective one since with the proposed algorithm we can interpolate the field solutions on the boundary interfaces. We use the MED data structure as the base structure where all the field operations are
Applying Hamming Code to Memory System of Safety Grade PLC (POSAFE-Q) Processor Module
Energy Technology Data Exchange (ETDEWEB)
Kim, Taehee; Hwang, Sungjae; Park, Gangmin [POSCO Nuclear Technology, Seoul (Korea, Republic of)
2013-05-15
If some errors such as inverted bits occur in the memory, instructions and data will be corrupted. As a result, the PLC may execute the wrong instructions or refer to the wrong data. Hamming Code can be considered as the solution for mitigating this mis operation. In this paper, we apply hamming Code, then, we inspect whether hamming code is suitable for to the memory system of the processor module. In this paper, we applied hamming code to existing safety grade PLC (POSAFE-Q). Inspection data are collected and they will be referred for improving the PLC in terms of the soundness. In our future work, we will try to improve time delay caused by hamming calculation. It will include CPLD optimization and memory architecture or parts alteration. In addition to these hamming code-based works, we will explore any methodologies such as mirroring for the soundness of safety grade PLC. Hamming code-based works can correct bit errors, but they have limitation in multi bits errors.
An expert system for the calculation of sample size.
Ebell, M H; Neale, A V; Hodgkins, B J
1994-06-01
Calculation of sample size is a useful technique for researchers who are designing a study, and for clinicians who wish to interpret research findings. The elements that must be specified to calculate the sample size include alpha, beta, Type I and Type II errors, 1- and 2-tail tests, confidence intervals, and confidence levels. A computer software program written by one of the authors (MHE), Sample Size Expert, facilitates sample size calculations. The program uses an expert system to help inexperienced users calculate sample sizes for analytic and descriptive studies. The software is available at no cost from the author or electronically via several on-line information services.
Energy Technology Data Exchange (ETDEWEB)
Silvestre, Larissa Jacome Barros
2016-07-01
Nuclear accidents in the world led to the establishment of rigorous criteria and requirements for nuclear power plant operations by the international regulatory bodies. By using specific computer programs, simulations of various accidents and transients likely to occur at any nuclear power plant are required for certifying and licensing a nuclear power plant. Based on this scenario, some sophisticated computational tools have been used such as the Reactor Excursion and Leak Analysis Program (RELAP5), which is the most widely used code for the thermo-hydraulic analysis of accidents and transients in nuclear reactors in Brazil and worldwide. A major difficulty in the simulation by using RELAP5 code is the amount of information required for the simulation of thermal-hydraulic accidents or transients. The preparation of the input data requires a great number of mathematical operations to calculate the geometry of the components. Thus, for those calculations performance and preparation of RELAP5 input data, a friendly mathematical preprocessor was designed. The Visual Basic for Application (VBA) for Microsoft Excel demonstrated to be an effective tool to perform a number of tasks in the development of the program. In order to meet the needs of RELAP5 users, the RELAP5 Calculation Program (Programa de Calculo do RELAP5 - PCRELAP5) was designed. The components of the code were codified; all entry cards including the optional cards of each one have been programmed. In addition, an English version for PCRELAP5 was provided. Furthermore, a friendly design was developed in order to minimize the time of preparation of input data and errors committed by users. In this work, the final version of this preprocessor was successfully applied for Safety Injection System (SIS) of Angra 2. (author)
Temporal code in the vibrissal system-Part II: Roughness surface discrimination
Energy Technology Data Exchange (ETDEWEB)
Farfan, F D [Departamento de BioingenierIa, FACET, Universidad Nacional de Tucuman, INSIBIO - CONICET, CC 327, Postal Code CP 4000 (Argentina); AlbarracIn, A L [Catedra de Neurociencias, Facultad de Medicina, Universidad Nacional de Tucuman (Argentina); Felice, C J [Departamento de BioingenierIa, FACET, Universidad Nacional de Tucuman, INSIBIO - CONICET, CC 327, Postal Code CP 4000 (Argentina)
2007-11-15
Previous works have purposed hypotheses about the neural code of the tactile system in the rat. One of them is based on the physical characteristics of vibrissae, such as frequency of resonance; another is based on discharge patterns on the trigeminal ganglion. In this work, the purpose is to find a temporal code analyzing the afferent signals of two vibrissal nerves while vibrissae sweep surfaces of different roughness. Two levels of pressure were used between the vibrissa and the contact surface. We analyzed the afferent discharge of DELTA and GAMMA vibrissal nerves. The vibrissae movements were produced using electrical stimulation of the facial nerve. The afferent signals were analyzed using an event detection algorithm based on Continuous Wavelet Transform (CWT). The algorithm was able to detect events of different duration. The inter-event times detected were calculated for each situation and represented in box plot. This work allowed establishing the existence of a temporal code at peripheral level.
A calculational system to aid economical use of MTRs
Energy Technology Data Exchange (ETDEWEB)
Reitsma, F.; Joubert, W.R. [Radiation and Reactor Theory Department, Atomic Energy Corporation of South Africa Ltd., Pretoria (South Africa)
1999-07-01
The availability of a fast and accurate core neutronic calculational system is a valuable asset in the operation and utilization of research reactors. Its primary value lies in optimum reload design, fuel management, safety and utilization studies. In this paper the OSCAR-3 calculational system of the Atomic Energy Corporation of South Africa (AEC) is discussed in detail. The different components and important features are explained with a short summary of some comparisons with experiments. (author)
Photovoltaic Power Systems and the National Electrical Code: Suggested Practices
Energy Technology Data Exchange (ETDEWEB)
None
2002-02-01
This guide provides information on how the National Electrical Code (NEC) applies to photovoltaic systems. The guide is not intended to supplant or replace the NEC; it paraphrases the NEC where it pertains to photovoltaic systems and should be used with the full text of the NEC. Users of this guide should be thoroughly familiar with the NEC and know the engineering principles and hazards associated with electrical and photovoltaic power systems. The information in this guide is the best available at the time of publication and is believed to be technically accurate; it will be updated frequently.
Photovoltaic power systems and the National Electrical Code: Suggested practices
Energy Technology Data Exchange (ETDEWEB)
Wiles, J. [New Mexico State Univ., Las Cruces, NM (United States). Southwest Technology Development Inst.
1996-12-01
This guide provides information on how the National Electrical Code (NEC) applies to photovoltaic systems. The guide is not intended to supplant or replace the NEC; it paraphrases the NEC where it pertains to photovoltaic systems and should be used with the full text of the NEC. Users of this guide should be thoroughly familiar with the NEC and know the engineering principles and hazards associated with electrical and photovoltaic power systems. The information in this guide is the best available at the time of publication and is believed to be technically accurate; it will be updated frequently. Application of this information and results obtained are the responsibility of the user.
Channel estimation for physical layer network coding systems
Gao, Feifei; Wang, Gongpu
2014-01-01
This SpringerBrief presents channel estimation strategies for the physical later network coding (PLNC) systems. Along with a review of PLNC architectures, this brief examines new challenges brought by the special structure of bi-directional two-hop transmissions that are different from the traditional point-to-point systems and unidirectional relay systems. The authors discuss the channel estimation strategies over typical fading scenarios, including frequency flat fading, frequency selective fading and time selective fading, as well as future research directions. Chapters explore the performa
Cygnus Code Simulation of Magnetoshell Aerocapture and Entry System
Shimazu, Akihisa; Kirtley, David; Barnes, Dan; Slough, John
2017-10-01
A Magnetoshell Aerocapture and Entry System (MAC) is a novel concept for planetary atmospheric entry, which enables both manned and planetary deep space orbiter space missions that are difficult with present day technologies. The MAC uses a low-beta dipole plasma magnetoshell to produce a drag effect on the spacecraft through the collisional interactions between the entry atmospheric neutrals and the confined plasma in the magnetoshell, creating a dynamic and controllable plasma parachute for entry. To understand the performance and the behavior of the MAC, the Cygnus 2D Hall MHD code is used for this study. The Cygnus code is a 2D Hall MHD code with coupled external circuits, which has been originally developed for studying FRC formation, translation, merging, and compression. In this study, the Cygnus code is modified to support the MAC geometry with a simplified plasma-neutral model that accounts for electron-impact ionization, radiative recombination, and resonant charge exchange to simulate the collisional interaction processes for the MAC.
MC/sup 2/-2: a code to calculate fast neutron spectra and multigroup cross sections. [LMFBR
Energy Technology Data Exchange (ETDEWEB)
Henryson, H. II; Toppel, B. J.; Stenberg, C. G.
1976-06-01
MC/sup 2/-2 is a program to solve the neutron slowing down problem using basic neutron data derived from the ENDF/B data files. The spectrum calculated by MC/sup 2/-2 is used to collapse the basic data to multigroup cross sections for use in standard reactor neutronics codes. Four different slowing down formulations are used by MC/sup 2/-2: multigroup, continuous slowing down using the Goertzel-Greuling or Improved Goertzel-Greuling moderating parameters, and a hyper-fine-group integral transport calculation. Resolved and unresolved resonance cross sections are calculated accounting for self-shielding, broadening and overlap effects. This document provides a description of the MC/sup 2/-2 program. The physics and mathematics of the neutron slowing down problem are derived and detailed information is provided to aid the MC/sup 2/-2 user in preparing input for the program and implementation of the program on IBM 370 or CDC 7600 computers.
García-Jerez, Antonio; Sánchez-Sesma, Francisco J; Luzón, Francisco; Perton, Mathieu
2016-01-01
During a quarter of a century, the main characteristics of the horizontal-to-vertical spectral ratio of ambient noise HVSRN have been extensively used for site effect assessment. In spite of the uncertainties about the optimum theoretical model to describe these observations, several schemes for inversion of the full HVSRN curve for near surface surveying have been developed over the last decade. In this work, a computer code for forward calculation of H/V spectra based on the diffuse field assumption (DFA) is presented and tested.It takes advantage of the recently stated connection between the HVSRN and the elastodynamic Green's function which arises from the ambient noise interferometry theory. The algorithm allows for (1) a natural calculation of the Green's functions imaginary parts by using suitable contour integrals in the complex wavenumber plane, and (2) separate calculation of the contributions of Rayleigh, Love, P-SV and SH waves as well. The stability of the algorithm at high frequencies is preserv...
Validation of system codes RELAP5 and SPECTRA for natural convection boiling in narrow channels
Energy Technology Data Exchange (ETDEWEB)
Stempniewicz, M.M., E-mail: stempniewicz@nrg.eu; Slootman, M.L.F.; Wiersema, H.T.
2016-10-15
Highlights: • Computer codes RELAP5/Mod3.3 and SPECTRA 3.61 validated for boiling in narrow channels. • Validated codes can be used for LOCA analyses in research reactors. • Code validation based on natural convection boiling in narrow channels experiments. - Abstract: Safety analyses of LOCA scenarios in nuclear power plants are performed with so called thermal–hydraulic system codes, such as RELAP5. Such codes are validated for typical fuel geometries applied in nuclear power plants. The question considered by this article is if the codes can be applied for LOCA analyses in research reactors, in particular exceeding CHF in very narrow channels. In order to answer this question, validation calculations were performed with two thermal–hydraulic system codes: RELAP and SPECTRA. The validation was based on natural convection boiling in narrow channels experiments, performed by Prof. Monde et al. in the years 1990–2000. In total 42 vertical tube and annulus experiments were simulated with both codes. A good agreement of the calculated values with the measured data was observed. The main conclusions are: • The computer codes RELAP5/Mod 3.3 (US NRC version) and SPECTRA 3.61 have been validated for natural convection boiling in narrow channels using experiments of Monde. The dimensions applied in the experiments were performed for a range that covers the values observed in typical research reactors. Therefore it is concluded that both codes are validated and can be used for LOCA analyses in research reactors, including natural convection boiling. The applicability range of the present validation is: hydraulic diameters of 1.1 ⩽ D{sub hyd} ⩽ 9.0 mm, heated lengths of 0.1 ⩽ L ⩽ 1.0 m, pressures of 0.10 ⩽ P ⩽ 0.99 MPa. In most calculations the burnout was predicted to occur at lower power than that observed in the experiments. In several cases the burnout was observed at higher power. The overprediction was not larger than 16% in RELAP and 15% in
Status of the LAHET{trademark} Code System
Energy Technology Data Exchange (ETDEWEB)
Waters, L.S.; Prael, R.E.
1995-12-31
The LAHET Code System (LCS) is extensively used for medium energy accelerator applications, including spallation target design and deep penetration shielding problems. Current applications include Accelerator Production of Tritium (APT), Accelerator Driven Transmutation Technologies (ADTT), LANSCE and WNR spallation target upgrades, as well as various medical projects. We will discuss recent upgrades to the MCNP and LAHET components of LCS, AND review the work in progress now funded under the APT program.
Coding system of therapeutic focus on action and insight.
Samoilov, A; Goldfried, M R; Shapiro, D A
2000-06-01
This study (a) used an established comprehensive process measure to uncover a latent pattern of therapeutic focus in cognitive-behavioral and psychodynamic-interpersonal sessions; (b) used these results to develop the coding system of Therapeutic Focus on Action and Insight, which makes it possible to evaluate therapists' relative emphasis on the Constructing Meaning and Facilitating Action domains of in-session focus; and (c) evaluated its reliability and validity.
Energy Technology Data Exchange (ETDEWEB)
Young, P.G.; Arthur, E.D.
1977-11-01
A new multistep Hauser--Feshbach code that includes corrections for preequilibrium effects is described. The code can calculate up to 60 decay reactions (cross sections and energy spectra) in one computation, and thereby provide considerable flexibility for handling processes with complicated reaction chains. Input parameter setup, problem output, and subroutine descriptions are given along with a sample problem calculation. A brief theoretical description is also included. 8 figures, 3 tables.
Recriticality calculation with GENFLO code for the BWR core after steam explosion in the lower head
Energy Technology Data Exchange (ETDEWEB)
Miettinen, J. [VTT Processes (Finland)
2002-12-01
Recriticality of the partially degraded BWR core has been studied by assuming a severe accident phase during which the fuel rods are still intact but the control rods have experienced extensive damage. Previous NKS and EU projects have studied the same case assuming reflooding by the ECCS system In the present study it was assumed that coolant enters the core due to melt-coolant interaction in the lower plenum. In the first case specified the relocation and fragmentation of the molten control rod metal causes the level swell in the core but no steam explosion. In the second case a steam explosion in the lower head was assumed. I n the first case a prompt recriticality peak can occur, but after the peak no semistable power generation remains. In the second case the consequence of the slug entrance into the core is so violent that the fuel disintegration and melting during the first power peak may occur. After the large power peak water is rapidly pushed back from the core and no semistable power generation maintains. The fuel disintegration studies have been based on a coarse assumption that the acceptable local energy addition into the fresh fuel may be 170 cal/g, but with increasing burn-up it can be as low as 60-70 cal/g. In the level swell variations the maximum energy addition was between these limits, but in most of the steam explosion variations much above these limits. Additional variation of the assumptions related to the neutronics demonstrated that for the converged analysis result some interactions would be useful with respect to the boundary conditions and neutronic options.
Investigation of NACOK air ingress experiment using different system analysis codes
Energy Technology Data Exchange (ETDEWEB)
Zheng Yanhua, E-mail: zhengyh@tsinghua.edu.cn [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Stempniewicz, Marek M. [NRG Arnhem, Utrechtseweg 310, P.O. Box 9034, 6800 ES Arnhem (Netherlands)
2012-10-15
Air ingress into to the core after the primary circuit depressurization due to large breaks of the pressure boundary is considered as one of the severe hypothetical accidents for the high temperature gas-cooled reactor (HTR). If the air source and the natural convection cannot be impeded, the continuous graphite oxidation reaction along with the formation of burnable gas mixtures resulting in the corrosion of the fuel elements and the reflectors might damage the reactor structure integrity and endanger the reactor safety. In order to study the effects of air flow driven by natural convection as well as to investigate the corrosion of graphite, the NACOK (Naturzug im Core mit Korrosion) facility was built at Juelich Research Center in Germany. A complete 2A-rupture of the coaxial duct in the HTR primary system, as well as the chimney effect caused by breaks in both upper and lower parts of the pressure boundary was simulated in the test facility. Several series of experiments and the related code validations (TINTE, DIREKT, THERMIX/REACT, etc.) have been performed on this facility since the 1990s. In this paper, the latest NACOK air ingress experiment, carried out on October 23, 2008 to simulate the chimney effect, was preliminarily analyzed at NRG with the SPECTRA code, as well as at INET, Tsinghua University of China with the TINTE code. The calculating results of air flow rate of natural convection, time-dependent graphite corrosion, and temperature distribution are compared with the NACOK test results. The preliminary code-to-experiment and code-to-code validation successfully proves the code capability to simulate and predict the air-ingress accident. In addition, more research work, including parameter sensitivity analysis, modeling refinement, code amelioration, etc., should be performed to improve the simulation accuracy in the future.
Eisenbach, Markus
The Locally Self-consistent Multiple Scattering (LSMS) code solves the first principles Density Functional theory Kohn-Sham equation for a wide range of materials with a special focus on metals, alloys and metallic nano-structures. It has traditionally exhibited near perfect scalability on massively parallel high performance computer architectures. We present our efforts to exploit GPUs to accelerate the LSMS code to enable first principles calculations of O(100,000) atoms and statistical physics sampling of finite temperature properties. Using the Cray XK7 system Titan at the Oak Ridge Leadership Computing Facility we achieve a sustained performance of 14.5PFlop/s and a speedup of 8.6 compared to the CPU only code. This work has been sponsored by the U.S. Department of Energy, Office of Science, Basic Energy Sciences, Material Sciences and Engineering Division and by the Office of Advanced Scientific Computing. This work used resources of the Oak Ridge Leadership Computing Facility, which is supported by the Office of Science of the U.S. Department of Energy under Contract No. DE-AC05-00OR22725.
Energy Technology Data Exchange (ETDEWEB)
Botta, F.; Mairani, A.; Battistoni, G.; Cremonesi, M.; Di Dia, A.; Fasso, A.; Ferrari, A.; Ferrari, M.; Paganelli, G.; Pedroli, G.; Valente, M. [Medical Physics Department, European Institute of Oncology, Via Ripamonti 435, 20141 Milan (Italy); Istituto Nazionale di Fisica Nucleare (I.N.F.N.), Via Celoria 16, 20133 Milan (Italy); Medical Physics Department, European Institute of Oncology, Via Ripamonti 435, 20141 Milan (Italy); Jefferson Lab, 12000 Jefferson Avenue, Newport News, Virginia 23606 (United States); CERN, 1211 Geneva 23 (Switzerland); Medical Physics Department, European Institute of Oncology, Milan (Italy); Nuclear Medicine Department, European Institute of Oncology, Via Ripamonti 435, 2014 Milan (Italy); Medical Physics Department, European Institute of Oncology, Via Ripamonti 435, 20141 Milan (Italy); FaMAF, Universidad Nacional de Cordoba and CONICET, Cordoba, Argentina C.P. 5000 (Argentina)
2011-07-15
Purpose: The calculation of patient-specific dose distribution can be achieved by Monte Carlo simulations or by analytical methods. In this study, fluka Monte Carlo code has been considered for use in nuclear medicine dosimetry. Up to now, fluka has mainly been dedicated to other fields, namely high energy physics, radiation protection, and hadrontherapy. When first employing a Monte Carlo code for nuclear medicine dosimetry, its results concerning electron transport at energies typical of nuclear medicine applications need to be verified. This is commonly achieved by means of calculation of a representative parameter and comparison with reference data. Dose point kernel (DPK), quantifying the energy deposition all around a point isotropic source, is often the one. Methods: fluka DPKs have been calculated in both water and compact bone for monoenergetic electrons (10{sup -3} MeV) and for beta emitting isotopes commonly used for therapy ({sup 89}Sr, {sup 90}Y, {sup 131}I, {sup 153}Sm, {sup 177}Lu, {sup 186}Re, and {sup 188}Re). Point isotropic sources have been simulated at the center of a water (bone) sphere, and deposed energy has been tallied in concentric shells. fluka outcomes have been compared to penelope v.2008 results, calculated in this study as well. Moreover, in case of monoenergetic electrons in water, comparison with the data from the literature (etran, geant4, mcnpx) has been done. Maximum percentage differences within 0.8{center_dot}R{sub CSDA} and 0.9{center_dot}R{sub CSDA} for monoenergetic electrons (R{sub CSDA} being the continuous slowing down approximation range) and within 0.8{center_dot}X{sub 90} and 0.9{center_dot}X{sub 90} for isotopes (X{sub 90} being the radius of the sphere in which 90% of the emitted energy is absorbed) have been computed, together with the average percentage difference within 0.9{center_dot}R{sub CSDA} and 0.9{center_dot}X{sub 90} for electrons and isotopes, respectively. Results: Concerning monoenergetic electrons
Botta, F; Mairani, A; Battistoni, G; Cremonesi, M; Di Dia, A; Fassò, A; Ferrari, A; Ferrari, M; Paganelli, G; Pedroli, G; Valente, M
2011-07-01
The calculation of patient-specific dose distribution can be achieved by Monte Carlo simulations or by analytical methods. In this study, FLUKA Monte Carlo code has been considered for use in nuclear medicine dosimetry. Up to now, FLUKA has mainly been dedicated to other fields, namely high energy physics, radiation protection, and hadrontherapy. When first employing a Monte Carlo code for nuclear medicine dosimetry, its results concerning electron transport at energies typical of nuclear medicine applications need to be verified. This is commonly achieved by means of calculation of a representative parameter and comparison with reference data. Dose point kernel (DPK), quantifying the energy deposition all around a point isotropic source, is often the one. FLUKA DPKS have been calculated in both water and compact bone for monoenergetic electrons (10-3 MeV) and for beta emitting isotopes commonly used for therapy (89Sr, 90Y, 131I 153Sm, 177Lu, 186Re, and 188Re). Point isotropic sources have been simulated at the center of a water (bone) sphere, and deposed energy has been tallied in concentric shells. FLUKA outcomes have been compared to PENELOPE v.2008 results, calculated in this study as well. Moreover, in case of monoenergetic electrons in water, comparison with the data from the literature (ETRAN, GEANT4, MCNPX) has been done. Maximum percentage differences within 0.8.RCSDA and 0.9.RCSDA for monoenergetic electrons (RCSDA being the continuous slowing down approximation range) and within 0.8.X90 and 0.9.X90 for isotopes (X90 being the radius of the sphere in which 90% of the emitted energy is absorbed) have been computed, together with the average percentage difference within 0.9.RCSDA and 0.9.X90 for electrons and isotopes, respectively. Concerning monoenergetic electrons, within 0.8.RCSDA (where 90%-97% of the particle energy is deposed), FLUKA and PENELOPE agree mostly within 7%, except for 10 and 20 keV electrons (12% in water, 8.3% in bone). The
Development and evaluation of a Monte Carlo Code System for analysis of ionization chamber responses
Energy Technology Data Exchange (ETDEWEB)
Johnson, J.O.; Gabriel, T.A.
1987-07-01
This document reports on the development and testing of a Monte Carlo code system for calculating the response of an ionization chamber to mixed neutron and photon radiation environment. The code system, Monte Carlo Ionization Chamber Analysis Package (MICAP), determines the neutron, photon, and total responses of the ionization chamber to the mixed field radiation environment. The Monte Carlo method performs accurate simulations of the physical processes involved in detecting radiation using ionization chambers, and eliminates limitations inherent in approximate methods. To evaluate MICAP, comparisons were made with results obtained using other code systems and with experimental results. Separate comparisons with other code systems verified the validity of the neutron, photon, and charged particle transport processes and the nuclear models used to describe the individual neutron reactions, respectively. Comparisons with mono-energetic photon calibration experiments and with mixed neutron and photon radiation experiments verified the applicability of MICAP for analyzing the response of ionization chambers to mixed field radiation environments. 47 refs., 23 figs., 20 tabs.
Jabbari, Keivan
A fast and accurate treatment planning system is essential for radiation therapy and Monte Carlo (MC) techniques produce the most accurate results for dose calculation in treatment planning. In this work, we developed a fast Monte Carlo code based on pre-calculated data (PMC, Pre-calculated Monte Carlo) for applications in radiation therapy treatment planning. The PMC code takes advantage of large available memory in current computer hardware for extensive generation of pre-calculated data. Primary tracks of electrons are generated in the middle of homogeneous materials (water, air, bone, lung) and with energies between 0.2 and 18 MeV using the EGSnrc code. Secondary electrons are not transported but their position, energy, charge and direction are saved and used as a primary particle. Based on medium type and incident electron energy, a track is selected from the pre-calculated set. The performance of the method is tested in various homogeneous and heterogeneous configurations and the results were generally within 2% compared to EGSnrc but with a 40-60 times speed improvement. The limitations of various techniques for the improvement of speed and accuracy of particle transport have been evaluated. We studied the obstacles for further increased speed ups in voxel based geometries by including ray-tracing and particle fluence information in the pre-generated track information. The latter method leads to speed-increases of about a factor of 500 over EGSnrc for voxel-based geometries. In both approaches, no physical calculation is carried out during the runtime phase after the pre-generated data has been stored even in the presence of heterogeneities. The pre-calculated data is generated for each particular material and this improves the performance of the pre-calculated Monte Carlo code both in terms of accuracy and speed. The PMC is also extended for proton transport in radiation therapy. The pre-calculated data is based on tracks of 1000 primary protons using
Evolution of the CYCLE code for the system analysis of the nuclear fuel cycle
Directory of Open Access Journals (Sweden)
A.G. Kalashnikov
2016-06-01
Full Text Available The CYCLE code is intended to simulate mathematically the operation of a nuclear power system (NPS with thermal and fast reactors in an open or closed nuclear fuel cycle, to develop scenarios of efficient nuclear power evolution in Russia and to analyze trends in global nuclear power. The code is based on a well-known software program, WIMSD-5B, broadly used for the design of thermal reactor cells, and on a 2D multi-group software system, RZA, for the fast neutron reactor simulation. The CYCLE code was developed at IPPE in Obninsk. This paper presents a brief review of the capabilities and information on the current status of the CYCLE code. The code allows simulation of key facilities of the external fuel cycle (fuel fabrication and reprocessing facilities, SNF storage, uranium, plutonium, neptunium, americium and curium stores, RW long-term storage sites, nuclear reactors, including RBMK-1000 reactors, existing and advanced VVER reactors (using different fuel types, and fast reactors (both existing and innovative. As an important feature, the CYCLE code allows the evolution of the fuel's nuclide composition both in reactors and at the external fuel cycle phase to be considered in details. Offered as an extra option is the capability to calculate a variety of the nuclear fuel cycle cost parameters for nuclear power plants with thermal and fast reactors. For years, the code has been successfully used as part of INPRO, an international innovative nuclear reactor and fuel cycle project. The results of studies into the Russian NPS evolution scenarios were presented at Global 2011. Some other of the CYCLE-based simulation results were presented at Global 2015.
Energy Technology Data Exchange (ETDEWEB)
NONE
1997-03-01
This Manual represents Revision 5 of the user documentation for the modular code system referred to as SCALE. The history of the SCALE code system dates back to 1969 when the current Computational Physics and Engineering Division at Oak Ridge National Laboratory (ORNL) began providing the transportation package certification staff at the U.S. Atomic Energy Commission with computational support in the use of the new KENO code for performing criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976 the certification staff relied on the ORNL staff to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analyses of transportation packages. However, the certification staff learned that, with only occasional use of the codes, it was difficult to become proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the move of the certification staff to the U.S. Nuclear Regulatory Commission (NRC), the NRC staff proposed the development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system was born. This volume consists of the section of the manual dealing with three of the functional modules in the code. Those are the Morse-SGC for the SCALE system, Heating 7.2, and KENO V.a. The manual describes the latest released versions of the codes.
Energy Technology Data Exchange (ETDEWEB)
Keney, G.S.
1981-08-01
A computer code has been written to calculate neutron induced activation of neutral-beam injector components and the corresponding dose rates as a function of geometry, component composition, and time after shutdown. The code, ACDOS1, was written in FORTRAN IV to calculate both activity and dose rates for up to 30 target nuclides and 50 neutron groups. Sufficient versatility has also been incorporated into the code to make it applicable to a variety of general activation problems due to neutrons of energy less than 20 MeV.
Security Concerns and Countermeasures in Network Coding Based Communications Systems
DEFF Research Database (Denmark)
Talooki, Vahid; Bassoli, Riccardo; Roetter, Daniel Enrique Lucani
2015-01-01
This survey paper shows the state of the art in security mechanisms, where a deep review of the current research and the status of this topic is carried out. We start by introducing network coding and its variety applications in enhancing current traditional networks. In particular, we analyze two...... key protocol types, namely, state-aware and stateless protocols, specifying the benefits and disadvantages of each one of them. We also present the key security assumptions of network coding (NC) systems as well as a detailed analysis of the security goals and threats, both passive and active....... This paper also presents a detailed taxonomy and a timeline of the different NC security mechanisms and schemes reported in the literature. Current proposed security mechanisms and schemes for NC in the literature are classified later. Finally a timeline of these mechanism and schemes is presented....
[Behavior ethogram and PAE coding system of Cervus nippon sichuanicus].
Qi, Wen-Hua; Yue, Bi-Song; Ning, Ji-Zu; Jiang, Xue-Mei; Quan, Qiu-Mei; Guo, Yan-Shu; Mi, Jun; Zuo, Lin; Xiong, Yuan-Qing
2010-02-01
A monthly 5-day periodic observation at 06:00-18:00 from March to November 2007 was conducted to record the behavioral processes, contents, and results, and the surrounding habitats of Sichuan sika deer (Cervus nippon sichuanicus) in Donglie, Chonger, and Reer villages of Tiebu Natural Reserve of Sichuan Province. The behavioral ethogram, vigilance behaviors ethogram and its PAE (posture, act, and environment) coding system of the Sichuan sika deer were established, which filled the gap of the PAE coding of ungulates vigilance behaviors. A total of 11 kinds of postures, 83 acts, and 136 behaviors were recorded and distinguished, with the relative frequency of each behavior in relation to gender, age, and season described. Compared with other ungulates, the behavioral repertoire of Sichuan sika deer was mostly similar to that of other cervid animals.
The penelope code system. Specific features and recent improvements
Salvat, Francesc
2014-06-01
Since its first release, back in 1996, the Monte Carlo code system penelope has evolved into a flexible and reliable tool for describing coupled electron-photon transport in complex material structures. The present article contains an overview of the physical interaction models, particle tracking methods, geometry tools, and variance-reduction techniques implemented in penelope. Recent refinements aimed at improving the accuracy of the code, and its stability under variations of user-defined simulation parameters, are also described. These include the use of reliable cross sections for the ionization of inner atomic electron shells by electron/positron impact, a reformulation of the random-hinge method, and the use of fuzzy quadric surfaces in the description of the geometry.
Method of laser beam coding for control systems
Pałys, Tomasz; Arciuch, Artur; Walczak, Andrzej; Murawski, Krzysztof
2017-08-01
The article presents the method of encoding a laser beam for control systems. The experiments were performed using a red laser emitting source with a wavelength of λ = 650 nm and a power of P ≍ 3 mW. The aim of the study was to develop methods of modulation and demodulation of the laser beam. Results of research, in which we determined the effect of selected camera parameters, such as image resolution, number of frames per second on the result of demodulation of optical signal, is also shown in the paper. The experiments showed that the adopted coding method provides sufficient information encoded in a single laser beam (36 codes with the effectiveness of decoding at 99.9%).
Energy Technology Data Exchange (ETDEWEB)
Gil, Soo Lee; Nam, Zin Cho [Korea Advanced Institute of Science and Technology, Dept. of Nuclear and Quantum Engineering, Yusong-gu, Daejeon (Korea, Republic of)
2005-07-01
The OECD 3-dimensional benchmark problem C5G7 MOX was calculated by the CRX code. For 3-dimensional heterogeneous calculation, the CRX code uses a fusion technique of 2-dimensional/1-dimensional methods: the method of characteristics for radial 2-dimensional calculation and diamond difference scheme (DD) that is an S{sub N}-like method for axial 1-dimensional calculation. We improve the fusion method by using a linear characteristics (LC) solver in the 1-dimensional calculation. Here, we present brief structure of 2-dimensional/1-dimensional fusion method and the results of 3 configurations of benchmark problem. We also present results of several different 1-dimensional calculation options. Numerical results show that the LC scheme presents better performance than DD. In the results of the benchmark problem, k(eff) errors are less than 0.05% and the averages of pin power errors are less than 1% for all calculations.
Koeman, Tom; Offermans, Nadine S M; Christopher-de Vries, Yvette; Slottje, Pauline; Van Den Brandt, Piet A; Goldbohm, R Alexandra; Kromhout, Hans; Vermeulen, Roel
2013-01-01
In epidemiological studies, occupational exposure estimates are often assigned through linkage of job histories to job-exposure matrices (JEMs). However, available JEMs may have a coding system incompatible with the coding system used to code the job histories, necessitating a translation of the originally assigned job codes. Since manual recoding is usually not feasible in large studies, this is often done by use of automated crosswalks translating job codes from one system to another. We set out to investigate whether automatically translating job codes led to different exposure estimates compared with those resulting from manual recoding using the original job descriptions. One hundred job histories were randomly drawn from the Netherlands Cohort Study on diet and cancer (NLCS), using a sampling strategy designed to oversample potentially exposed jobs. This resulted in 220 job codes that were automatically translated from the original Dutch coding system to the International Standard Classification of Occupations (ISCO)-68 and ISCO-88 as well as manually recoded from the job descriptions in the original questionnaire by two coders. Exposure to several agents (i.e. chromium, asbestos, silica, pesticides, aromatic solvents, and extremely low-frequency magnetic fields) was assigned by JEMs based on job codes resulting from automatic and manual recodings. The agreement between occupational exposure estimates based on the crosswalk versus those based on manual recoding reached a Cohen's Kappa (κ) of 0.66 or higher and were similar to the agreements between the two coders. Results of this study indicate that using automated crosswalks to recode job codes from one occupational classification system to another results only in a limited loss in agreement in assigned occupational exposure estimates compared with direct manual recoding. Therefore, in this case, crosswalks provide an efficient alternative to the costly and time-consuming direct manual recoding from job
Forming the Calculated Dynamic Transmission Systems of Wheeled Vehicles
Directory of Open Access Journals (Sweden)
A. B. Fominykh
2017-01-01
Full Text Available To calculate dynamic loading of transmission parts of wheeled vehicles, it is necessary to build up the appropriate calculated dynamic systems and determine their inertial, elastic, and damping parameters.The initial point of this process is to form an initial dynamic system. Hereafter, to cut the time of computations there is a need to reduce the number of masses of this system, and sometimes simplify its structure. The main requirement to be fulfilled in this case is that the calculated dynamical system is to be equivalent to the initial one (in terms of similarity of the vibrational process characteristics in these systems, i.e., the frequencies and modes of oscillations of both systems, their amplitude-frequency characteristics. This is possible when the energy characteristics of the corresponding systems are equal, i.e. their kinetic and potential energies, dissipative functions, and external force energies.Usually, when forming the initial and calculated dynamic systems, all types of friction are reduced to a linearly viscous one. However, it disables us to investigate the motion of these systems if there is an arbitrary, in particular, poly-harmonic action (for example, on the side of the internal combustion engine, since in this case the linear friction coefficients given will depend on the frequency and amplitude of the oscillations.The paper is aimed at determining the equivalent parameters of calculated dynamic systems of wheeled vehicles, including the dissipative parameters for the general case of friction.On the basis of energy principles, the expressions are obtained to determine the equivalent inertial, elastic, and damping parameters of the calculated dynamical systems of wheeled vehicles when the structure is changed and the number of masses of the system is decreased. The presented technique enables us to investigate the motion of these systems under arbitrary, including poly-harmonic, action on the system, using the
Trip and Control System Models in SPACE Code
Energy Technology Data Exchange (ETDEWEB)
Lee, Eun Ju; Park, Chan Eok; Lee, Gyu Cheon [Korea Power Engineering Company, Daejeon (Korea, Republic of)
2009-05-15
KOPEC has been developing a hydraulic solver of SPACE, which is a nuclear power plant safety analysis code, using two-fluid, three-field governing equations. Several numerical schemes, such as collocated, staggered, semi-implicit, and implicit schemes, have been tried so far. In this paper, the trip and control system model of SPACE will be described. The trip system of SPACE is developed to evaluate logical statements. Each trip statement is a simple logical statement that has a true or false result and an associated variable. The control system provides the capability to evaluate simultaneous algebraic and ordinary differential equations. The capability is primarily intended to simulate control systems typically used in nuclear reactor systems, but it can also model other phenomena described by algebraic and ordinary differential equations.
Energy Technology Data Exchange (ETDEWEB)
NONE
1997-03-01
This Manual represents Revision 5 of the user documentation for the modular code system referred to as SCALE. The history of the SCALE code system dates back to 1969 when the current Computational Physics and Engineering Division at Oak Ridge National Laboratory (ORNL) began providing the transportation package certification staff at the U.S. Atomic Energy Commission with computational support in the use of the new KENO code for performing criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976 the certification staff relied on the ORNL staff to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analyses of transportation packages. However, the certification staff learned that, with only occasional use of the codes, it was difficult to become proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the move of the certification staff to the U.S. Nuclear Regulatory Commission (NRC), the NRC staff proposed the development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system was born. This volume consists of the section of the manual dealing with eight of the functional modules in the code. Those are: BONAMI - resonance self-shielding by the Bondarenko method; NITAWL-II - SCALE system module for performing resonance shielding and working library production; XSDRNPM - a one-dimensional discrete-ordinates code for transport analysis; XSDOSE - a module for calculating fluxes and dose rates at points outside a shield; KENO IV/S - an improved monte carlo criticality program; COUPLE; ORIGEN-S - SCALE system module to calculate fuel depletion, actinide transmutation, fission product buildup and decay, and associated radiation source terms; ICE.
The ICPC coding system in pharmacy : developing a subset, ICPC-Ph
van Mil, JWF; Brenninkmeijer, R; Tromp, TFJ
The ICPC system is a coding system developed for general medical practice, to be able to code the GP-patient encounters and other actions. Some of the codes can be easily used by community pharmacists to code complaints and diseases in pharmaceutical care practice. We developed a subset of the ICPC
Energy Technology Data Exchange (ETDEWEB)
Romero, L.; Travesi, A.
1983-07-01
A codes, BETAL, was developed, written in FORTRAN IV, to automatize calculations and presentations of the result of the total alpha-beta activities measurements in environmental samples. This code performs the necessary calculations for transformation the activities measured in total counts, to pCi/1., bearing in mind the efficiency of the detector used and the other necessary parameters. Further more, it appraise the standard deviation of the result, and calculus the Lower limit of detection for each measurement. This code is written in iterative way by screen-operator dialogue, and asking the necessary data to perform the calculation of the activity in each case by a screen label. The code could be executed through any screen and keyboard terminal, (whose computer accepts Fortran IV) with a printer connected to the said computer. (Author) 5 refs.
Advanced Error-Control Coding Methods Enhance Reliability of Transmission and Storage Data Systems
Directory of Open Access Journals (Sweden)
K. Vlcek
2003-04-01
Full Text Available Iterative coding systems are currently being proposed and acceptedfor many future systems as next generation wireless transmission andstorage systems. The text gives an overview of the state of the art initerative decoded FEC (Forward Error-Correction error-control systems.Such systems can typically achieve capacity to within a fraction of adB at unprecedented low complexities. Using a single code requires verylong code words, and consequently very complex coding system. One wayaround the problem of achieving very low error probabilities is turbocoding (TC application. A general model of concatenated coding systemis shown - an algorithm of turbo codes is given in this paper.
Error correcting coding-theory for structured light illumination systems
Porras-Aguilar, Rosario; Falaggis, Konstantinos; Ramos-Garcia, Ruben
2017-06-01
Intensity discrete structured light illumination systems project a series of projection patterns for the estimation of the absolute fringe order using only the temporal grey-level sequence at each pixel. This work proposes the use of error-correcting codes for pixel-wise correction of measurement errors. The use of an error correcting code is advantageous in many ways: it allows reducing the effect of random intensity noise, it corrects outliners near the border of the fringe commonly present when using intensity discrete patterns, and it provides a robustness in case of severe measurement errors (even for burst errors where whole frames are lost). The latter aspect is particular interesting in environments with varying ambient light as well as in critical safety applications as e.g. monitoring of deformations of components in nuclear power plants, where a high reliability is ensured even in case of short measurement disruptions. A special form of burst errors is the so-called salt and pepper noise, which can largely be removed with error correcting codes using only the information of a given pixel. The performance of this technique is evaluated using both simulations and experiments.
An intelligent management system for corporate information- calculating network development
Pirogov, V V; Khristenko, D V
2001-01-01
An approach to solving the problem of managerial control quality improvement corporate information-calculating networks (CICN) is proposed. The approach is based on the concept of a flexible toolkit. A macro-model, system-science and system-engineering models of an intelligent managerial control system (IMCS) for CICN development are considered as well as its dynamics. The practical opportunity of IMCS implementation is assessed. (5 refs).
TCODE: a computer code for analysis of tritium and vacuum systems for tokamak fusion reactors
Energy Technology Data Exchange (ETDEWEB)
Clemmer, R.G.
1978-08-01
TCODE can be used for either near-term experimental reactors or for commercial reactors. The code provides options for items that may be included in a commercial reactor such as a divertor, neutral beam heating, and a breeding blanket. The code was used to calculate tritium and vacuum system parameters for the near term reactors ITR, TNS-UP and EPR as well as for some commercial reactor designs, the UWMAK series. A selected sample of the tritium and vacuum parameters for these reactor designs is shown. Also shown are parameters for a hypothetical reactor UWMAK-III M having similar characteristics to UWMAK-III but with a higher fractional burnup (5.0% cf. 0.83%). The impact of the reactor design scenario upon major tritium and vacuum systems is discussed.
ATHENA AIDE: An expert system for ATHENA code input model preparation
Energy Technology Data Exchange (ETDEWEB)
Fink, R.K.; Callow, R.A.; Larson, T.K.; Ransom, V.H.
1987-01-01
An expert system called the ATHENA AIDE that assists in the preparation of input models for the ATHENA thermal-hydraulics code has been developed by researchers at the Idaho National Engineering Laboratory. The ATHENA AIDE uses a menu driven graphics interface and rule-based and object-oriented programming techniques to assist users of the ATHENA code in performing the tasks involved in preparing the card image input files required to run ATHENA calculations. The ATHENA AIDE was developed and currently runs on single-user Xerox artificial intelligence workstations. Experience has shown that the intelligent modeling environment provided by the ATHENA AIDE expert system helps ease the modeling task by relieving the analyst of many mundane, repetitive, and error prone procedures involved in the construction of an input model. This reduces errors in the resulting models, helps promote standardized modeling practices, and allows models to be constructed more quickly than was previously possible. 5 refs., 4 figs.
EquiFACS: The Equine Facial Action Coding System.
Directory of Open Access Journals (Sweden)
Jen Wathan
Full Text Available Although previous studies of horses have investigated their facial expressions in specific contexts, e.g. pain, until now there has been no methodology available that documents all the possible facial movements of the horse and provides a way to record all potential facial configurations. This is essential for an objective description of horse facial expressions across a range of contexts that reflect different emotional states. Facial Action Coding Systems (FACS provide a systematic methodology of identifying and coding facial expressions on the basis of underlying facial musculature and muscle movement. FACS are anatomically based and document all possible facial movements rather than a configuration of movements associated with a particular situation. Consequently, FACS can be applied as a tool for a wide range of research questions. We developed FACS for the domestic horse (Equus caballus through anatomical investigation of the underlying musculature and subsequent analysis of naturally occurring behaviour captured on high quality video. Discrete facial movements were identified and described in terms of the underlying muscle contractions, in correspondence with previous FACS systems. The reliability of others to be able to learn this system (EquiFACS and consistently code behavioural sequences was high--and this included people with no previous experience of horses. A wide range of facial movements were identified, including many that are also seen in primates and other domestic animals (dogs and cats. EquiFACS provides a method that can now be used to document the facial movements associated with different social contexts and thus to address questions relevant to understanding social cognition and comparative psychology, as well as informing current veterinary and animal welfare practices.
Electronic health record standards, coding systems, frameworks, and infrastructures
Sinha, Pradeep K; Bendale, Prashant; Mantri, Manisha; Dande, Atreya
2013-01-01
Discover How Electronic Health Records Are Built to Drive the Next Generation of Healthcare Delivery The increased role of IT in the healthcare sector has led to the coining of a new phrase ""health informatics,"" which deals with the use of IT for better healthcare services. Health informatics applications often involve maintaining the health records of individuals, in digital form, which is referred to as an Electronic Health Record (EHR). Building and implementing an EHR infrastructure requires an understanding of healthcare standards, coding systems, and frameworks. This book provides an
Fracchiolla, F.; Lorentini, S.; Widesott, L.; Schwarz, M.
2015-11-01
We propose a method of creating and validating a Monte Carlo (MC) model of a proton Pencil Beam Scanning (PBS) machine using only commissioning measurements and avoiding the nozzle modeling. Measurements with a scintillating screen coupled with a CCD camera, ionization chamber and a Faraday Cup were used to model the beam in TOPAS without using any machine parameter information but the virtual source distance from the isocenter. Then the model was validated on simple Spread Out Bragg Peaks (SOBP) delivered in water phantom and with six realistic clinical plans (many involving 3 or more fields) on an anthropomorphic phantom. In particular the behavior of the moveable Range Shifter (RS) feature was investigated and its modeling has been proposed. The gamma analysis (3%,3 mm) was used to compare MC, TPS (XiO-ELEKTA) and measured 2D dose distributions (using radiochromic film). The MC modeling proposed here shows good results in the validation phase, both for simple irradiation geometry (SOBP in water) and for modulated treatment fields (on anthropomorphic phantoms). In particular head lesions were investigated and both MC and TPS data were compared with measurements. Treatment plans with no RS always showed a very good agreement with both of them (γ -Passing Rate (PR) > 95%). Treatment plans in which the RS was needed were also tested and validated. For these treatment plans MC results showed better agreement with measurements (γ -PR > 93%) than the one coming from TPS (γ -PR < 88%). This work shows how to simplify the MC modeling of a PBS machine for proton therapy treatments without accounting for any hardware components and proposes a more reliable RS modeling than the one implemented in our TPS. The validation process has shown how this code is a valid candidate for a completely independent treatment plan dose calculation algorithm. This makes the code an important future tool for the patient specific QA verification process.
Dioguardi, Fabio; Dellino, Pierfrancesco
2014-05-01
PYFLOW is a computer code designed for quantifying the hazard related to Dilute Pyroclastic Density Currents (DPDC). DPDCs are multiphase flows that form during explosive volcanic eruptions. They are the major source of hazard related to volcanic eruptions, as they exert a significant stress over buildings and transport significant amounts of volcanic ash, which is hot and unbreathable. The program calculates the DPDC's impact parameters (e.g. dynamic pressure and particle volumetric concentration) and is founded on the turbulent boundary layer theory adapted to a multiphase framework. Fluid-dynamic variables are searched with a probabilistic approach, meaning that for each variable the average, maximum and minimum solutions are calculated. From these values, PYFLOW creates probability functions that allow to calculate the parameter at a given percentile. The code is written in Fortran 90 and can be compiled and installed on Windows, Mac OS X, Linux operating systems (OS). A User's manual is provided, explaining the details of the theoretical background, the setup and running procedure and the input data. The model inputs are DPDC deposits data, e.g. particle grainsize, layer thickness, particles shape factor and density. PYFLOW reads input data from a specifically designed input file or from the user's direct typing by command lines. Guidelines for writing input data are also contained in the package. PYFLOW guides the user at each step of execution, asking for additional data and inputs. The program is a tool for DPDC hazard assessment and, as an example, an application to the DPDC deposits of the Agnano-Monte Spina eruption (4.1 ky BP) at Campi Flegrei (Italy) is presented.
Grid point extraction and coding for structured light system
Song, Zhan; Chung, Ronald
2011-09-01
A structured light system simplifies three-dimensional reconstruction by illuminating a specially designed pattern to the target object, thereby generating a distinct texture on it for imaging and further processing. Success of the system hinges upon what features are to be coded in the projected pattern, extracted in the captured image, and matched between the projector's display panel and the camera's image plane. The codes have to be such that they are largely preserved in the image data upon illumination from the projector, reflection from the target object, and projective distortion in the imaging process. The features also need to be reliably extracted in the image domain. In this article, a two-dimensional pseudorandom pattern consisting of rhombic color elements is proposed, and the grid points between the pattern elements are chosen as the feature points. We describe how a type classification of the grid points plus the pseudorandomness of the projected pattern can equip each grid point with a unique label that is preserved in the captured image. We also present a grid point detector that extracts the grid points without the need of segmenting the pattern elements, and that localizes the grid points in subpixel accuracy. Extensive experiments are presented to illustrate that, with the proposed pattern feature definition and feature detector, more features points in higher accuracy can be reconstructed in comparison with the existing pseudorandomly encoded structured light systems.
DEFF Research Database (Denmark)
Nielsen, Mogens; Rozenberg, Grzegorz; Salomaa, Arto
1974-01-01
Continuing the work begun in Part I of this paper, we consider now variations of nondeterministic OL-systems. The present Part II of the paper contains a systematic classification of the effect of nonterminals, codings, weak codings, nonerasing homomorphisms and homomorphisms for all basic variat...
Energy Technology Data Exchange (ETDEWEB)
Napier, B.A.; Peloquin, R.A.; Strenge, D.L.; Ramsdell, J.V.
1988-09-01
The Hanford Environmental Dosimetry Upgrade Project was undertaken to incorporate the internal dosimetry models recommended by the International Commission on Radiological Protection (ICRP) in updated versions of the environmental pathway analysis models used at Hanford. The resulting second generation of Hanford environmental dosimetry computer codes is compiled in the Hanford Environmental Dosimetry System (Generation II, or GENII). This coupled system of computer codes is intended for analysis of environmental contamination resulting from acute or chronic releases to, or initial contamination of, air, water, or soil, on through the calculation of radiation doses to individuals or populations. GENII is described in three volumes of documentation. This volume is a Code Maintenance Manual for the serious user, including code logic diagrams, global dictionary, worksheets to assist with hand calculations, and listings of the code and its associated data libraries. The first volume describes the theoretical considerations of the system. The second volume is a Users' Manual, providing code structure, users' instructions, required system configurations, and QA-related topics. 7 figs., 5 tabs.
The Application Programming Interface for the PVMEXEC Program and Associated Code Coupling System
Energy Technology Data Exchange (ETDEWEB)
Walter L. Weaver III
2005-03-01
This report describes the Application Programming Interface for the PVMEXEC program and the code coupling systems that it implements. The information in the report is intended for programmers wanting to add a new code into the coupling system.
Energy Technology Data Exchange (ETDEWEB)
Shirakawa, Toshihiko [Computer Software Development Co., Ltd., Tokyo (Japan); Hatanaka, Koichiro [Japan Nuclear Cycle Development Inst., Tokai Works, Tokai, Ibaraki (Japan)
2001-11-01
In order to document a basic manual about input data, output data, execution of computer code on groundwater flow and radionuclide transport calculation in heterogeneous porous rock, we investigated the theoretical background about geostatistical computer codes and the user's manual for the computer code on groundwater flow and radionuclide transport which calculates water flow in three dimension, the path of moving radionuclide, and one dimensional radionuclide migration. In this report, based on above investigation we describe the geostatistical background about simulating heterogeneous permeability field. And we describe construction of files, input and output data, a example of calculating of the programs which simulates heterogeneous permeability field, and calculates groundwater flow and radionuclide transport. Therefore, we can document a manual by investigating the theoretical background about geostatistical computer codes and the user's manual for the computer code on groundwater flow and radionuclide transport calculation. And we can model heterogeneous porous rock and analyze groundwater flow and radionuclide transport by utilizing the information from this report. (author)
System for Processing Coded OFDM Under Doppler and Fading
Tsou, Haiping; Darden, Scott; Lee, Dennis; Yan, Tsun-Yee
2005-01-01
An advanced communication system has been proposed for transmitting and receiving coded digital data conveyed as a form of quadrature amplitude modulation (QAM) on orthogonal frequency-division multiplexing (OFDM) signals in the presence of such adverse propagation-channel effects as large dynamic Doppler shifts and frequency-selective multipath fading. Such adverse channel effects are typical of data communications between mobile units or between mobile and stationary units (e.g., telemetric transmissions from aircraft to ground stations). The proposed system incorporates novel signal processing techniques intended to reduce the losses associated with adverse channel effects while maintaining compatibility with the high-speed physical layer specifications defined for wireless local area networks (LANs) as the standard 802.11a of the Institute of Electrical and Electronics Engineers (IEEE 802.11a). OFDM is a multi-carrier modulation technique that is widely used for wireless transmission of data in LANs and in metropolitan area networks (MANs). OFDM has been adopted in IEEE 802.11a and some other industry standards because it affords robust performance under frequency-selective fading. However, its intrinsic frequency-diversity feature is highly sensitive to synchronization errors; this sensitivity poses a challenge to preserve coherence between the component subcarriers of an OFDM system in order to avoid intercarrier interference in the presence of large dynamic Doppler shifts as well as frequency-selective fading. As a result, heretofore, the use of OFDM has been limited primarily to applications involving small or zero Doppler shifts. The proposed system includes a digital coherent OFDM communication system that would utilize enhanced 802.1la-compatible signal-processing algorithms to overcome effects of frequency-selective fading and large dynamic Doppler shifts. The overall transceiver design would implement a two-frequency-channel architecture (see figure
76 FR 4113 - Federal Procurement Data System Product Service Code Manual Update
2011-01-24
... ADMINISTRATION Federal Procurement Data System Product Service Code Manual Update AGENCY: Office of... the Products and Services Code (PSC) Manual, which provides codes to describe products, services, and... [email protected] . SUPPLEMENTARY INFORMATION: The Products and Services Code (PSC) Manual provides...
Simulation of High Power Amplifier Calculation in VSAT System
Directory of Open Access Journals (Sweden)
Indri Neforawati
2009-08-01
Full Text Available Arithmatical simulation of High Power Amplifier (HPA on VSAT system is a program which used to calculate the capacity of HPA as a working test of maximum power on each remote station of the VSAT network system, afterward can be obtained the available capacity value and power capacity used, therefore able to reallocate residual power below its available power spare. VSAT system can be used for several telecommunication application such as video broadcast, data broadcast,audio broadcast, banking operation, ATM and others. Due to the easy operational, maintanance and its instalment, VSAT system is more prifitable compare to ordinary terestrial band, its capability for multiservice application become more flexible in using its network. The software used is Visual Basic 6.0 version and database Microsoft Access. These software take a role as visualization and planning for remote station development and also power capasity needed for each remote in the calculation of HPA.
Magro, G.; Dahle, T. J.; Molinelli, S.; Ciocca, M.; Fossati, P.; Ferrari, A.; Inaniwa, T.; Matsufuji, N.; Ytre-Hauge, K. S.; Mairani, A.
2017-05-01
Particle therapy facilities often require Monte Carlo (MC) simulations to overcome intrinsic limitations of analytical treatment planning systems (TPS) related to the description of the mixed radiation field and beam interaction with tissue inhomogeneities. Some of these uncertainties may affect the computation of effective dose distributions; therefore, particle therapy dedicated MC codes should provide both absorbed and biological doses. Two biophysical models are currently applied clinically in particle therapy: the local effect model (LEM) and the microdosimetric kinetic model (MKM). In this paper, we describe the coupling of the NIRS (National Institute for Radiological Sciences, Japan) clinical dose to the FLUKA MC code. We moved from the implementation of the model itself to its application in clinical cases, according to the NIRS approach, where a scaling factor is introduced to rescale the (carbon-equivalent) biological dose to a clinical dose level. A high level of agreement was found with published data by exploring a range of values for the MKM input parameters, while some differences were registered in forward recalculations of NIRS patient plans, mainly attributable to differences with the analytical TPS dose engine (taken as reference) in describing the mixed radiation field (lateral spread and fragmentation). We presented a tool which is being used at the Italian National Center for Oncological Hadrontherapy to support the comparison study between the NIRS clinical dose level and the LEM dose specification.
A method for calculating active feedback system to provide vertical ...
Indian Academy of Sciences (India)
Home; Journals; Pramana – Journal of Physics; Volume 68; Issue 4. A method for calculating active feedback system to provide vertical position control of plasma in a tokamak. Nizami Gasilov. Research ... Nizami Gasilov1. Faculty of Engineering, Baskent University, Eskisehir Yolu 20. km, Baglica, 06530 Ankara, Turkey ...
Greenhouse Gas Emissions Calculator for Grain and Biofuel Farming Systems
McSwiney, Claire P.; Bohm, Sven; Grace, Peter R.; Robertson, G. Philip
2010-01-01
Opportunities for farmers to participate in greenhouse gas (GHG) credit markets require that growers, students, extension educators, offset aggregators, and other stakeholders understand the impact of agricultural practices on GHG emissions. The Farming Systems Greenhouse Gas Emissions Calculator, a web-based tool linked to the SOCRATES soil…
A novel reporting approach to coronary angiography: "segmental coding system".
Konuralp, Cüneyt; Idiz, Mustafa; Ateş, Mehmet
2005-01-01
A new systematic reporting system for coronary angiography has been developed, which is capable of describing any visible intraluminal or extraluminal conditions with the exact coordinates. In this method, called "segmental coding system"(SCS), the part of the artery that is located between its two subsequent branches is considered to be an "angiographic segment". Conditions are localized according to their relationship with these angiographic segments and the anatomic border of the segments (coronary ostiums, primary, secondary and tertiary branches, grafts and proximal and distal anastomosis sites). They are also described by using a special coding system that consists of letters, numbers and signs. SCS can supply the name (stenosis, occlusion, contour deformity, aneurysm, rupture, anatomical variation, existence of stent, etc.) and the exact localization (coordinates) of the condition with its properties; filling direction, and the collateral system that fills the vessel. We applied SCS to more than 500 cineangiograpies. According to our experience, SCS provides more objective, detailed, and even correct information than the current narrative reporting system. SCS also offers many extra advantages. (a) It can describe all imaginable types of lesion combinations. (b) All of the existing conditions can be listed without missing. (c) The definitions are very precise and clear. They can easily be understood by everyone in the same way. (d) It is more advantageous on archiving, searching the database, and comparing the subsequent reports for the same patient. (e) In the future, by using specially tailored software, personal and detailed angiographic images will be reproduced from the SCS data. By being introduced into clinical practice, we believe, SCS will prove a very useful tool for both surgeons and cardiologists.
Source Code Verification for Embedded Systems using Prolog
Directory of Open Access Journals (Sweden)
Frank Flederer
2017-01-01
Full Text Available System relevant embedded software needs to be reliable and, therefore, well tested, especially for aerospace systems. A common technique to verify programs is the analysis of their abstract syntax tree (AST. Tree structures can be elegantly analyzed with the logic programming language Prolog. Moreover, Prolog offers further advantages for a thorough analysis: On the one hand, it natively provides versatile options to efficiently process tree or graph data structures. On the other hand, Prolog's non-determinism and backtracking eases tests of different variations of the program flow without big effort. A rule-based approach with Prolog allows to characterize the verification goals in a concise and declarative way. In this paper, we describe our approach to verify the source code of a flash file system with the help of Prolog. The flash file system is written in C++ and has been developed particularly for the use in satellites. We transform a given abstract syntax tree of C++ source code into Prolog facts and derive the call graph and the execution sequence (tree, which then are further tested against verification goals. The different program flow branching due to control structures is derived by backtracking as subtrees of the full execution sequence. Finally, these subtrees are verified in Prolog. We illustrate our approach with a case study, where we search for incorrect applications of semaphores in embedded software using the real-time operating system RODOS. We rely on computation tree logic (CTL and have designed an embedded domain specific language (DSL in Prolog to express the verification goals.
Gérard, F.; Clément, A.; Fritz, B.
1998-04-01
It is demonstrated that at steady state, the 1D thermo-kinetic hydrochemical Eulerian mass balance equations in pure advective mode are indeed identical to the governing mass balance equations of a single reaction path (or geochemical) code in open system mode. Thus, both calculated reaction paths should be theoretically identical whatever the chemical complexity of the water-rock system (i.e., multicomponent, multireaction zones kinetically and equilibrium-controlled). We propose to use this property to numerically test the thermo-kinetic hydrochemical Eulerian codes and we employ it to verify the algorithm of the 1D finite difference code KIRMAT. Compared to the other methods to perform such numerical tests (i.e., comparisons with analytical, semi-analytical solutions, between two Eulerian hydrochemical codes), the advantage of this new method is the absence of constraints on the chemical complexity of the modelled water-rock systems. Moreover, the same thermo-kinetic databases and geochemical functions can be easily and mechanically used in both calculations, when the numerical reference comes from the Eulerian code with no transport terms ( u and D=0) and modify to be consistent with the definition of the open system mode in geochemical modelling. The ability of KIRMAT to treat multicomponent pure advective transport, subjected to several kinetically equilibrium-controlled dissolution and precipitation reactions, and to track their boundaries has been successfully verified with the property of interest. The required numerical validation of the reference calculations is bypassed in developing the Eulerian code from an already checked single reaction path code. A forward time-upstream weighting scheme (a mixing cell scheme) is used in this study. An appropriate choice of grid spacing allows to calculate within the grid size uncertainty the correct mineral reaction zone boundaries, despite the presence of numerical dispersion. Its correction enables us to improve
V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009
Energy Technology Data Exchange (ETDEWEB)
Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.
2010-07-15
V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)
Spatiotemporal Coding of Individual Chemicals by the Gustatory System.
Reiter, Sam; Campillo Rodriguez, Chelsey; Sun, Kui; Stopfer, Mark
2015-09-02
Four of the five major sensory systems (vision, olfaction, somatosensation, and audition) are thought to use different but partially overlapping sets of neurons to form unique representations of vast numbers of stimuli. The only exception is gustation, which is thought to represent only small numbers of basic taste categories. However, using new methods for delivering tastant chemicals and making electrophysiological recordings from the tractable gustatory system of the moth Manduca sexta, we found chemical-specific information is as follows: (1) initially encoded in the population of gustatory receptor neurons as broadly distributed spatiotemporal patterns of activity; (2) dramatically integrated and temporally transformed as it propagates to monosynaptically connected second-order neurons; and (3) observed in tastant-specific behavior. Our results are consistent with an emerging view of the gustatory system: rather than constructing basic taste categories, it uses a spatiotemporal population code to generate unique neural representations of individual tastant chemicals. Our results provide a new view of taste processing. Using a new, relatively simple model system and a new set of techniques to deliver taste stimuli and to examine gustatory receptor neurons and their immediate followers, we found no evidence for labeled line connectivity, or basic taste categories such as sweet, salty, bitter, and sour. Rather, individual tastant chemicals are represented as patterns of spiking activity distributed across populations of receptor neurons. These representations are transformed substantially as multiple types of receptor neurons converge upon follower neurons, leading to a combinatorial coding format that uniquely, rapidly, and efficiently represents individual taste chemicals. Finally, we found that the information content of these neurons can drive tastant-specific behavior. Copyright © 2015 the authors 0270-6474/15/3512309-13$15.00/0.
Carbon footprint calculation model for the Mexican food equivalent system
Directory of Open Access Journals (Sweden)
Salvador Ruiz Cerrillo
2017-06-01
Full Text Available Introduction: the impact environment trough the anthropogenic action has been contributed to the fast production of greenhouse gases effect (GHG, a way to estimate the quantity of these substances is the carbon footprint (CF, nowadays it does not exist enough models for the calculation of food carbon footprint. Objective: the aim of this study was to design a calculation model for the measurement of the carbon footprint on the Mexican food equivalent system. Methods: it was about a retrospective study, a bibliographic review was made with original and review articles in different specialized researchers, there were included publications in English and Spanish, also published from 2000 to 2016. Results: a reference table was proposed for the food carbon footprint calculation on the Mexican food equivalent system trough the carbon intensity indicator, which is determined by the grams of emissions equivalents of carbon dioxide (CO2 in relation with the energetic contribution of each food equivalent. Conclusion: in a conclusion manner, estimating food carbon footprint is still a challenge, mean while the calculation models proposal is important to estimate the production of GHG trough a more sustainable food system.
FPGA based digital phase-coding quantum key distribution system
Lu, XiaoMing; Zhang, LiJun; Wang, YongGang; Chen, Wei; Huang, DaJun; Li, Deng; Wang, Shuang; He, DeYong; Yin, ZhenQiang; Zhou, Yu; Hui, Cong; Han, ZhengFu
2015-12-01
Quantum key distribution (QKD) is a technology with the potential capability to achieve information-theoretic security. Phasecoding is an important approach to develop practical QKD systems in fiber channel. In order to improve the phase-coding modulation rate, we proposed a new digital-modulation method in this paper and constructed a compact and robust prototype of QKD system using currently available components in our lab to demonstrate the effectiveness of the method. The system was deployed in laboratory environment over a 50 km fiber and continuously operated during 87 h without manual interaction. The quantum bit error rate (QBER) of the system was stable with an average value of 3.22% and the secure key generation rate is 8.91 kbps. Although the modulation rate of the photon in the demo system was only 200 MHz, which was limited by the Faraday-Michelson interferometer (FMI) structure, the proposed method and the field programmable gate array (FPGA) based electronics scheme have a great potential for high speed QKD systems with Giga-bits/second modulation rate.
Thermodynamic calculations in ternary titanium–aluminium–manganese system
Directory of Open Access Journals (Sweden)
ANA I. KOSTOV
2008-04-01
Full Text Available Thermodynamic calculations in the ternary Ti–Al–Mn system are shown in this paper. The thermodynamic calculations were performed using the FactSage thermochemical software and database, with the aim of determining thermodynamic properties, such as activities, coefficient of activities, partial and integral values of the enthalpies and Gibbs energies of mixing and excess energies at two different temperatures: 2000 and 2100 K. Bearing in mind that no experimental data for the Ti–Al–Mn ternary system have been obtained or reported. The obtained results represent a good base for further thermodynamic analysis and may be useful as a comparison with some future critical experimental results and thermodynamic optimization of this system.
Calculating Outcrossing Rates used in Decision Support Systems for Ships
DEFF Research Database (Denmark)
Nielsen, Ulrik Dam
2008-01-01
Onboard decision support systems (DSS) are used to increase the operational safety of ships. Ideally, DSS can estimate - in the statistical sense - future ship responses on a time scale of the order of 1-3 hours taking into account speed and course changes. The calculations depend on both...... operational and environmental parameters that are known only in the statistical sense. The present paper suggests a procedure to incorporate random variables and associated uncertainties in calculations of outcrossing rates, which are the basis for risk-based DSS. The procedure is based on parallel system...... analysis, and the paper derives and describes the main ideas. The concept is illustrated by an example, where the limit state of a non-linear ship response is considered. The results from the parallel system analysis are in agreement with corresponding Monte Carlo simulations. However, the computational...
Design of ACM system based on non-greedy punctured LDPC codes
Lu, Zijun; Jiang, Zihong; Zhou, Lin; He, Yucheng
2017-08-01
In this paper, an adaptive coded modulation (ACM) scheme based on rate-compatible LDPC (RC-LDPC) codes was designed. The RC-LDPC codes were constructed by a non-greedy puncturing method which showed good performance in high code rate region. Moreover, the incremental redundancy scheme of LDPC-based ACM system over AWGN channel was proposed. By this scheme, code rates vary from 2/3 to 5/6 and the complication of the ACM system is lowered. Simulations show that more and more obvious coding gain can be obtained by the proposed ACM system with higher throughput.
Overview of Particle and Heavy Ion Transport Code System PHITS
Sato, Tatsuhiko; Niita, Koji; Matsuda, Norihiro; Hashimoto, Shintaro; Iwamoto, Yosuke; Furuta, Takuya; Noda, Shusaku; Ogawa, Tatsuhiko; Iwase, Hiroshi; Nakashima, Hiroshi; Fukahori, Tokio; Okumura, Keisuke; Kai, Tetsuya; Chiba, Satoshi; Sihver, Lembit
2014-06-01
A general purpose Monte Carlo Particle and Heavy Ion Transport code System, PHITS, is being developed through the collaboration of several institutes in Japan and Europe. The Japan Atomic Energy Agency is responsible for managing the entire project. PHITS can deal with the transport of nearly all particles, including neutrons, protons, heavy ions, photons, and electrons, over wide energy ranges using various nuclear reaction models and data libraries. It is written in Fortran language and can be executed on almost all computers. All components of PHITS such as its source, executable and data-library files are assembled in one package and then distributed to many countries via the Research organization for Information Science and Technology, the Data Bank of the Organization for Economic Co-operation and Development's Nuclear Energy Agency, and the Radiation Safety Information Computational Center. More than 1,000 researchers have been registered as PHITS users, and they apply the code to various research and development fields such as nuclear technology, accelerator design, medical physics, and cosmic-ray research. This paper briefly summarizes the physics models implemented in PHITS, and introduces some important functions useful for specific applications, such as an event generator mode and beam transport functions.
Laser system range calculations and the Lambert W function.
Steinvall, Ove
2009-02-01
The knowledge of range performance versus atmospheric transmission, often given by the visibility, is critical for the design, use, and prediction of laser and passive electro-optic systems. I present a solution of the ladar-lidar equation based on Lambert's W function. This solution will reveal the dependence of the maximum range on the system and target parameters for different atmospheric attenuations and will also allow us to take the signal statistics into account by studying the influence on the threshold signal-to-noise ratio. The method is also applicable to many range calculations for passive systems where the atmospheric loss can be approximated by an exponential term.
Energy Technology Data Exchange (ETDEWEB)
Takada, Tomoyuki; Miyoshi, Yoshinori; Katakura, Jun-ichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2003-03-01
In order to perform accuracy evaluation of the critical calculation by the combination of multi-group constant library MGCL and 3-dimensional Monte Carlo code KENO-IV among critical safety evaluation code system JACS, benchmark calculation was carried out from 1980 in 1982. Some cases where the neutron multiplication factor calculated in the heterogeneous system in it was less than 0.95 were seen. In this report, it re-calculated by considering the cause about the heterogeneous system of the U+Pu nitric acid solution systems containing the neutron poison shown in JAERI-M 9859. The present study has shown that the k{sub eff} value less than 0.95 given in JAERI-M 9859 is caused by the fact that the water reflector below a cylindrical container was not taken into consideration in the KENO-IV calculation model. By taking into the water reflector, the KENO-IV calculation gives a k{sub eff} value greater than 0.95 and a good agreement with the experiment. (author)
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
Energy Technology Data Exchange (ETDEWEB)
Artmann, Andreas; Meyering, Henrich
2016-11-15
The GRS research project was aimed to the test the appropriateness of the software package ''residual radioactivity'' (RESRAD) for the calculation of clearance values according to German and European regulations. Comparative evaluations were performed with RESRAD-OFFSITE and the code SiWa-PRO DSS used by GRS and the GRS program code ARTM. It is recommended to use RESRAD-OFFSITE for comparative calculations. The dose relevant air-path dispersion of radionuclides should not be modeled using RESRAD-OFFSITE, the use of ARTM is recommended. The sensitivity analysis integrated into RESRAD-OFFSITE allows a fast identification of crucial parameters.
WIPP shaft seal system parameters recommended to support compliance calculations
Energy Technology Data Exchange (ETDEWEB)
Hurtado, L.D.; Knowles, M.K. [Sandia National Labs., Albuquerque, NM (United States); Kelley, V.A.; Jones, T.L.; Ogintz, J.B. [INTERA Inc., Austin, TX (United States); Pfeifle, T.W. [RE/SPEC, Inc., Rapid City, SD (United States)
1997-12-01
The US Department of Energy plans to dispose of transuranic waste at the Waste Isolation Pilot Plant (WIPP), which is sited in southeastern New Mexico. The WIPP disposal facility is located approximately 2,150 feet (650 m) below surface in the bedded halite of the Salado Formation. Prior to initiation of disposal activities, the Department of Energy must demonstrate that the WIPP will comply with all regulatory requirements. Applicable regulations require that contaminant releases from the WIPP remain below specified levels for a period of 10,000 years. To demonstrate that the WIPP will comply with these regulations, the Department of Energy has requested that Sandia National Laboratories develop and implement a comprehensive performance assessment of the WIPP repository for the regulatory period. This document presents the conceptual model of the shaft sealing system to be implemented in performance assessment calculations conducted in support of the Compliance Certification Application for the WIPP. The model was developed for use in repository-scale calculations and includes the seal system geometry and materials to be used in grid development as well as all parameters needed to describe the seal materials. These calculations predict the hydrologic behavior of the system. Hence conceptual model development is limited to those processes that could impact the fluid flow through the seal system.
Simulation realization of 2-D wavelength/time system utilizing MDW code for OCDMA system
Azura, M. S. A.; Rashidi, C. B. M.; Aljunid, S. A.; Endut, R.; Ali, N.
2017-11-01
This paper presents a realization of Wavelength/Time (W/T) Two-Dimensional Modified Double Weight (2-D MDW) code for Optical Code Division Multiple Access (OCDMA) system based on Spectral Amplitude Coding (SAC) approach. The MDW code has the capability to suppress Phase-Induce Intensity Noise (PIIN) and minimizing the Multiple Access Interference (MAI) noises. At the permissible BER 10-9, the 2-D MDW (APD) had shown minimum effective received power (Psr) = -71 dBm that can be obtained at the receiver side as compared to 2-D MDW (PIN) only received -61 dBm. The results show that 2-D MDW (APD) has better performance in achieving same BER with longer optical fiber length and with less received power (Psr). Also, the BER from the result shows that MDW code has the capability to suppress PIIN ad MAI.
Energy Technology Data Exchange (ETDEWEB)
Hesse, U.
1991-01-01
There are plans to also use plutonium containing fuel elements (mixed oxide fuel) in the BWR type reactors, with a proportion of up to one third of the entire fuel core. The new concept uses complete MOX fuel elements, as are used in the PWR type reactors. The OREST computer code has been designed for burnup calculations in PWRs. The situation in BWRs is different, as in these reactor types, fuel elements are heterogenous in design, and burnup calculations have to take into account the axial variations of the void fraction, so that multi-dimensional effects have to be calculated. The report explains that the one-dimensional OREST code can be enhanced by supplementing calculations, performed with the Monte-Carlo type KENO code in this case, and is thus suitable without restrictions for performing burnup calculations for MOX fuel elements in BWRs. The calculation method and performance is illustrated by the example of a UO{sub 2} fuel element of the Wuergassen reactor. The model calculations predict a relatively high residual activity in the upper part of the fuel element, and a distinct curium buildup in the lower third of the BWR fuel element. (orig./HP).
Energy Technology Data Exchange (ETDEWEB)
Casanovas, R., E-mail: ramon.casanovas@urv.cat [Unitat de Fisica Medica, Facultat de Medicina i Ciencies de la Salut, Universitat Rovira i Virgili, ES-43201 Reus (Tarragona) (Spain); Morant, J.J. [Servei de Proteccio Radiologica, Facultat de Medicina i Ciencies de la Salut, Universitat Rovira i Virgili, ES-43201 Reus (Tarragona) (Spain); Salvado, M. [Unitat de Fisica Medica, Facultat de Medicina i Ciencies de la Salut, Universitat Rovira i Virgili, ES-43201 Reus (Tarragona) (Spain)
2012-05-21
The radiation detectors yield the optimal performance if they are accurately calibrated. This paper presents the energy, resolution and efficiency calibrations for two scintillation detectors, NaI(Tl) and LaBr{sub 3}(Ce). For the two former calibrations, several fitting functions were tested. To perform the efficiency calculations, a Monte Carlo user code for the EGS5 code system was developed with several important implementations. The correct performance of the simulations was validated by comparing the simulated spectra with the experimental spectra and reproducing a number of efficiency and activity calculations. - Highlights: Black-Right-Pointing-Pointer NaI(Tl) and LaBr{sub 3}(Ce) scintillation detectors are used for gamma-ray spectrometry. Black-Right-Pointing-Pointer Energy, resolution and efficiency calibrations are discussed for both detectors. Black-Right-Pointing-Pointer For the two former calibrations, several fitting functions are tested. Black-Right-Pointing-Pointer A Monte Carlo user code for EGS5 was developed for the efficiency calculations. Black-Right-Pointing-Pointer The code was validated reproducing some efficiency and activity calculations.
Verification of the CONPAS (CONtainment Performance Analysis System) code package
Energy Technology Data Exchange (ETDEWEB)
Kim, See Darl; Ahn, Kwang Il; Song, Yong Man; Choi, Young; Park, Soo Yong; Kim, Dong Ha; Jin, Young Ho
1997-09-01
CONPAS is a computer code package to integrate the numerical, graphical, and results-oriented aspects of Level 2 probabilistic safety assessment (PSA) for nuclear power plants under a PC window environment automatically. For the integrated analysis of Level 2 PSA, the code utilizes four distinct, but closely related modules: (1) ET Editor, (2) Computer, (3) Text Editor, and (4) Mechanistic Code Plotter. Compared with other existing computer codes for Level 2 PSA, and CONPAS code provides several advanced features: computational aspects including systematic uncertainty analysis, importance analysis, sensitivity analysis and data interpretation, reporting aspects including tabling and graphic as well as user-friendly interface. The computational performance of CONPAS has been verified through a Level 2 PSA to a reference plant. The results of the CONPAS code was compared with an existing level 2 PSA code (NUCAP+) and the comparison proves that CONPAS is appropriate for Level 2 PSA. (author). 9 refs., 8 tabs., 14 figs.
Hybrid Compton camera/coded aperture imaging system
Mihailescu, Lucian [Livermore, CA; Vetter, Kai M [Alameda, CA
2012-04-10
A system in one embodiment includes an array of radiation detectors; and an array of imagers positioned behind the array of detectors relative to an expected trajectory of incoming radiation. A method in another embodiment includes detecting incoming radiation with an array of radiation detectors; detecting the incoming radiation with an array of imagers positioned behind the array of detectors relative to a trajectory of the incoming radiation; and performing at least one of Compton imaging using at least the imagers and coded aperture imaging using at least the imagers. A method in yet another embodiment includes detecting incoming radiation with an array of imagers positioned behind an array of detectors relative to a trajectory of the incoming radiation; and performing Compton imaging using at least the imagers.
The Tubulin Code: A Navigation System for Chromosomes during Mitosis.
Barisic, Marin; Maiato, Helder
2016-10-01
Before chromosomes segregate during mitosis in metazoans, they align at the cell equator by a process known as chromosome congression. This is in part mediated by the coordinated activities of kinetochore motors with opposite directional preferences that transport peripheral chromosomes along distinct spindle microtubule populations. Because spindle microtubules are all made from the same α/β-tubulin heterodimers, a critical longstanding question has been how chromosomes are guided to specific locations during mitosis. This implies the existence of spatial cues/signals on specific spindle microtubules that are read by kinetochore motors on chromosomes and ultimately indicate the way towards the equator. Here, we discuss the emerging concept that tubulin post-translational modifications (PTMs), as part of the so-called tubulin code, work as a navigation system for kinetochore-based chromosome motility during early mitosis. Copyright © 2016 Elsevier Ltd. All rights reserved.
Biometric iris image acquisition system with wavefront coding technology
Hsieh, Sheng-Hsun; Yang, Hsi-Wen; Huang, Shao-Hung; Li, Yung-Hui; Tien, Chung-Hao
2013-09-01
Biometric signatures for identity recognition have been practiced for centuries. Basically, the personal attributes used for a biometric identification system can be classified into two areas: one is based on physiological attributes, such as DNA, facial features, retinal vasculature, fingerprint, hand geometry, iris texture and so on; the other scenario is dependent on the individual behavioral attributes, such as signature, keystroke, voice and gait style. Among these features, iris recognition is one of the most attractive approaches due to its nature of randomness, texture stability over a life time, high entropy density and non-invasive acquisition. While the performance of iris recognition on high quality image is well investigated, not too many studies addressed that how iris recognition performs subject to non-ideal image data, especially when the data is acquired in challenging conditions, such as long working distance, dynamical movement of subjects, uncontrolled illumination conditions and so on. There are three main contributions in this paper. Firstly, the optical system parameters, such as magnification and field of view, was optimally designed through the first-order optics. Secondly, the irradiance constraints was derived by optical conservation theorem. Through the relationship between the subject and the detector, we could estimate the limitation of working distance when the camera lens and CCD sensor were known. The working distance is set to 3m in our system with pupil diameter 86mm and CCD irradiance 0.3mW/cm2. Finally, We employed a hybrid scheme combining eye tracking with pan and tilt system, wavefront coding technology, filter optimization and post signal recognition to implement a robust iris recognition system in dynamic operation. The blurred image was restored to ensure recognition accuracy over 3m working distance with 400mm focal length and aperture F/6.3 optics. The simulation result as well as experiment validates the proposed code
Control code for laboratory adaptive optics teaching system
Jin, Moonseob; Luder, Ryan; Sanchez, Lucas; Hart, Michael
2017-09-01
By sensing and compensating wavefront aberration, adaptive optics (AO) systems have proven themselves crucial in large astronomical telescopes, retinal imaging, and holographic coherent imaging. Commercial AO systems for laboratory use are now available in the market. One such is the ThorLabs AO kit built around a Boston Micromachines deformable mirror. However, there are limitations in applying these systems to research and pedagogical projects since the software is written with limited flexibility. In this paper, we describe a MATLAB-based software suite to interface with the ThorLabs AO kit by using the MATLAB Engine API and Visual Studio. The software is designed to offer complete access to the wavefront sensor data, through the various levels of processing, to the command signals to the deformable mirror and fast steering mirror. In this way, through a MATLAB GUI, an operator can experiment with every aspect of the AO system's functioning. This is particularly valuable for tests of new control algorithms as well as to support student engagement in an academic environment. We plan to make the code freely available to the community.
Perturbation theory calculations of model pair potential systems
Energy Technology Data Exchange (ETDEWEB)
Gong, Jianwu [Iowa State Univ., Ames, IA (United States)
2016-01-01
Helmholtz free energy is one of the most important thermodynamic properties for condensed matter systems. It is closely related to other thermodynamic properties such as chemical potential and compressibility. It is also the starting point for studies of interfacial properties and phase coexistence if free energies of different phases can be obtained. In this thesis, we will use an approach based on the Weeks-Chandler-Anderson (WCA) perturbation theory to calculate the free energy of both solid and liquid phases of Lennard-Jones pair potential systems and the free energy of liquid states of Yukawa pair potentials. Our results indicate that the perturbation theory provides an accurate approach to the free energy calculations of liquid and solid phases based upon comparisons with results from molecular dynamics (MD) and Monte Carlo (MC) simulations.
Energy Technology Data Exchange (ETDEWEB)
New, Joshua Ryan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Levinson, Ronnen [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Huang, Yu [White Box Technologies, Salt Lake City, UT (United States); Sanyal, Jibonananda [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Miller, William A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mellot, Joe [The Garland Company, Cleveland, OH (United States); Childs, Kenneth W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kriner, Scott [Green Metal Consulting, Inc., Macungie, PA (United States)
2014-06-01
The Roof Savings Calculator (RSC) was developed through collaborations among Oak Ridge National Laboratory (ORNL), White Box Technologies, Lawrence Berkeley National Laboratory (LBNL), and the Environmental Protection Agency in the context of a California Energy Commission Public Interest Energy Research project to make cool-color roofing materials a market reality. The RSC website and a simulation engine validated against demonstration homes were developed to replace the liberal DOE Cool Roof Calculator and the conservative EPA Energy Star Roofing Calculator, which reported different roof savings estimates. A preliminary analysis arrived at a tentative explanation for why RSC results differed from previous LBNL studies and provided guidance for future analysis in the comparison of four simulation programs (doe2attic, DOE-2.1E, EnergyPlus, and MicroPas), including heat exchange between the attic surfaces (principally the roof and ceiling) and the resulting heat flows through the ceiling to the building below. The results were consolidated in an ORNL technical report, ORNL/TM-2013/501. This report is an in-depth inter-comparison of four programs with detailed measured data from an experimental facility operated by ORNL in South Carolina in which different segments of the attic had different roof and attic systems.
Modelling guidelines for core exit temperature simulations with system codes
Energy Technology Data Exchange (ETDEWEB)
Freixa, J., E-mail: jordi.freixa-terradas@upc.edu [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain); Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland); Martínez-Quiroga, V., E-mail: victor.martinez@nortuen.com [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain); Zerkak, O., E-mail: omar.zerkak@psi.ch [Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland); Reventós, F., E-mail: francesc.reventos@upc.edu [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain)
2015-05-15
Highlights: • Core exit temperature is used in PWRs as an indication of core heat up. • Modelling guidelines of CET response with system codes. • Modelling of heat transfer processes in the core and UP regions. - Abstract: Core exit temperature (CET) measurements play an important role in the sequence of actions under accidental conditions in pressurized water reactors (PWR). Given the difficulties in placing measurements in the core region, CET readings are used as criterion for the initiation of accident management (AM) procedures because they can indicate a core heat up scenario. However, the CET responses have some limitation in detecting inadequate core cooling and core uncovery simply because the measurement is not placed inside the core. Therefore, it is of main importance in the field of nuclear safety for PWR power plants to assess the capabilities of system codes for simulating the relation between the CET and the peak cladding temperature (PCT). The work presented in this paper intends to address this open question by making use of experimental work at integral test facilities (ITF) where experiments related to the evolution of the CET and the PCT during transient conditions have been carried out. In particular, simulations of two experiments performed at the ROSA/LSTF and PKL facilities are presented. The two experiments are part of a counterpart exercise between the OECD/NEA ROSA-2 and OECD/NEA PKL-2 projects. The simulations are used to derive guidelines in how to correctly reproduce the CET response during a core heat up scenario. Three aspects have been identified to be of main importance: (1) the need for a 3-dimensional representation of the core and Upper Plenum (UP) regions in order to model the heterogeneity of the power zones and axial areas, (2) the detailed representation of the active and passive heat structures, and (3) the use of simulated thermocouples instead of steam temperatures to represent the CET readings.
Development of system analysis code for thermal-hydraulic simulation of integral reactor, Rex-10
Energy Technology Data Exchange (ETDEWEB)
NONE
2010-10-15
Rex-10 is an environment-friendly and economical small-scale nuclear reactor to provide the energy for district heating as well as the electric power in micro-grid. This integral reactor comprises several innovative concepts supported by advanced primary circuit components, low coolant parameters and natural circulation cooling. To evaluate the system performance and thermal-hydraulic behavior of the reactor, a system analysis code is being developed so that the new designs and technologies adopted in Rex-10 can be reflected. The research efforts are absorbed in programming the simple and fast-running thermal-hydraulic analysis software. The details of hydrodynamic governing equations component models and numerical solution scheme used in this code are presented in this paper. On the basis of one-dimensional momentum integral model, the models of point reactor neutron kinetics for thorium-fueled core, physical processes in the steam-gas pressurizer, and heat transfers in helically coiled steam generator are implemented to the system code. Implicit numerical scheme is employed to momentum and energy equations to assure the numerical stability. The accuracy of simulation is validated by applying the solution method to the Rex-10 test facility. Calculated natural circulation flow rate and coolant temperature at steady-state are compared to the experimental data. The validation is also carried out for the transients in which the sudden reduction in the core power or the feedwater flow takes place. The code's capability to predict the steady-state flow by natural convection and the qualitative behaviour of the primary system in the transients is confirmed. (Author)
Energy Technology Data Exchange (ETDEWEB)
Delbecq, J.M
1999-07-01
The Aster code is a 2D or 3D finite-element calculation code for structures developed by the R and D direction of Electricite de France (EdF). This dossier presents a complete overview of the characteristics and uses of the Aster code: introduction of version 4; the context of Aster (organisation of the code development, versions, systems and interfaces, development tools, quality assurance, independent validation); static mechanics (linear thermo-elasticity, Euler buckling, cables, Zarka-Casier method); non-linear mechanics (materials behaviour, big deformations, specific loads, unloading and loss of load proportionality indicators, global algorithm, contact and friction); rupture mechanics (G energy restitution level, restitution level in thermo-elasto-plasticity, 3D local energy restitution level, KI and KII stress intensity factors, calculation of limit loads for structures), specific treatments (fatigue, rupture, wear, error estimation); meshes and models (mesh generation, modeling, loads and boundary conditions, links between different modeling processes, resolution of linear systems, display of results etc..); vibration mechanics (modal and harmonic analysis, dynamics with shocks, direct transient dynamics, seismic analysis and aleatory dynamics, non-linear dynamics, dynamical sub-structuring); fluid-structure interactions (internal acoustics, mass, rigidity and damping); linear and non-linear thermal analysis; steels and metal industry (structure transformations); coupled problems (internal chaining, internal thermo-hydro-mechanical coupling, chaining with other codes); products and services. (J.S.)
Two-Layer Coding Rate Optimization in Relay-Aided Systems
DEFF Research Database (Denmark)
Sun, Fan
2011-01-01
We consider a three-node transmission system, where a source node conveys a data block to a destination node with the help of a half-duplex decode and-forward (DF) relay node. The whole data block is transmitted as a sequence of packets. For reliable transmission in the three-node system, a two......-layer coding scheme is proposed, where physical layer channel coding is utilized within each packet for error-correction and random network coding is applied on top of channel coding for network error-control. There is a natural tradeoff between the physical layer coding rate and the network coding rate given...
Calculating life? Duelling discourses in interdisciplinary systems biology.
Calvert, Jane; Fujimura, Joan H
2011-06-01
A high profile context in which physics and biology meet today is in the new field of systems biology. Systems biology is a fascinating subject for sociological investigation because the demands of interdisciplinary collaboration have brought epistemological issues and debates front and centre in discussions amongst systems biologists in conference settings, in publications, and in laboratory coffee rooms. One could argue that systems biologists are conducting their own philosophy of science. This paper explores the epistemic aspirations of the field by drawing on interviews with scientists working in systems biology, attendance at systems biology conferences and workshops, and visits to systems biology laboratories. It examines the discourses of systems biologists, looking at how they position their work in relation to previous types of biological inquiry, particularly molecular biology. For example, they raise the issue of reductionism to distinguish systems biology from molecular biology. This comparison with molecular biology leads to discussions about the goals and aspirations of systems biology, including epistemic commitments to quantification, rigor and predictability. Some systems biologists aspire to make biology more similar to physics and engineering by making living systems calculable, modelable and ultimately predictable-a research programme that is perhaps taken to its most extreme form in systems biology's sister discipline: synthetic biology. Other systems biologists, however, do not think that the standards of the physical sciences are the standards by which we should measure the achievements of systems biology, and doubt whether such standards will ever be applicable to 'dirty, unruly living systems'. This paper explores these epistemic tensions and reflects on their sociological dimensions and their consequences for future work in the life sciences. Copyright © 2010 Elsevier Ltd. All rights reserved.
An efficient code for calculation of the 6C, 9C and 12C symbols for C3v, T, and O point groups
Nikitin, A. V.
2012-03-01
A new code designed to calculate the 6 C, 9 C, and 12 C symbols for C3v, T, and O point groups is presented. The program is based on an algorithm that uses the symmetry property between pair and impair representations. This algorithm allows one to speed up the C-symbols calculation and increase the efficiency of spectroscopic programs based on the irreducible tensorial formalism. Program summaryProgram title: 6912C Catalogue identifier: AEKZ_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEKZ_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 1214 No. of bytes in distributed program, including test data, etc.: 22 097 Distribution format: tar.gz Programming language: C++ Computer: Any computer with C, C++ compiler Operating system: Linux SUSE, Windows XP64 RAM: 400 Kb Classification: 4.2, 16.2, 16.3 Nature of problem: Spectroscopy of symmetric atmospheric molecules. Solution method: The program is based on an algorithm that uses the symmetry property between pair and impair representations. Running time: The test program provided takes a few seconds for C3v, a few minutes for T and a few days for O.
Directory of Open Access Journals (Sweden)
Gholamzadeh Zohreh
2014-12-01
Full Text Available Decrease of the economically accessible uranium resources and the inherent proliferation resistance of thorium fuel motivate its application in nuclear power systems. Estimation of the nuclear reactor’s neutronic parameters during different operational situations is of key importance for the safe operation of nuclear reactors. In the present research, thorium oxide fuel burn-up calculations for a demonstrative model of a heavy water- -cooled reactor have been performed using MCNPX 2.6 code. Neutronic parameters for three different thorium fuel matrices loaded separately in the modelled thermal core have been investigated. 233U, 235U and 239Pu isotopes have been used as fissile element in the thorium oxide fuel, separately. Burn-up of three different fuels has been calculated at 1 MW constant power. 135X and 149Sm concentration variations have been studied in the modelled core during 165 days burn-up. Burn-up of thorium oxide enriched with 233U resulted in the least 149Sm and 135Xe productions and net fissile production of 233U after 165 days. The negative fuel, coolant and void reactivity of the used fuel assures safe operation of the modelled thermal core containing (233U-Th O2 matrix. Furthermore, utilisation of thorium breeder fuel demonstrates several advantages, such as good neutronic economy, 233U production and less production of long-lived α emitter high radiotoxic wastes in biological internal exposure point of view
Energy Technology Data Exchange (ETDEWEB)
Marie, S. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France)], E-mail: stephane.marie@cea.fr; Chapuliot, S.; Kayser, Y. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France); Lacire, M.H. [CEA Saclay, DEN/DDIN, 91191 Gif sur Yvette Cedex (France); Drubay, B. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France); Barthelet, B. [EDF/EPN, Site Cap Ampere, 1 Place Pleyel 93207, Saint Denis Cedex 1 (France); Le Delliou, P. [EDF Pole Industrie - Division R and D, Site des Renardieres, Route de Sens, Ecuelles, 77250 Moret sur Loing Cedex (France); Rougier, V. [EDF/UTO, SIS/GAM, 6, Avenue Montaigne, 93192 Noisy le Grand (France); Naudin, C. [EDF/SEPTEN, 12-14, avenue Dutrievoz, 69628 Villeurbanne Cedex (France); Gilles, P.; Triay, M. [AREVA ANP, Tour AREVA, 92084 Paris La Defense Cedex 16 (France)
2007-10-15
Two French nuclear codes include flaw assessment procedures: the RSE-M Code 'Rules for In-service Inspection of Nuclear Power Plant Components' and the RCC-MR code 'Design and Construction rules for mechanical components of FBR nuclear islands and high temperature applications'. Development of analytical methods has been made for the last 10 years through a collaboration between CEA, EDF and AREVA-NP, and through R and D actions involving CEA and IRSN. These activities have led to unification of the common methods of the two codes. The calculation of fracture mechanics parameters, and in particular the stress intensity factor K{sub I} and the J integral, has been widely developed for industrial configurations. All the developments have been integrated in the 2005 edition of RSE-M and in 2007 edition of RCC-MR. This series of papers is composed of five parts: the first presents an overview of the methods proposed in the RCC-MR and RSE-M codes. Parts II-IV provide compendia for specific components. The geometries are plates (part II), pipes (part III) and elbows (part IV). Part V presents validation of the methods, with details on their accuracy. This paper presents the stress intensity factor and J calculation for cracked elbows. General data applicable for all defect geometries are first presented, and then, compendia for K{sub I} and {sigma}{sub ref} calculations are provided for the available defect geometries.
KIM, Jong Woon; LEE, Young-Ouk
2017-09-01
As computing power gets better and better, computer codes that use a deterministic method seem to be less useful than those using the Monte Carlo method. In addition, users do not like to think about space, angles, and energy discretization for deterministic codes. However, a deterministic method is still powerful in that we can obtain a solution of the flux throughout the problem, particularly as when particles can barely penetrate, such as in a deep penetration problem with small detection volumes. Recently, a new state-of-the-art discrete-ordinates code, ATTILA, was developed and has been widely used in several applications. ATTILA provides the capabilities to solve geometrically complex 3-D transport problems by using an unstructured tetrahedral mesh. Since 2009, we have been developing our own code by benchmarking ATTILA. AETIUS is a discrete ordinates code that uses an unstructured tetrahedral mesh such as ATTILA. For pre- and post- processing, Gmsh is used to generate an unstructured tetrahedral mesh by importing a CAD file (*.step) and visualizing the calculation results of AETIUS. Using a CAD tool, the geometry can be modeled very easily. In this paper, we describe a brief overview of AETIUS and provide numerical results from both AETIUS and a Monte Carlo code, MCNP5, in a deep penetration problem with small detection volumes. The results demonstrate the effectiveness and efficiency of AETIUS for such calculations.
Energy Technology Data Exchange (ETDEWEB)
Maconald, J.L.; Cashwell, E.D.
1978-09-01
The techniques of learning theory and pattern recognition are used to learn splitting surface locations for the Monte Carlo neutron transport code MCN. A study is performed to determine default values for several pattern recognition and learning parameters. The modified MCN code is used to reduce computer cost for several nontrivial example problems.
Low-density parity-check codes for volume holographic memory systems.
Pishro-Nik, Hossein; Rahnavard, Nazanin; Ha, Jeongseok; Fekri, Faramarz; Adibi, Ali
2003-02-10
We investigate the application of low-density parity-check (LDPC) codes in volume holographic memory (VHM) systems. We show that a carefully designed irregular LDPC code has a very good performance in VHM systems. We optimize high-rate LDPC codes for the nonuniform error pattern in holographic memories to reduce the bit error rate extensively. The prior knowledge of noise distribution is used for designing as well as decoding the LDPC codes. We show that these codes have a superior performance to that of Reed-Solomon (RS) codes and regular LDPC counterparts. Our simulation shows that we can increase the maximum storage capacity of holographic memories by more than 50 percent if we use irregular LDPC codes with soft-decision decoding instead of conventionally employed RS codes with hard-decision decoding. The performance of these LDPC codes is close to the information theoretic capacity.
Neutron and photon transport calculations in fusion system. 2
Energy Technology Data Exchange (ETDEWEB)
Sato, Satoshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment
1998-03-01
On the application of MCNP to the neutron and {gamma}-ray transport calculations for fusion reactor system, the wide range design calculation has been carried out in the engineering design activities for the international thermonuclear fusion experimental reactor (ITER) being developed jointly by Japan, USA, EU and Russia. As the objects of shielding calculation for fusion reactors, there are the assessment of dose equivalent rate for living body shielding and the assessment of the nuclear response for the soundness of in-core structures. In the case that the detailed analysis of complicated three-dimensional shapes is required, the assessment using MCNP has been carried out. Also when the nuclear response of peripheral equipment due to the gap streaming between blanket modules is evaluated with good accuracy, the calculation with MCNP has been carried out. The analyses of the shieldings for blanket modules and NBI port are explained, and the examples of the results of analyses are shown. In the blanket modules, there are penetrating holes and continuous gap. In the case of the NBI port, shielding plug cannot be installed. These facts necessitate the MCNP analysis with high accuracy. (K.I.)
Research on Traceability System of Food Safety Based on PDF417 Two-Dimensional Bar Code
Xu, Shipu; Liu, Muhua; Zhao, Jingyin; Yuan, Tao; Wang, Yunsheng
2010-01-01
International audience; In this paper, through the research on two-dimensional bar codes of food source tracking; a set of traceability tracking system was designed based on the two bar coding algorithm and decoding algorithms. This system could generate PDF417 code based on food production, logistics and transport, supermarket and other storage information. Consumers could identify PDF417 code through a client based on B/S architecture so that a simple process of food traceability has been c...
Multilevel LDPC Codes Design for Multimedia Communication CDMA System
Directory of Open Access Journals (Sweden)
Hou Jia
2004-01-01
Full Text Available We design multilevel coding (MLC with a semi-bit interleaved coded modulation (BICM scheme based on low density parity check (LDPC codes. Different from the traditional designs, we joined the MLC and BICM together by using the Gray mapping, which is suitable to transmit the data over several equivalent channels with different code rates. To perform well at signal-to-noise ratio (SNR to be very close to the capacity of the additive white Gaussian noise (AWGN channel, random regular LDPC code and a simple semialgebra LDPC (SA-LDPC code are discussed in MLC with parallel independent decoding (PID. The numerical results demonstrate that the proposed scheme could achieve both power and bandwidth efficiency.
Energy Technology Data Exchange (ETDEWEB)
Watson, S.B.; Ford, M.R.
1980-02-01
A computer code has been developed that implements the recommendations of ICRP Committee 2 for computing limits for occupational exposure of radionuclides. The purpose of this report is to describe the various modules of the computer code and to present a description of the methods and criteria used to compute the tables published in the Committee 2 report. The computer code contains three modules of which: (1) one computes specific effective energy; (2) one calculates cumulated activity; and (3) one computes dose and the series of ICRP tables. The description of the first two modules emphasizes the new ICRP Committee 2 recommendations in computing specific effective energy and cumulated activity. For the third module, the complex criteria are discussed for calculating the tables of committed dose equivalent, weighted committed dose equivalents, annual limit of intake, and derived air concentration.
Energy Technology Data Exchange (ETDEWEB)
NONE
1997-03-01
This Manual represents Revision 5 of the user documentation for the modular code system referred to as SCALE. The history of the SCALE code system dates back to 1969 when the current Computational Physics and Engineering Division at Oak Ridge National Laboratory (ORNL) began providing the transportation package certification staff at the U. S. Atomic Energy Commission with computational support in the use of the new KENO code for performing criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976 the certification staff relied on the ORNL staff to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analyses of transportation packages. However, the certification staff learned that, with only occasional use of the codes, it was difficult to become proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the move of the certification staff to the U.S. Nuclear Regulatory Commission (NRC), the NRC staff proposed the development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system was born. This volume is part of the manual related to the control modules for the newest updated version of this computational package.
Directory of Open Access Journals (Sweden)
KIM Jong Woon
2017-01-01
In this paper, we describe a brief overview of AETIUS and provide numerical results from both AETIUS and a Monte Carlo code, MCNP5, in a deep penetration problem with small detection volumes. The results demonstrate the effectiveness and efficiency of AETIUS for such calculations.
ATHENA calculation model for the ITER-FEAT divertor cooling system. Final report with updates
Energy Technology Data Exchange (ETDEWEB)
Eriksson, John; Sjoeberg, A.; Sponton, L.L
2001-05-01
An ATHENA model of the ITER-FEAT divertor cooling system has been developed for the purpose of calculating and evaluating consequences of different thermal-hydraulic accidents as specified in the Accident Analysis Specifications for the ITER-FEAT Generic Site Safety Report. The model is able to assess situations for a variety of conceivable operational transients from small flow disturbances to more critical conditions such as total blackout caused by a loss of offsite and emergency power. The main objective for analyzing this type of scenarios is to determine margins against jeopardizing the integrity of the divertor cooling system components and pipings. The model of the divertor primary heat transport system encompasses the divertor cassettes, the port limiter systems, the pressurizer, the heat exchanger and all feed and return pipes of these components. The development was pursued according to practices and procedures outlined in the ATHENA code manuals using available modelling components such as volumes, junctions, heat structures and process controls.
Software for calculation of electric power systems parameters
Energy Technology Data Exchange (ETDEWEB)
Rodriguez, Gustavo; Aromataris, Luis; Donolo, Marcos; Hernandez, Jose; Moitre, Diego [Universidad Nacional de Rio Cuarto, Grupo de Analisis de Sistemas Electricos de Potencia (GASEP), Cordoba (Argentina)
2003-07-01
The calculation of power transmission lines parameters of electric energy has a fundamental importance in the design, construction and simulation of electric power systems. Generally, the results obtained are used in the database for the studies of power flow, short circuit and stability . This paper presents a software with graphical interface which allows the estimation of parameters of power transmission lines of electric energy. By means of a tutorial, the user is oriented step by step about the input of data and other features of the line, passing to the calculation stage when the data are correctly and completely entered. The output of the results is shown on the screen on a Pi circuit and through graphics that simulate the voltage drop for a determined load. (Author)
Prototype demonstration of radiation therapy planning code system
Energy Technology Data Exchange (ETDEWEB)
Little, R.C.; Adams, K.J.; Estes, G.P.; Hughes, L.S. III; Waters, L.S. [and others
1996-09-01
This is the final report of a one-year, Laboratory-Directed Research and Development project at the Los Alamos National Laboratory (LANL). Radiation therapy planning is the process by which a radiation oncologist plans a treatment protocol for a patient preparing to undergo radiation therapy. The objective is to develop a protocol that delivers sufficient radiation dose to the entire tumor volume, while minimizing dose to healthy tissue. Radiation therapy planning, as currently practiced in the field, suffers from inaccuracies made in modeling patient anatomy and radiation transport. This project investigated the ability to automatically model patient-specific, three-dimensional (3-D) geometries in advanced Los Alamos radiation transport codes (such as MCNP), and to efficiently generate accurate radiation dose profiles in these geometries via sophisticated physics modeling. Modem scientific visualization techniques were utilized. The long-term goal is that such a system could be used by a non-expert in a distributed computing environment to help plan the treatment protocol for any candidate radiation source. The improved accuracy offered by such a system promises increased efficacy and reduced costs for this important aspect of health care.
Energy Technology Data Exchange (ETDEWEB)
McGrail, B.P.; Bacon, D.H.
1998-02-01
Planned performance assessments for the proposed disposal of low-activity waste (LAW) glass produced from remediation of wastes stored in underground tanks at Hanford, Washington will require calculations of radionuclide release rates from the subsurface disposal facility. These calculations will be done with the aid of computer codes. The available computer codes with suitable capabilities at the time Revision 0 of this document was prepared were ranked in terms of the feature sets implemented in the code that match a set of physical, chemical, numerical, and functional capabilities needed to assess release rates from the engineered system. The needed capabilities were identified from an analysis of the important physical and chemical processes expected to affect LAW glass corrosion and the mobility of radionuclides. This analysis was repeated in this report but updated to include additional processes that have been found to be important since Revision 0 was issued and to include additional codes that have been released. The highest ranked computer code was found to be the STORM code developed at PNNL for the US Department of Energy for evaluation of arid land disposal sites.
Energy Technology Data Exchange (ETDEWEB)
Pialla, David, E-mail: david.pialla@cea.fr [Commissariat à l’Energie Atomique et aux Energies Alternatives (CEA), DEN/DM2S/STMF, 17 rue des martyrs, 38054 Grenoble Cedex 9 (France); Tenchine, Denis [Commissariat à l’Energie Atomique et aux Energies Alternatives (CEA), DEN/DM2S/STMF, 17 rue des martyrs, 38054 Grenoble Cedex 9 (France); Li, Simon [Commissariat à l’Energie Atomique et aux Energies Alternatives (CEA), DEN/DM2S/STMF, 91191 Gif-sur-Yvette Cedex (France); Gauthe, Paul; Vasile, Alfredo [Commissariat à l’Energie Atomique et aux Energies Alternatives (CEA), DEN/DER/SESI, 13108 Saint Paul Lez Durance Cedex (France); Baviere, Roland; Tauveron, Nicolas; Perdu, Fabien [Commissariat à l’Energie Atomique et aux Energies Alternatives (CEA), DEN/DM2S/STMF, 17 rue des martyrs, 38054 Grenoble Cedex 9 (France); Maas, Ludovic; Cocheme, François [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN/SEMIA/BAST, B.P. 17, 92262 Fontenay-aux-Roses Cedex (France); Huber, Klaus; Cheng, Xu [Karlsruhe Institute of Technology (KIT), Institute of Fusion and Reactor Technology (IFRT), Kaiserstraße 12, Building 07.08, 76131 Karlsruhe (Germany)
2015-08-15
Highlights: • The PHENIX natural convection test performed during the end of life tests program. • The calculation with system codes and theirs limits. • The calculation with coupling CFD and system code, which allows better prediction. • The tasks of code validation have been done in the frame of the THINS project. - Abstract: The PHENIX sodium cooled fast reactor started operation in 1973 and was shut down in 2009. Before decommissioning, an ultimate test program was designed and performed to provide valuable data for the development of future sodium cooled fast reactors, as the so-called Astrid prototype in France. Among these ultimate tests, a thermal-hydraulic Natural Convection Test (NCT) was set-up in June 2009. Starting from a reduced power state of 120 MWt, the NCT consists of a loss of the heat sink combined with a reactor scram and a primary pumps trip leading to stabilized natural circulation in the primary sodium system. The thermal-hydraulics innovative system project (THINS project), sponsored by the European Community in the frame of the 7th FP has selected this transient for validation of both stand-alone system code simulations and coupled simulations using system and CFD codes. Participants from three organizations (CEA, IRSN and KIT) have addressed this transient using different system codes (CATHARE, DYN2B and ATHLET) and CFD codes (TRIO-U and OPEN FOAM). The present paper depicts the different modeling approaches, methodologies and compares the numerical results with the available experimental data. Finally, the main lessons learned from the work performed within the THINS project on the PHENIX NCT with respect to code development and validation are summarized.
Evaluation and implementation of QR Code Identity Tag system for Healthcare in Turkey.
Uzun, Vassilya; Bilgin, Sami
2016-01-01
For this study, we designed a QR Code Identity Tag system to integrate into the Turkish healthcare system. This system provides QR code-based medical identification alerts and an in-hospital patient identification system. Every member of the medical system is assigned a unique QR Code Tag; to facilitate medical identification alerts, the QR Code Identity Tag can be worn as a bracelet or necklace or carried as an ID card. Patients must always possess the QR Code Identity bracelets within hospital grounds. These QR code bracelets link to the QR Code Identity website, where detailed information is stored; a smartphone or standalone QR code scanner can be used to scan the code. The design of this system allows authorized personnel (e.g., paramedics, firefighters, or police) to access more detailed patient information than the average smartphone user: emergency service professionals are authorized to access patient medical histories to improve the accuracy of medical treatment. In Istanbul, we tested the self-designed system with 174 participants. To analyze the QR Code Identity Tag system's usability, the participants completed the System Usability Scale questionnaire after using the system.
Speech coding technology in CDMA mobile system and its implementation
Liu, Lan; Wu, Wei
2004-03-01
The fundamental of QCELP speech coding technology is introduced. According to the features of TMS320C54X family DSP of TI Inc., the implementation approach of QCELP speech coding with fixed-point DSP (digital signal processor) is presented.
Energy Technology Data Exchange (ETDEWEB)
Rojas C, E. L. [ININ, Carretera Mexico-Toluca s/n, Ocoyoacac 52750, Estado de Mexico (Mexico)
2008-07-01
The objective of this study is to investigate the changes observed in the absorbed doses in mammary gland tissue when irradiated with a equipment of high dose rate known as Mammosite and introducing material resources contrary to the tissue that constitutes the mammary gland. The modeling study is performed with the code MCNPX, 2005 version, the equipment and the mammary gland and calculating the absorbed doses in tissue when introduced small volumes of air or calcium in the system. (Author)
System Level Evaluation of Innovative Coded MIMO-OFDM Systems for Broadcasting Digital TV
Directory of Open Access Journals (Sweden)
Y. Nasser
2008-01-01
Full Text Available Single-frequency networks (SFNs for broadcasting digital TV is a topic of theoretical and practical interest for future broadcasting systems. Although progress has been made in the characterization of its description, there are still considerable gaps in its deployment with MIMO technique. The contribution of this paper is multifold. First, we investigate the possibility of applying a space-time (ST encoder between the antennas of two sites in SFN. Then, we introduce a 3D space-time-space block code for future terrestrial digital TV in SFN architecture. The proposed 3D code is based on a double-layer structure designed for intercell and intracell space time-coded transmissions. Eventually, we propose to adapt a technique called effective exponential signal-to-noise ratio (SNR mapping (EESM to predict the bit error rate (BER at the output of the channel decoder in the MIMO systems. The EESM technique as well as the simulations results will be used to doubly check the efficiency of our 3D code. This efficiency is obtained for equal and unequal received powers whatever is the location of the receiver by adequately combining ST codes. The 3D code is then a very promising candidate for SFN architecture with MIMO transmission.
Program developed for CO{sub 2} system calculations
Energy Technology Data Exchange (ETDEWEB)
Lewis, E.; Wallace, D. [Brookhaven National Lab., Upton, NY (United States). Dept. of Applied Science; Allison, L.J. [Oak Ridge National Lab., TN (United States). Carbon Dioxide Information Analysis Center
1998-02-01
The program CO2SYS performs calculations relating parameters of the carbon dioxide (CO{sub 2}) system in seawater and freshwater. The program uses two of the four measurable parameters of the CO{sub 2} system [total alkalinity (TA), total inorganic CO{sub 2} (TCO{sub 2}), pH, and either fugacity (fCO{sub 2}) or partial pressure of CO{sub 2} (pCO{sub 2})] to calculate the other two parameters at a set of input conditions (temperature and pressure) and a set of output conditions chosen by the user. It replaces and extends the programs CO2SYSTM.EXE, FCO2TCO2.EXE, PHTCO2.EXE, and CO2BTCH.EXE, which were released in May 1995. It may be run in single-input mode or batch-input mode and has a variety of options for the various constants and parameters used. An on-screen information section is available that includes documentation on various topics relevant to the program. This program may be run on any 80 x 86 computer equipped with the DOS operating system by simply typing CO2SYS at the prompt after loading the executable file CO2SYS.EXE.
Proliferation Potential of Accelerator-Drive Systems: Feasibility Calculations
Energy Technology Data Exchange (ETDEWEB)
Riendeau, C.D.; Moses, D.L.; Olson, A.P.
1998-11-01
Accelerator-driven systems for fissile materials production have been proposed and studied since the early 1950s. Recent advances in beam power levels for small accelerators have raised the possibility that such use could be feasible for a potential proliferator. The objective of this study is to review the state of technology development for accelerator-driven spallation neutron sources and subcritical reactors. Energy and power requirements were calculated for a proton accelerator-driven neutron spallation source and subcritical reactors to produce a significant amount of fissile material--plutonium.
Energy Technology Data Exchange (ETDEWEB)
Kang, Sarah; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)
2014-10-15
These advantages make the MSR attractive and to be one of the six candidates for the Generation IV Reactor. Therefore, the researches related to the MSR are being conducted. To analyze the molten salt-cooled systems in the laboratory, this study generated the properties of molten salt using MARS-LMR. In this research, the implemented salts were Flibe (LiF-BeF{sub 2}) in a molar mixture that is 66% LiF and 34% BeF{sub 2}, respectively. Table 1 indicates the comparison of thermal properties of various coolants in nuclear power plants. Molten salt was added to the MARS-LMR code to support the analysis of Flibe-cooled systems. The molten salt includes LiF-BeF{sub 2} in a molar mixture that is 66% LiF and 34% BeF{sub 2}, respectively. MARS-LMR code for liquid metals uses the soft sphere model based on Monte Carlo calculations for particles interacting with pair potentials. Although MARS was originally intended for a safety analysis of light water reactor, Flibe properties were newly added to this code as so-called MARS-FLIBE which is applicable for Flibe-cooled systems. By using this thermodynamic property table file, the thermal hydraulic systems of Flibe can be simulated for numerical and parametric studies. In this study, the natural convection phenomena in the rectangular natural convection loop and IVR-ERVC in APR 1400 were simulated. Through the simulations in Flibe-cooled systems, the temperature distribution and mass flowrate of Flibe can be calculated and the heat transfer coefficients of Flibe in natural convection loop will be calculated by adding the related heat transfer correlations in the MARS-FLIBE code. MARS-FLIBE code will be used to predict and design of Flibe-cooled systems.
New versions of VENTURE/PC, a multigroup, multidimensional diffusion-depletion code system
Energy Technology Data Exchange (ETDEWEB)
Shapiro, A.; Huria, H.C. (Univ. of Cincinnati, OH (United States))
1990-01-01
VENTURE/PC is a microcomputer version of the BOLD VENTURE code system developed over a period of years at Oak Ridge National Laboratory (ORNL). It is a very complete and flexible multigroup, multidimensional, diffusion-depletion code system, which was developed for the Idaho National Engineering Laboratory personal computer (PC) based reactor physics and radiation shielding analysis package. The major characteristics of the code system were reported previously. Since that time, new versions of the code system have been developed. These new versions were designed to speed the convergence process, simplify the input stream, and extend the code to the state-of-the-art 32-bit microcomputers. The 16-bit version of the code is distributed by the Radiation Shielding Information Center (RSIC) at ORNL. The code has received widespread usage.
Matsumoto, Masaki; Yamanaka, Tsuneyasu; Hayakawa, Nobuhiro; Iwai, Satoshi; Sugiura, Nobuyuki
2015-03-01
This paper describes the Basic Radionuclide vAlue for Internal Dosimetry (BRAID) code, which was developed to calculate the time-dependent activity distribution in each organ and tissue characterised by the biokinetic compartmental models provided by the International Commission on Radiological Protection (ICRP). Translocation from one compartment to the next is taken to be governed by first-order kinetics, which is formulated by the first-order differential equations. In the source program of this code, the conservation equations are solved for the mass balance that describes the transfer of a radionuclide between compartments. This code is applicable to the evaluation of the radioactivity of nuclides in an organ or tissue without modification of the source program. It is also possible to handle easily the cases of the revision of the biokinetic model or the application of a uniquely defined model by a user, because this code is designed so that all information on the biokinetic model structure is imported from an input file. The sample calculations are performed with the ICRP model, and the results are compared with the analytic solutions using simple models. It is suggested that this code provides sufficient result for the dose estimation and interpretation of monitoring data. © The Author 2014. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.
Energy Technology Data Exchange (ETDEWEB)
Sienicki, J.J. [Argonne National Lab., IL (United States). Reactor Engineering Div.
1997-06-01
A fast running and simple computer code has been developed to calculate pressure loadings inside light water reactor containments/confinements under loss-of-coolant accident conditions. PACER was originally developed to calculate containment/confinement pressure and temperature time histories for loss-of-coolant accidents in Soviet-designed VVER reactors and is relevant to the activities of the US International Nuclear Safety Center. The code employs a multicompartment representation of the containment volume and is focused upon application to early time containment phenomena during and immediately following blowdown. PACER has been developed for FORTRAN 77 and earlier versions of FORTRAN. The code has been successfully compiled and executed on SUN SPARC and Hewlett-Packard HP-735 workstations provided that appropriate compiler options are specified. The code incorporates both capabilities built around a hardwired default generic VVER-440 Model V230 design as well as fairly general user-defined input. However, array dimensions are hardwired and must be changed by modifying the source code if the number of compartments/cells differs from the default number of nine. Detailed input instructions are provided as well as a description of outputs. Input files and selected output are presented for two sample problems run on both HP-735 and SUN SPARC workstations.
A Spanish version for the new ERA-EDTA coding system for primary renal disease
Directory of Open Access Journals (Sweden)
Óscar Zurriaga
2015-07-01
Conclusions: Translation and adaptation into Spanish represent an improvement that will help to introduce and use the new coding system for PKD, as it can help reducing the time devoted to coding and also the period of adaptation of health workers to the new codes.
Jabbari, N.; Hashemi-Malayeri, B.; Farajollahi, A. R.; Kazemnejad, A.; Shafaei, A.; Jabbari, S.
2007-08-01
Comparison of different Monte Carlo codes for understanding their limitations is essential to avoid systematic errors in the simulation, and to suggest further improvement for the codes. MCNP4C and EGSnrc, two Monte Carlo codes commonly used in medical physics, were compared and evaluated against electron depth-dose data and experimental results obtained using clinical radiotherapy beams. Different physical models and algorithms used in the codes give significantly different depth-dose curves. The default version of MCNP4C calculates electron depth-dose curves which are too much penetrating. The MCNP4C results agree better with the experiment if the Integrated Tiger Series-style energy-indexing algorithm is used. EGSnrc uses a class II-Condensed History (CH) scheme for the simulation of electron transport. To conclude the comparison, a timing study was performed. It was noted that EGSnrc is generally faster than MCNP4C and the use of a large number of scoring voxels dramatically slows down the MCNP4C calculation. However, the use of a large number of geometry voxels in MCNP4C only slightly affects the speed of the calculation.
Environmental performance of green building code and certification systems.
Suh, Sangwon; Tomar, Shivira; Leighton, Matthew; Kneifel, Joshua
2014-01-01
We examined the potential life-cycle environmental impact reduction of three green building code and certification (GBCC) systems: LEED, ASHRAE 189.1, and IgCC. A recently completed whole-building life cycle assessment (LCA) database of NIST was applied to a prototype building model specification by NREL. TRACI 2.0 of EPA was used for life cycle impact assessment (LCIA). The results showed that the baseline building model generates about 18 thousand metric tons CO2-equiv. of greenhouse gases (GHGs) and consumes 6 terajoule (TJ) of primary energy and 328 million liter of water over its life-cycle. Overall, GBCC-compliant building models generated 0% to 25% less environmental impacts than the baseline case (average 14% reduction). The largest reductions were associated with acidification (25%), human health-respiratory (24%), and global warming (GW) (22%), while no reductions were observed for ozone layer depletion (OD) and land use (LU). The performances of the three GBCC-compliant building models measured in life-cycle impact reduction were comparable. A sensitivity analysis showed that the comparative results were reasonably robust, although some results were relatively sensitive to the behavioral parameters, including employee transportation and purchased electricity during the occupancy phase (average sensitivity coefficients 0.26-0.29).
Evaluation of the analysis models in the ASTRA nuclear design code system
Energy Technology Data Exchange (ETDEWEB)
Cho, Nam Jin; Park, Chang Jea; Kim, Do Sam; Lee, Kyeong Taek; Kim, Jong Woon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)
2000-11-15
In the field of nuclear reactor design, main practice was the application of the improved design code systems. During the process, a lot of basis and knowledge were accumulated in processing input data, nuclear fuel reload design, production and analysis of design data, et al. However less efforts were done in the analysis of the methodology and in the development or improvement of those code systems. Recently, KEPO Nuclear Fuel Company (KNFC) developed the ASTRA (Advanced Static and Transient Reactor Analyzer) code system for the purpose of nuclear reactor design and analysis. In the code system, two group constants were generated from the CASMO-3 code system. The objective of this research is to analyze the analysis models used in the ASTRA/CASMO-3 code system. This evaluation requires indepth comprehension of the models, which is important so much as the development of the code system itself. Currently, most of the code systems used in domestic Nuclear Power Plant were imported, so it is very difficult to maintain and treat the change of the situation in the system. Therefore, the evaluation of analysis models in the ASTRA nuclear reactor design code system in very important.
On the implementation of new technology modules for fusion reactor systems codes
Energy Technology Data Exchange (ETDEWEB)
Franza, F., E-mail: fabrizio.franza@kit.edu [Institute of Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen, 76344 (Germany); Boccaccini, L.V.; Fisher, U. [Institute of Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen, 76344 (Germany); Gade, P.V.; Heller, R. [Institute for Technical Physics, Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen, 76344 (Germany)
2015-10-15
Highlights: • At KIT a new technology modules for systems code are under development. • A new algorithm for the definition of the main reactor's components is defined. • A new blanket model based on 1D neutronics analysis is described. • A new TF coil stress model based on 3D electromagnetic analysis is described. • The models were successfully benchmarked against more detailed models. - Abstract: In the frame of the pre-conceptual design of the next generation fusion power plant (DEMO), systems codes are being used from nearly 20 years. In such computational tools the main reactor components (e.g. plasma, blanket, magnets, etc.) are integrated in a unique computational algorithm and simulated by means of rather simplified mathematical models (e.g. steady state and zero dimensional models). The systems code tries to identify the main design parameters (e.g. major radius, net electrical power, toroidal field) and to make the reactor's requirements and constraints to be simultaneously accomplished. In fusion applications, requirements and constraints can be either of physics or technology kind. Concerning the latest category, at Karlsruhe Institute of Technology a new modelling activity has been recently launched aiming to develop improved models focusing on the main technology areas, such as neutronics, thermal-hydraulics, electromagnetics, structural mechanics, fuel cycle and vacuum systems. These activities started by developing: (1) a geometry model for the definition of poloidal profiles for the main reactors components, (2) a blanket model based on neutronics analyses and (3) a toroidal field coil model based on electromagnetic analysis, firstly focusing on the stresses calculations. The objective of this paper is therefore to give a short outline of these models.
Energy Technology Data Exchange (ETDEWEB)
Gonzalez, Alejandro; Milian, Daniel [Instituto Superior de Ciencias y Tecnologias Nucleares (ISCTN), La Habana (Cuba). E-mail: agg@ctn.isctn.edu.cu
2000-07-01
The task of the physical calculation of the reactor demand of the management of a great volume of information and inclose the stages for processing of data, calculations and analysis of their results. These stages are highly sensible to human mistakes, that's why is imprescindible that them undergo automatization, doing tracked all the process against mistake or unexpected result. The user-interface SYSMOD was developed over the platform IDE Delphi 3.0, visual language driven to events. It to consist in of the principal menu, which inclose between its options the preparation of the input data (File and Edit) to the pre-processors for the calculation codes of reactors. The output information may be showed in graphic and/or alphanumeric format (Data-Process). SYSMOD endures two applications for the management of the data base for the data during the preparation of the input for the pre-processors of the spectral calculation, so as for the organization, conservation and presentation for the obtained results. The carried out of the lattices and global codes, takes place from this application, over the platform MS-DOS (Run). SYSMOD regards the possibility for the debugging of the codes (Debugging), so as the benchmarks qualified to so effect (Benchmark). SYSMOD has been applied for the analysis of te WWER-440 of the first unity of Juragua Nuclear Power Plant. (author)
Energy Technology Data Exchange (ETDEWEB)
Eschbach, Romain; Riffard, Cecile; San Felice, Laurence; Marimbeau, Pierre; Venard, Christophe [CEA/DEN/DER/SPRC, CE Cadarache, St Paul-Lez-Durance (France); Laugier, Frederic [EdF, R and D, 1 av. General de Gaulle, 92131 Clamart Cedex (France); Thro, Jean-Francois [AREVA-NC, Tour Areva, 92084 Paris La Defense (France)
2008-07-01
This paper deals with the experimental qualification of the French package DARWIN for the PWR UOX fuel inventory calculation, focused on the isotopes involved in Burnup Credit (BUC) applications and decay heat calculation. The calculation - experiment (C/E-1) discrepancies are calculated with the latest European Evaluation File JEFF3 associated with the SHEM energy mesh, and compared with the previous results obtained with JEF2.2. An overview of the tendencies for the JEFF3 / JEF2 comparison is obtained on a complete range of burnup from 10 to 70 GWd/t, from the French Post Irradiation Examination (P.I.E.) database and the experimental results obtained in the framework of the international benchmark MALIBU. For the most previously unsatisfactory isotopes ({sup 236}U, {sup 242}Pu/{sup 243}Am, {sup 143}Nd, {sup 147,150,152}Sm, {sup 155}Gd and {sup 153,154,155}Eu), the (C/E-1) trends are considerably reduced. Uncertainty on decay heat calculation is systematically reduced with JEFF3, especially for short and very long cooling time, up to 40 % compared to JEF2.2 (authors)
Patel, Ronak Y; Shah, Neethu; Jackson, Andrew R; Ghosh, Rajarshi; Pawliczek, Piotr; Paithankar, Sameer; Baker, Aaron; Riehle, Kevin; Chen, Hailin; Milosavljevic, Sofia; Bizon, Chris; Rynearson, Shawn; Nelson, Tristan; Jarvik, Gail P; Rehm, Heidi L; Harrison, Steven M; Azzariti, Danielle; Powell, Bradford; Babb, Larry; Plon, Sharon E; Milosavljevic, Aleksandar
2017-01-12
The success of the clinical use of sequencing based tests (from single gene to genomes) depends on the accuracy and consistency of variant interpretation. Aiming to improve the interpretation process through practice guidelines, the American College of Medical Genetics and Genomics (ACMG) and the Association for Molecular Pathology (AMP) have published standards and guidelines for the interpretation of sequence variants. However, manual application of the guidelines is tedious and prone to human error. Web-based tools and software systems may not only address this problem but also document reasoning and supporting evidence, thus enabling transparency of evidence-based reasoning and resolution of discordant interpretations. In this report, we describe the design, implementation, and initial testing of the Clinical Genome Resource (ClinGen) Pathogenicity Calculator, a configurable system and web service for the assessment of pathogenicity of Mendelian germline sequence variants. The system allows users to enter the applicable ACMG/AMP-style evidence tags for a specific allele with links to supporting data for each tag and generate guideline-based pathogenicity assessment for the allele. Through automation and comprehensive documentation of evidence codes, the system facilitates more accurate application of the ACMG/AMP guidelines, improves standardization in variant classification, and facilitates collaborative resolution of discordances. The rules of reasoning are configurable with gene-specific or disease-specific guideline variations (e.g. cardiomyopathy-specific frequency thresholds and functional assays). The software is modular, equipped with robust application program interfaces (APIs), and available under a free open source license and as a cloud-hosted web service, thus facilitating both stand-alone use and integration with existing variant curation and interpretation systems. The Pathogenicity Calculator is accessible at http://calculator
Energy Technology Data Exchange (ETDEWEB)
Wohl, C.G.
1982-01-01
Seven calculator programs that do simple chores that arise in elementary particle physics are given. LEGENDRE evaluates the Legendre polynomial series ..sigma..a/sub n/P/sub n/(x) at a series of values of x. ASSOCIATED LEGENDRE evaluates the first-associated Legendre polynomial series ..sigma..b/sub n/P/sub n//sup 1/(x) at a series of values of x. CONFIDENCE calculates confidence levels for chi/sup 2/, Gaussian, or Poisson probability distributions. TWO BODY calculates the c.m. energy, the initial- and final-state c.m. momenta, and the extreme values of t and u for a 2-body reaction. ELLIPSE calculates coordinates of points for drawing an ellipse plot showing the kinematics of a 2-body reaction or decay. DALITZ RECTANGULAR calculates coordinates of points on the boundary of a rectangular Dalitz plot. DALITZ TRIANGULAR calculates coordinates of points on the boundary of a triangular Dalitz plot. There are short versions of CONFIDENCE (EVEN N and POISSON) that calculate confidence levels for the even-degree-of-freedom-chi/sup 2/ and the Poisson cases, and there is a short version of TWO BODY (CM) that calculates just the c.m. energy and initial-state momentum. The programs are written for the HP-97 calculator. (WHK)
Energy Technology Data Exchange (ETDEWEB)
Son, Han Seong; Song, Deok Yong [ENESYS, Taejon (Korea, Republic of); Kim, Ma Woong; Shin, Hyeong Ki; Lee, Sang Kyu; Kim, Hyun Koon [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)
2006-07-01
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.
Bobin, C; Thiam, C; Bouchard, J
2016-03-01
At LNE-LNHB, a liquid scintillation (LS) detection setup designed for Triple to Double Coincidence Ratio (TDCR) measurements is also used in the β-channel of a 4π(LS)β-γ coincidence system. This LS counter based on 3 photomultipliers was first modeled using the Monte Carlo code Geant4 to enable the simulation of optical photons produced by scintillation and Cerenkov effects. This stochastic modeling was especially designed for the calculation of double and triple coincidences between photomultipliers in TDCR measurements. In the present paper, this TDCR-Geant4 model is extended to 4π(LS)β-γ coincidence counting to enable the simulation of the efficiency-extrapolation technique by the addition of a γ-channel. This simulation tool aims at the prediction of systematic biases in activity determination due to eventual non-linearity of efficiency-extrapolation curves. First results are described in the case of the standardization (59)Fe. The variation of the γ-efficiency in the β-channel due to the Cerenkov emission is investigated in the case of the activity measurements of (54)Mn. The problem of the non-linearity between β-efficiencies is featured in the case of the efficiency tracing technique for the activity measurements of (14)C using (60)Co as a tracer. Copyright © 2015 Elsevier Ltd. All rights reserved.
Brandon, Paul R.; Harrison, George M.; Lawton, Brian E.
2013-01-01
When evaluators plan site-randomized experiments, they must conduct the appropriate statistical power analyses. These analyses are most likely to be valid when they are based on data from the jurisdictions in which the studies are to be conducted. In this method note, we provide software code, in the form of a SAS macro, for producing statistical…
1979-06-01
UCID- 16189 , January 1973. 18. Lacetera, J., Jr., Lacetera, J. M. and Schmitt, J. A., "The BRL 7600 Version of the HELP Code," USA Ballistic Research...State Coeffi- cients for High Explosives," Lawrence Livermore Laboratory, Report No. UCID- 16189 , January 1973. 18. Lacetera, J., Jr., Lacetera, J. N. and
Monte Carlo Capabilities of the SCALE Code System
Rearden, B. T.; Petrie, L. M.; Peplow, D. E.; Bekar, K. B.; Wiarda, D.; Celik, C.; Perfetti, C. M.; Ibrahim, A. M.; Hart, S. W. D.; Dunn, M. E.
2014-06-01
SCALE is a widely used suite of tools for nuclear systems modeling and simulation that provides comprehensive, verified and validated, user-friendly capabilities for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, licensees, and research institutions around the world have used SCALE for nuclear safety analysis and design. SCALE provides a "plug-and-play" framework that includes three deterministic and three Monte Carlo radiation transport solvers that can be selected based on the desired solution, including hybrid deterministic/Monte Carlo simulations. SCALE includes the latest nuclear data libraries for continuous-energy and multigroup radiation transport as well as activation, depletion, and decay calculations. SCALE's graphical user interfaces assist with accurate system modeling, visualization, and convenient access to desired results. SCALE 6.2, to be released in 2014, will provide several new capabilities and significant improvements in many existing features, especially with expanded continuous-energy Monte Carlo capabilities for criticality safety, shielding, depletion, and sensitivity and uncertainty analysis. An overview of the Monte Carlo capabilities of SCALE is provided here, with emphasis on new features for SCALE 6.2.
Energy Technology Data Exchange (ETDEWEB)
Marie, S. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France)], E-mail: stephane.marie@cea.fr; Chapuliot, S.; Kayser, Y. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France); Lacire, M.H. [CEA Saclay, DEN/DDIN, 91191 Gif sur Yvette Cedex (France); Drubay, B. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France); Barthelet, B. [EDF/EPN, Site Cap Ampere, 1 place Pleyel 93207, Saint Denis Cedex 1 (France); Le Delliou, P. [EDF Pole Industrie-Division R and D, Site des Renardieres, Route de Sens, Ecuelles, 77250 Moret sur Loing Cedex (France); Rougier, V. [EDF/UTO SIS/GAM, 6, avenue Montaigne, 93192 Noisy le Grand (France); Naudin, C. [EDF/SEPTEN, 12-14, avenue Dutrievoz, 69628 Villeurbanne Cedex (France); Gilles, P.; Triay, M. [AREVA ANP, Tour AREVA, 92084 Paris La Defense Cedex 16 (France)
2007-10-15
French nuclear codes include flaw assessment procedures: the RSE-M Code 'Rules for In-service Inspection of Nuclear Power Plant Components' and the RCC-MR code 'Design and Construction rules for mechanical components of FBR nuclear islands and high temperature applications'. An important effort of development of these analytical methods has been made for the last 10 years in the frame of a collaboration between CEA, EDF and AREVA-NP, and in the frame of R and D actions involving CEA and IRSN. These activities have led to a unification of the common methods of the two codes. The calculation of fracture mechanics parameters, and in particular the stress intensity factor K{sub I} and the J integral, has been widely developed for industrial configurations. All the developments have been integrated in the 2005 edition of RSE-M and in the 2007 edition of RCC-MR. This series of articles is composed of 5 parts: the first part presents an overview of the methods proposed in the RCC-MR and RSE-M codes. Parts II-IV provide compendia for specific components. The geometries are plates (part II), pipes (part III) and elbows (part IV). Finally, part V presents the validation elements of the methods, with details on the process followed for the development and evaluation of the accuracy of the proposed analytical methods. This second article in the series presents all details for the stress intensity factor and J calculations for cracked plates. General data applicable for all defect geometries are first presented, and then, available defect geometries where compendia for K{sub I} and {sigma}{sub ref} calculation are provided are given.
Energy Technology Data Exchange (ETDEWEB)
Marie, S. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France)], E-mail: stephane.marie@cea.fr; Chapuliot, S.; Kayser, Y. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France); Lacire, M.H. [CEA Saclay, DEN/DDIN, 91191 Gif sur Yvette Cedex (France); Drubay, B. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France); Barthelet, B. [EDF/EPN, Site Cap Ampere, 1 place Pleyel 93207, Saint Denis Cedex 1 (France); Le Delliou, P. [EDF Pole Industrie-Division R and D, Site des Renardieres, Route de Sens, Ecuelles, 77250 Moret sur Loing Cedex (France); Rougier, V. [EDF/UTO SIS/GAM, 6, avenue Montaigne, 93192 Noisy le Grand (France); Naudin, C. [EDF/SEPTEN, 12-14, Avenue Dutrievoz, 69628 Villeurbanne Cedex (France); Gilles, P.; Triay, M. [AREVA ANP, Tour AREVA, 92084 Paris La Defense Cedex 16 (France)
2007-10-15
French nuclear codes include flaw assessment procedures: the RSE-M Code 'Rules for In-service Inspection of Nuclear Power Plant Components' and the RCC-MR code 'Design and Construction rules for mechanical components of FBR nuclear islands and high-temperature applications'. An important effort of development of these analytical methods has been made for the last 10 years in the frame of collaboration between CEA, EDF and AREVA-NP, and in the frame of R and D actions involving CEA and IRSN. These activities have led to a unification of the common methods of the two codes. The calculation of fracture mechanics parameters, and in particular the stress-intensity factor K{sub I} and the J integral, has been widely developed for industrial configurations. All the developments have been integrated in the 2005 edition of the RSE-M and in the 2007 edition of the RCC-MR. This series of articles is composed of 5 parts: the first part presents an overview of the methods proposed in the RCC-MR and RSE-M codes. Parts II-IV provide compendia for specific components. The geometries are plates (part II), pipes (part III) and elbows (part IV). Part V presents validation, with details on the accuracy of the proposed analytical method. This third part in the series presents details of the stress intensity factor and J calculations for cracked pipes. General data applicable for all defect geometries are first presented, and then, compendia for K{sub I} and {sigma}{sub ref} calculations are provided for specific cases.
Energy Technology Data Exchange (ETDEWEB)
Marie, S. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France)], E-mail: stephane.marie@cea.fr; Chapuliot, S.; Kayser, Y. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France); Lacire, M.H. [CEA Saclay, DEN/DDIN, 91191 Gif sur Yvette Cedex (France); Drubay, B. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France); Barthelet, B. [EDF/EPN, Site Cap Ampere, 1 place Pleyel 93207, Saint Denis Cedex 1 (France); Le Delliou, P. [EDF Pole Industrie-Division R and D, Site des Renardieres, Route de Sens, Ecuelles, 77250 Moret sur Loing Cedex (France); Rougier, V. [EDF/UTO SIS/GAM, 6, avenue Montaigne, 93192 Noisy le Grand (France); Naudin, C. [EDF/SEPTEN, 12-14, avenue Dutrievoz, 69628 Villeurbanne Cedex (France); Gilles, P.; Triay, M. [AREVA ANP, Tour AREVA, 92084 Paris La Defense Cedex 16 (France)
2007-10-15
Two French nuclear codes include flaw assessment procedures: the RSE-M code 'Rules for In-service Inspection of Nuclear Power Plant Components' and the RCC-MR code 'Design and Construction rules for mechanical components of FBR nuclear islands and high temperature applications'. An important effort of development of these analytical methods has been made for the last 10 years in the frame of a collaboration between CEA, EDF and AREVA-NP, and in the frame of R and D actions involving CEA and IRSN. These activities have led to a unification of the common methods of the two codes. The calculation of fracture mechanics parameters, and in particular the stress intensity factor K{sub I} and the J integral, has been widely developed for industrial configurations. All developments have been integrated in the 2005 edition of RSE-M and in the 2007 edition of RCC-MR. This series of articles is composed of five parts: this first one presents an overview of the methods proposed in the RCC-MR and RSE-M codes. Parts II-IV provide compendia for specific components: plates (part II), pipes (part III) and elbows (part IV). Finally, part V presents the validation elements of the methods, with details on the process followed for their development and on evaluation of the accuracy of the proposed analytical methods. This first article of the series presents an overview of the calculation of K{sub I} and J in these two codes and describes briefly the defect assessment analyses. Specific details in the Appendix A16 of RCC-MR (LBB procedure and creep analyses) are also introduced in this article.
Energy Technology Data Exchange (ETDEWEB)
Lischke, W.; Vandreier, B. [Univ. for Applied Sciences, Zittau/Goerlitz (Germany). Dept. of Nuclear Technology
1997-12-31
At the University for Applied Sciences Zittau/Goerlitz (FH) calculations for the verification of the ATHLET-code for reactors of type VVER are carried out since 1991, sponsored by the German Ministry for Education, Science and Technology (BMBF). The special features of these reactors in comparison to reactors of western countries are characterized by the duct route of reactor coolant pipes and the horizontal steam generators. Because of these special features, a check of validity of the ATHLET-models is necessary. For further verification of the ATHLET-code the post test calculation of the experiment SBL-22 (Small break loss-of-coolant test) realized at the finnish facility PACTEL was carried out. The experiment served for the examination of the natural circulation behaviour of the loop over a continuous range of primary side water inventory. 5 refs.
High-Speed Turbo-TCM-Coded Orthogonal Frequency-Division Multiplexing Ultra-Wideband Systems
Directory of Open Access Journals (Sweden)
2006-01-01
Full Text Available One of the UWB proposals in the IEEE P802.15 WPAN project is to use a multiband orthogonal frequency-division multiplexing (OFDM system and punctured convolutional codes for UWB channels supporting a data rate up to 480 Mbps. In this paper, we improve the proposed system using turbo TCM with QAM constellation for higher data rate transmission. We construct a punctured parity-concatenated trellis codes, in which a TCM code is used as the inner code and a simple parity-check code is employed as the outer code. The result shows that the system can offer a much higher spectral efficiency, for example, 1.2 Gbps, which is 2.5 times higher than the proposed system. We identify several essential requirements to achieve the high rate transmission, for example, frequency and time diversity and multilevel error protection. Results are confirmed by density evolution.
Energy Technology Data Exchange (ETDEWEB)
Lee, Won Jae; Chung, Bub Dong; Jeong, Jae Jun; Ha, Kwi Seok [Korea Atomic Energy Research Institute, Taejon (Korea)
1998-06-01
A multi-dimensional realistic thermal-hydraulic system analysis code, MARS version 1.3 has been developed. Main purpose of MARS 1.3 development is to have the realistic analysis capability of transient two-phase thermal-hydraulics of Pressurized Water Reactors (PWRs) especially during Large Break Loss of Coolant Accidents (LBLOCAs) where the multi-dimensional phenomena domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, three-dimensional (3D) reactor vessel analysis code, and RELAP5/MOD3.2.1.2, one-dimensional (1D) reactor system analysis code., Developmental requirements for MARS are chosen not only to best utilize the existing capability of the codes but also to have the enhanced capability in code maintenance, user accessibility, user friendliness, code portability, code readability, and code flexibility. For the maintenance of existing codes capability and the enhancement of code maintenance capability, user accessibility and user friendliness, MARS has been unified to be a single code consisting of 1D module (RELAP5) and 3D module (COBRA-TF). This is realized by implicitly integrating the system pressure matrix equations of hydrodynamic models and solving them simultaneously, by modifying the 1D/3D calculation sequence operable under a single Central Processor Unit (CPU) and by unifying the input structure and the light water property routines of both modules. In addition, the code structure of 1D module is completely restructured using the modular data structure of standard FORTRAN 90, which greatly improves the code maintenance capability, readability and portability. For the code flexibility, a dynamic memory management scheme is applied in both modules. MARS 1.3 now runs on PC/Windows and HP/UNIX platforms having a single CPU, and users have the options to select the 3D module to model the 3D thermal-hydraulics in the reactor vessel or other
Directory of Open Access Journals (Sweden)
Yettou Leila
2015-01-01
Full Text Available In this study, a new version EMPIRE 3.2 code was used in the cross section calculations of (n,p reactions and in the calculation of proton emission spectra produced by (n,xp reactions. Exciton model predictions combined with the Kalbach angular distribution systematics were used and some parameters such as those of mean free path, cluster emission in terms of Iwamoto-Harada model, optical model potentials of Morillon for neutrons and protons in the energy range up to 20 MeV, level density for spherical nuclei of Gilbert-Cameron model and width fluctuation correction in terms of compound nucleus have been investigated our calculations. The excitation functions and the proton emission spectra for 32,34S nuclei were calculated, discussed and found in good agreement with available experimental data.
Energy Technology Data Exchange (ETDEWEB)
Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.
1998-03-01
This document is the User`s Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code`s capabilities and limitations; Chapter 2 describes the code`s structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARC and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs.
Energy Technology Data Exchange (ETDEWEB)
Kobori, Hikaru [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hiwatari, Ryoji [Central Research Institute of Electric Power Industry, Tokyo (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan)
2016-11-01
Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO{sub 2} emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.
Garcia, F.; Mesa, J.; Arruda-Neto, J. D. T.; Helene, O.; Vanin, V.; Milian, F.; Deppman, A.; Rodrigues, T. E.; Rodriguez, O.
2007-03-01
The code STATFLUX, implementing a new and simple statistical procedure for the calculation of transfer coefficients in radionuclide transport to animals and plants, is proposed. The method is based on the general multiple-compartment model, which uses a system of linear equations involving geometrical volume considerations. Flow parameters were estimated by employing two different least-squares procedures: Derivative and Gauss-Marquardt methods, with the available experimental data of radionuclide concentrations as the input functions of time. The solution of the inverse problem, which relates a given set of flow parameter with the time evolution of concentration functions, is achieved via a Monte Carlo simulation procedure. Program summaryTitle of program:STATFLUX Catalogue identifier:ADYS_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/ADYS_v1_0 Program obtainable from: CPC Program Library, Queen's University of Belfast, N. Ireland Licensing provisions: none Computer for which the program is designed and others on which it has been tested:Micro-computer with Intel Pentium III, 3.0 GHz Installation:Laboratory of Linear Accelerator, Department of Experimental Physics, University of São Paulo, Brazil Operating system:Windows 2000 and Windows XP Programming language used:Fortran-77 as implemented in Microsoft Fortran 4.0. NOTE: Microsoft Fortran includes non-standard features which are used in this program. Standard Fortran compilers such as, g77, f77, ifort and NAG95, are not able to compile the code and therefore it has not been possible for the CPC Program Library to test the program. Memory required to execute with typical data:8 Mbytes of RAM memory and 100 MB of Hard disk memory No. of bits in a word:16 No. of lines in distributed program, including test data, etc.:6912 No. of bytes in distributed program, including test data, etc.:229 541 Distribution format:tar.gz Nature of the physical problem:The investigation of transport mechanisms for
Panettieri, Vanessa; Barsoum, Pierre; Westermark, Mathias; Brualla, Lorenzo; Lax, Ingmar
2009-10-01
In tangential beam treatments accurate dose calculation of the absorbed dose in the build-up region is of major importance, in particular when the target has superficial extension close to the skin. In most analytical treatment planning systems (TPSs) calculations depend on the experimental measurements introduced by the user in which accuracy might be limited by the type of detector employed to perform them. To quantify the discrepancy between analytically calculated and delivered dose in the build-up region, near the skin of a patient, independent Monte Carlo (MC) simulations using the penelope code were performed. Dose distributions obtained with MC simulations were compared with those given by the Pencil Beam Convolution (PBC) algorithm and the Analytical Anisotropic Algorithm (AAA) implemented in the commercial TPS Eclipse. A cylindrical phantom was used to approximate the breast contour of a patient for MC simulations and the TPS. Calculations of the absorbed doses were performed for 6 and 18MV beams for four different angles of incidence: 15 degrees , 30 degrees , 45 degrees and 75 degrees and different field sizes: 3x3cm(2), 10x10cm(2) and 40x40cm(2). Absorbed doses along the phantom central axis were obtained with both the PBC algorithm and the AAA and compared to those estimated by the MC simulations. Additionally, a breast patient case was calculated with two opposed 6MV photon beams using all the aforementioned analytical and stochastic algorithms. For the 6MV photon beam in the phantom case, both the PBC algorithm and the AAA tend to underestimate the absorbed dose in the build-up region in comparison to MC results. These differences are clinically irrelevant and are included in a 1mm range. This tendency is also confirmed in the breast patient case. For the 18MV beam the PBC algorithm underestimates the absorbed dose with respect to the AAA. In comparison to MC simulations the PBC algorithm tends to underestimate the dose after the first 2-3mm of
Development of A Monte Carlo Radiation Transport Code System For HEDS: Status Update
Townsend, Lawrence W.; Gabriel, Tony A.; Miller, Thomas M.
2003-01-01
Modifications of the Monte Carlo radiation transport code HETC are underway to extend the code to include transport of energetic heavy ions, such as are found in the galactic cosmic ray spectrum in space. The new HETC code will be available for use in radiation shielding applications associated with missions, such as the proposed manned mission to Mars. In this work the current status of code modification is described. Methods used to develop the required nuclear reaction models, including total, elastic and nuclear breakup processes, and their associated databases are also presented. Finally, plans for future work on the extended HETC code system and for its validation are described.
Performance Analysis of a CDMA VSAT System With Convoltional and Reed-Solomon Coding
Yigit, Ugur
2002-09-01
The purpose of this thesis is to model a satellite communication system with VSATs, using Spread Spectrum CDMA methods and Forward Error Correction (FEC), Walsh codes and PN sequences are used to generate a CDMA system and FEC is used to further improve the performance. Convolutional and block coding methods are examined and the results are obtained for each different case, including concatenated use of the codes, The performance of the system is given in terms of Bit Error Rate (BER), As observed from the results, the performance is mainly affected by the number of users and the code rates,
Marconi, F.; Yaeger, L.
1976-01-01
A numerical procedure was developed to compute the inviscid super/hypersonic flow field about complex vehicle geometries accurately and efficiently. A second-order accurate finite difference scheme is used to integrate the three-dimensional Euler equations in regions of continuous flow, while all shock waves are computed as discontinuities via the Rankine-Hugoniot jump conditions. Conformal mappings are used to develop a computational grid. The effects of blunt nose entropy layers are computed in detail. Real gas effects for equilibrium air are included using curve fits of Mollier charts. Typical calculated results for shuttle orbiter, hypersonic transport, and supersonic aircraft configurations are included to demonstrate the usefulness of this tool.
Energy Technology Data Exchange (ETDEWEB)
Marie, S. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France)], E-mail: stephane.marie@cea.fr; Chapuliot, S.; Kayser, Y. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France); Lacire, M.H. [CEA Saclay, DEN/DDIN, 91191 Gif sur Yvette Cedex (France); Drubay, B. [CEA Saclay, DEN/DM2S/SEMT/LISN, 91191 Gif sur Yvette Cedex (France); Barthelet, B. [EDF/EPN, Site Cap Ampere, 1 place Pleyel 93207, Saint Denis Cedex 1 (France); Le Delliou, P. [EDF Pole Industrie-Division R and D, Site des Renardieres, Route de Sens, Ecuelles, 77250 Moret sur Loing Cedex (France); Rougier, V. [EDF/UTO, SIS/GAM, 6, avenue Montaigne, 93192 Noisy le Grand (France); Naudin, C. [EDF/SEPTEN, 12-14, avenue Dutrievoz, 69628 Villeurbanne Cedex (France); Gilles, P.; Triay, M. [AREVA ANP, Tour AREVA, 92084 Paris La Defense Cedex 16 (France)
2007-10-15
French nuclear codes include flaw assessment procedures: the RSE-M Code 'Rules for In-service Inspection of Nuclear Power Plant Components' and the RCC-MR code 'Design and Construction Rules for Mechanical Components of FBR Nuclear Islands and High Temperature Applications'. Development of analytical methods has been made for the last 10 years in the framework of a collaboration between CEA, EDF and AREVA-NP, and by R and D actions involving CEA and IRSN. These activities have led to a unification of the common methods of the two codes. The calculation of fracture mechanics parameters, in particular the stress intensity factor K{sub I} and the J integral, has been widely developed for industrial configurations. All the developments have been integrated in the 2005 edition of RSE-M and in 2007 edition of RCC-MR. This series of articles consists of 5 parts: the first part presents an overview of the methods proposed in the RCC-MR and RSE-M codes. Parts II-IV provide the compendia for specific components. The geometries are plates (part II), pipes (part III) and elbows (part IV). This part presents validation of the methods, with details on the process followed for their development and of the evaluation accuracy of the proposed analytical methods.
An EGS4 user code with voxel geometry and a voxel phantom generation system
Energy Technology Data Exchange (ETDEWEB)
Funabiki, J.; Terabe, M. [Mitsubishi Research Institute, Inc., Tokyo (Japan); Zankl, M. [GSF, Neuherberg, Oberschleissheim (Germany); Koga, S. [Fujita Health Univ., Toyoake, Aichi (Japan); Saito, K. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2000-12-01
An EGS4 (Electron Gamma Shower Version 4) user code with voxel geometry (the UCPIXEL code) has been developed in order to make accurate dose evaluation by using human voxel phantoms. The voxel data have a format devised by GSF. This format can compress so large amount of high-resolution voxel data that required memory for computation is greatly reduced. UCPIXEL can treat 8 basic irradiation geometries (AP, PA, RLAT, LLAT, ROT, ISO, AB, BA) and cylindrical pseudo-environmental radiation source with arbitrary size, energy and directional distribution. In addition, UCPIXEL can model contaminated soil with arbitrary area and depth under a phantom and radiation from the soil. By using a post-processor, effective dose equivalent, effective dose and organ doses can be evaluated from the output of UCPIXEL. Preliminary results of effective dose calculated by using UCPIXEL for a Japanese voxel phantom are demonstrated and compared with the previous results for MIRD-type phantoms. We have also developed an intelligent system which automatically constructs a voxel phantom from CT data. We introduce this system and preliminary results are shown. (author)
Directory of Open Access Journals (Sweden)
Jingang Liang
2016-06-01
Full Text Available Because of prohibitive data storage requirements in large-scale simulations, the memory problem is an obstacle for Monte Carlo (MC codes in accomplishing pin-wise three-dimensional (3D full-core calculations, particularly for whole-core depletion analyses. Various kinds of data are evaluated and quantificational total memory requirements are analyzed based on the Reactor Monte Carlo (RMC code, showing that tally data, material data, and isotope densities in depletion are three major parts of memory storage. The domain decomposition method is investigated as a means of saving memory, by dividing spatial geometry into domains that are simulated separately by parallel processors. For the validity of particle tracking during transport simulations, particles need to be communicated between domains. In consideration of efficiency, an asynchronous particle communication algorithm is designed and implemented. Furthermore, we couple the domain decomposition method with MC burnup process, under a strategy of utilizing consistent domain partition in both transport and depletion modules. A numerical test of 3D full-core burnup calculations is carried out, indicating that the RMC code, with the domain decomposition method, is capable of pin-wise full-core burnup calculations with millions of depletion regions.
Energy Technology Data Exchange (ETDEWEB)
Panka, Istvan; Hegyi, Gyoergy; Maraczy, Csaba; Temesvari, Emese [Hungarian Academy of Sciences, Budapest (Hungary). Reactor Analysis Dept.
2017-11-15
The best-estimate KARATE code system has been widely used for core design calculations and simulations of slow transients of VVER reactors. Recently there has been an increasing need for assessing the uncertainties of such calculations by propagating the basic input uncertainties of the models through the full calculation chain. In order to determine the uncertainties of quantities of interest during the burnup, the statistical version of the KARATE code system has been elaborated. In the first part of the paper, the main features of the new code system are discussed. The applied statistical method is based on Monte-Carlo sampling of the considered input data taking into account mainly the covariance matrices of the cross sections and/or the technological uncertainties. In the second part of the paper, only the uncertainties of cross sections are considered and an equilibrium cycle related to a VVER-440 type reactor is investigated. The burnup dependence of the uncertainties of some safety related parameters (e.g. critical boron concentration, rod worth, feedback coefficients, assembly-wise radial power and burnup distribution) are discussed and compared to the recently used limits.
Rotated Walsh-Hadamard Spreading with Robust Channel Estimation for a Coded MC-CDMA System
Directory of Open Access Journals (Sweden)
Raulefs Ronald
2004-01-01
Full Text Available We investigate rotated Walsh-Hadamard spreading matrices for a broadband MC-CDMA system with robust channel estimation in the synchronous downlink. The similarities between rotated spreading and signal space diversity are outlined. In a multiuser MC-CDMA system, possible performance improvements are based on the chosen detector, the channel code, and its Hamming distance. By applying rotated spreading in comparison to a standard Walsh-Hadamard spreading code, a higher throughput can be achieved. As combining the channel code and the spreading code forms a concatenated code, the overall minimum Hamming distance of the concatenated code increases. This asymptotically results in an improvement of the bit error rate for high signal-to-noise ratio. Higher convolutional channel code rates are mostly generated by puncturing good low-rate channel codes. The overall Hamming distance decreases significantly for the punctured channel codes. Higher channel code rates are favorable for MC-CDMA, as MC-CDMA utilizes diversity more efficiently compared to pure OFDMA. The application of rotated spreading in an MC-CDMA system allows exploiting diversity even further. We demonstrate that the rotated spreading gain is still present for a robust pilot-aided channel estimator. In a well-designed system, rotated spreading extends the performance by using a maximum likelihood detector with robust channel estimation at the receiver by about 1 dB.
Channel coding for underwater acoustic single-carrier CDMA communication system
Liu, Lanjun; Zhang, Yonglei; Zhang, Pengcheng; Zhou, Lin; Niu, Jiong
2017-01-01
CDMA is an effective multiple access protocol for underwater acoustic networks, and channel coding can effectively reduce the bit error rate (BER) of the underwater acoustic communication system. For the requirements of underwater acoustic mobile networks based on CDMA, an underwater acoustic single-carrier CDMA communication system (UWA/SCCDMA) based on the direct-sequence spread spectrum is proposed, and its channel coding scheme is studied based on convolution, RA, Turbo and LDPC coding respectively. The implementation steps of the Viterbi algorithm of convolutional coding, BP and minimum sum algorithms of RA coding, Log-MAP and SOVA algorithms of Turbo coding, and sum-product algorithm of LDPC coding are given. An UWA/SCCDMA simulation system based on Matlab is designed. Simulation results show that the UWA/SCCDMA based on RA, Turbo and LDPC coding have good performance such that the communication BER is all less than 10-6 in the underwater acoustic channel with low signal to noise ratio (SNR) from -12 dB to -10dB, which is about 2 orders of magnitude lower than that of the convolutional coding. The system based on Turbo coding with Log-MAP algorithm has the best performance.
Directory of Open Access Journals (Sweden)
Machmud Roby Alhamidi
2013-12-01
Full Text Available One way to increase the performance of Orthogonal Frequency Division Multiplexing System (OFDM system is by adding a channel coding (error correction code in order to detect and correct errors that occur when sending data.At communication of acoustic underwater channel coding is required because of the characteristics of the channel bottom water is much different compared with the air channel and errors are likely to occur.In this research it was made simulation of acoustic underwater communication system with OFDM applied channel codingin which using Hamming code (7,4 and Hamming code (15,11 that is able to correct one error and detect two errors then BCH code capable to correct two errors for BCH (15,7 and correct 9 errors forBCH (127,64 and Reed Solomon code able to correct two errors for RS (15,11 and correct 8 errors for RS (31,15. Results of the study confirm the better performance when system usesOFDM with BCH Code (127.64 than other codes that are used, starting from 1 decibel (dB to 3 dB for the performance of BER as10 -3 on Additive Gaussian White Noise (AWGN channel while at the multipath channel, the performance of Bit Error Rate (BER got better result on 1 dB up to 8 dB for BER performance as10 -3. Keyword: Underwater, Orthogonal Frequency Division Multiplexing (OFDM, channel coding
Energy Technology Data Exchange (ETDEWEB)
Nishio, Gunji; Watanabe, Kouji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kouno, Kouji; Yamazaki, Noboru; Mukaide, Shigeo; Yoshioka, Itsuo
1998-03-01
The CELVA-1D computer code was developed to evaluate the confinement of radioactive materials during postulated fire and explosion in a cell of nuclear fuel reprocessing plants. The CELVA-1D code calculates a response of temperature, pressure, flow velocity of fluid in an air-ventilation system of the plants by one-dimensional thermofluid analysis and calculates an ability to confine radioactive aerosol particles by transport, deposition, and HEPA filtration. The mathematical models in CELVA-1D were verified by comparison of the calculation with the result of JAERI`s demonstration tests simulating hypothetical fire and explosion accidents in the cell. (author)
2009-10-01
Becker-Kistiakowsky-Wilson (BKW) 2. Jacobs-Cowperthwaite-Zwisler ( JCZ ) 3. Hayes 4. Davis 5. Williamsburg 6. JWL 7. HOM the JWL and HOM EOS...have often been used in hydrocode/CFD simulations. On the other hand, the BKW and JCZ equations remain the EOS of choice in chemical equilibrium code...development for condensed explosives. Various databases have been constructed for the BKW and JCZ equations of state. These include: BKWC
Energy Technology Data Exchange (ETDEWEB)
Grazevicius, Audrius; Kaliatka, Algirdas [Lithuanian Energy Institute, Kaunas (Lithuania). Lab. of Nuclear Installation Safety
2017-07-15
The main functions of spent fuel pools are to remove the residual heat from spent fuel assemblies and to perform the function of biological shielding. In the case of loss of heat removal from spent fuel pool, the fuel rods and pool water temperatures would increase continuously. After the saturated temperature is reached, due to evaporation of water the pool water level would drop, eventually causing the uncover of spent fuel assemblies, fuel overheating and fuel rods failure. This paper presents an analysis of loss of heat removal accident in spent fuel pool of BWR 4 and a comparison of two different modelling approaches. The one-dimensional system thermal-hydraulic computer code RELAP5 and CFD tool ANSYS Fluent were used for the analysis. The results are similar, but the local effects cannot be simulated using a one-dimensional code. The ANSYS Fluent calculation demonstrated that this three-dimensional treatment allows to avoid the need for many one-dimensional modelling assumptions in the pool modelling and enables to reduce the uncertainties associated with natural circulation flow calculation.
Calculation methods for SPF for heat pump systems for comparison, system choice and dimensioning
Energy Technology Data Exchange (ETDEWEB)
Nordman, Roger; Andersson, Kajsa; Axell, Monica; Lindahl, Markus
2010-09-15
In this project, results from field measurements of heat pumps have been collected and summarised. Also existing calculation methods have been compared and summarised. Analyses have been made on how the field measurements compare to existing calculation models for heat pumps Seasonal Performance Factor (SPF), and what deviations may depend on. Recommendations for new calculation models are proposed, which include combined systems (e.g. solar - HP), capacity controlled heat pumps and combined DHW and heating operation
Impact of advanced systems on LMFBR accident analysis code development
Energy Technology Data Exchange (ETDEWEB)
Dunn, F. E.; Kyser, J. M.
1979-01-01
In order to investigate the ability of an advanced computer, using currently available software, to handle large LMFBR accident analysis codes, the SAS3D code has been run on the NCAR CRAY-1. SAS3D is a large code (56,000 Fortran cards) using many different physical models and numerical algorithms, no one of which dominates the computing time. Even though SAS3D was developed on IBM computers, remarkably little effort was required to run it on the CRAY-1. Making limited use of the CRAY-1 vector capabilities, it runs a factor of 2.5 to 4 times faster on the NCAR CRAY-1 than on the ANL IBM 370-195. With minor modifications, an additional 20 to 30% speed improvement on the CRAY-1 is achieved. In the current process of completely re-writing SAS3D to make SAS4A, much of the coding is being vectorized for the CRAY-1 without sacrificing IBM, CDC 7600, or UNIVAC performance and portability. An initial SAS4A test case runs a factor of 7.1 faster on the CRAY-1 than on the IBM 370-195.
Internal Corrosion Control of Water Supply Systems Code of Practice
This Code of Practice is part of a series of publications by the IWA Specialist Group on Metals and Related Substances in Drinking Water. It complements the following IWA Specialist Group publications: 1. Best Practice Guide on the Control of Lead in Drinking Water 2. Best Prac...
Implications of Sepedi/English code switching for ASR systems
CSIR Research Space (South Africa)
Modipa, TI
2013-12-01
Full Text Available to the dominant language can become particularly frequent. We analyse one such scenario: Sepedi spoken in South Africa, where English is the dominant language; and determine the frequency and mechanisms of code switching through the analysis of radio broadcasts...
Directory of Open Access Journals (Sweden)
Surbhi Sharma
2011-06-01
Full Text Available Irregular low-density parity-check (LDPC codes have been found to show exceptionally good performance for single antenna systems over a wide class of channels. In this paper, the performance of LDPC codes with multiple antenna systems is investigated in flat Rayleigh and Rician fading channels for different modulation schemes. The focus of attention is mainly on the concatenation of irregular LDPC codes with complex orthogonal space-time codes. Iterative decoding is carried out with a density evolution method that sets a threshold above which the code performs well. For the proposed concatenated system, the simulation results show that the QAM technique achieves a higher coding gain of 8.8 dB and 3.2 dB over the QPSK technique in Rician (LOS and Rayleigh (NLOS faded environments respectively.
Energy Technology Data Exchange (ETDEWEB)
Galindo G, I. F.; Vergara del C, J. A.; Galvan A, S. J. [Instituto Nacional de Electricidad y Energias Limpias, Reforma 113, Col. Palmira, 62490 Cuernavaca, Morelos (Mexico); Tijerina S, F., E-mail: francisco.tijerina@cfe.gob.mx [CFE, Central Nucleoelectrica Laguna Verde, Carretera Federal Cardel-Nautla Km 42.5, 91476 Municipio Alto Lucero, Veracruz (Mexico)
2016-09-15
The use of specialized codes to estimate the radiation dose projection to an emergency postulated event at a nuclear power plant requires that certain plant data be available according to the event being simulated. The calculation of the possible radiological release is the critical activity to carry out the emergency actions. However, not all of the plant data required are obtained directly from the plant but need to be calculated. In this paper we present a computational tool that calculates the plant data required to use the radiological dose estimation codes. The tool provides the required information when there is a gas emergency venting event in the primary containment atmosphere, whether well or dry well and also calculates the time in which the spent fuel pool would be discovered in the event of a leak of water on some of the walls or floor of the pool. The tool developed has mathematical models for the processes involved such as: compressible flow in pipes considering area change and for constant area, taking into account the effects of friction and for the case of the spent fuel pool hydraulic models to calculate the time in which a container is emptied. The models implemented in the tool are validated with data from the literature for simulated cases. The results with the tool are very similar to those of reference. This tool will also be very supportive so that in postulated emergency cases can use the radiological dose estimation codes to adequately and efficiently determine the actions to be taken in a way that affects as little as possible. (Author)
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.
Moral, Cristian; de Antonio, Angelica; Ferre, Xavier; Lara, Graciela
2015-01-01
Introduction: In this article we propose a qualitative analysis tool--a coding system--that can support the formalisation of the information-seeking process in a specific field: research in computer science. Method: In order to elaborate the coding system, we have conducted a set of qualitative studies, more specifically a focus group and some…
Proposing a Web-Based Tutorial System to Teach Malay Language Braille Code to the Sighted
Wah, Lee Lay; Keong, Foo Kok
2010-01-01
The "e-KodBrailleBM Tutorial System" is a web-based tutorial system which is specially designed to teach, facilitate and support the learning of Malay Language Braille Code to individuals who are sighted. The targeted group includes special education teachers, pre-service teachers, and parents. Learning Braille code involves memorisation…
Chabert, I.; Barat, E.; Dautremer, T.; Montagu, T.; Agelou, M.; Croc de Suray, A.; Garcia-Hernandez, J. C.; Gempp, S.; Benkreira, M.; de Carlan, L.; Lazaro, D.
2016-07-01
This work aims at developing a generic virtual source model (VSM) preserving all existing correlations between variables stored in a Monte Carlo pre-computed phase space (PS) file, for dose calculation and high-resolution portal image prediction. The reference PS file was calculated using the PENELOPE code, after the flattening filter (FF) of an Elekta Synergy 6 MV photon beam. Each particle was represented in a mobile coordinate system by its radial position (r s ) in the PS plane, its energy (E), and its polar and azimuthal angles (φ d and θ d ), describing the particle deviation compared to its initial direction after bremsstrahlung, and the deviation orientation. Three sub-sources were created by sorting out particles according to their last interaction location (target, primary collimator or FF). For each sub-source, 4D correlated-histograms were built by storing E, r s , φ d and θ d values. Five different adaptive binning schemes were studied to construct 4D histograms of the VSMs, to ensure histogram efficient handling as well as an accurate reproduction of E, r s , φ d and θ d distribution details. The five resulting VSMs were then implemented in PENELOPE. Their accuracy was first assessed in the PS plane, by comparing E, r s , φ d and θ d distributions with those obtained from the reference PS file. Second, dose distributions computed in water, using the VSMs and the reference PS file located below the FF, and also after collimation in both water and heterogeneous phantom, were compared using a 1.5%-0 mm and a 2%-0 mm global gamma index, respectively. Finally, portal images were calculated without and with phantoms in the beam. The model was then evaluated using a 1%-0 mm global gamma index. Performance of a mono-source VSM was also investigated and led, as with the multi-source model, to excellent results when combined with an adaptive binning scheme.
Energy Technology Data Exchange (ETDEWEB)
Stuart, J.G.; Wright, A.D.; Butterfield, C.P.
1996-10-01
Mitigating the effects of damaging wind turbine loads and responses extends the lifetime of the turbine and, consequently, reduces the associated Cost of Energy (COE). Active control of aerodynamic devices is one option for achieving wind turbine load mitigation. Generally speaking, control system design and analysis requires a reasonable dynamic model of {open_quotes}plant,{close_quotes} (i.e., the system being controlled). This paper extends the wind turbine aileron control research, previously conducted at the National Wind Technology Center (NWTC), by presenting a more detailed development of the wind turbine dynamic model. In prior research, active aileron control designs were implemented in an existing wind turbine structural dynamics code, FAST (Fatigue, Aerodynamics, Structures, and Turbulence). In this paper, the FAST code is used, in conjunction with system identification, to generate a wind turbine dynamic model for use in active aileron control system design. The FAST code is described and an overview of the system identification technique is presented. An aileron control case study is used to demonstrate this modeling technique. The results of the case study are then used to propose ideas for generalizing this technique for creating dynamic models for other wind turbine control applications.
Energy Technology Data Exchange (ETDEWEB)
Nemoto, Toshiyuki; Watanabe, Hideo; Fujita, Toyozo [Fujitsu Ltd., Tokyo (Japan); Kawai, Wataru; Harada, Hiroo; Gorai, Kazuo; Yamasaki, Kazuhiko; Shoji, Makoto; Fujii, Minoru
1996-06-01
At Center for Promotion of Computational Science and Engineering, time consuming eight nuclear codes suggested by users have been vectorized, parallelized on the VPP500 computer system. In addition, two nuclear codes used on the VP2600 computer system were implemented on the VPP500 computer system. Neutron and photon transport calculation code MVP/GMVP and relativistic quantum molecular dynamics code QMDRELP have been parallelized. Extended quantum molecular dynamics code EQMD and adiabatic base calculation code HSABC have been parallelized and vectorized. Ballooning turbulence simulation code CURBAL, 3-D non-stationary compressible fluid dynamics code STREAM V3.1, operating plasma analysis code TOSCA and eddy current analysis code EDDYCAL have been vectorized. Reactor safety analysis code RELAP5/MOD2/C36-05 and RELAP5/MOD3 were implemented on the VPP500 computer system. (author)
THE TRANSFORM BETWEEN GEOGRAPHIC COORDINATES AND LOCATION CODES OF APERTURE 4 HEXAGONAL GRID SYSTEM
Directory of Open Access Journals (Sweden)
R. Wang
2017-09-01
Full Text Available Discrete global grid system is a new data model which supports the fusion processing of multi-source geospatial data. In discrete global grid systems, all cell operations can be completed by codes theoretically, but most of current spatial data are in the forms of geographic coordinates and projected coordinates. It is necessary to study the transform between geographic coordinates and grid codes, which will support data entering and getting out of the systems. This paper chooses the icosahedral hexagonal discrete global system as a base, and builds the mapping relationships between the sphere and the icosahedron. Then an encoding scheme of planar aperture 4 hexagonal grid system is designed and applied to the icosahedron. Basing on this, a new algorithm of transforms between geographic coordinates and grid codes is designed. Finally, experiments test the accuracy and efficiency of this algorithm. The efficiency of code addition of HLQT is about 5 times the efficiency of code addition of HQBS.
The Transform Between Geographic Coordinates and Location Codes of Aperture 4 Hexagonal Grid System
Wang, R.; Ben, J.; Li, Y.; Du, L.
2017-09-01
Discrete global grid system is a new data model which supports the fusion processing of multi-source geospatial data. In discrete global grid systems, all cell operations can be completed by codes theoretically, but most of current spatial data are in the forms of geographic coordinates and projected coordinates. It is necessary to study the transform between geographic coordinates and grid codes, which will support data entering and getting out of the systems. This paper chooses the icosahedral hexagonal discrete global system as a base, and builds the mapping relationships between the sphere and the icosahedron. Then an encoding scheme of planar aperture 4 hexagonal grid system is designed and applied to the icosahedron. Basing on this, a new algorithm of transforms between geographic coordinates and grid codes is designed. Finally, experiments test the accuracy and efficiency of this algorithm. The efficiency of code addition of HLQT is about 5 times the efficiency of code addition of HQBS.
Performance of Coded Systems with Generalized Selection Diversity in Nakagami Fading
Directory of Open Access Journals (Sweden)
Salam A. Zummo
2008-04-01
Full Text Available We investigate the performance of coded diversity systems employing generalized selection combining (GSC over Nakagami fading channels. In particular, we derive a numerical evaluation method for the cutoff rate of the GSC systems. In addition, we derive a new union bound on the bit-error probability based on the code's transfer function. The proposed bound is general to any coding scheme with a known weight distribution such as convolutional and trellis codes. Results show that the new bound is tight to simulation results for wide ranges of diversity order, Nakagami fading parameter, and signal-to-noise ratio (SNR.
Performance of Coded Systems with Generalized Selection Diversity in Nakagami Fading
Directory of Open Access Journals (Sweden)
Zummo SalamA
2008-01-01
Full Text Available Abstract We investigate the performance of coded diversity systems employing generalized selection combining (GSC over Nakagami fading channels. In particular, we derive a numerical evaluation method for the cutoff rate of the GSC systems. In addition, we derive a new union bound on the bit-error probability based on the code's transfer function. The proposed bound is general to any coding scheme with a known weight distribution such as convolutional and trellis codes. Results show that the new bound is tight to simulation results for wide ranges of diversity order, Nakagami fading parameter, and signal-to-noise ratio (SNR.
Energy Technology Data Exchange (ETDEWEB)
Chung, B. D.; Bae, S. W.; Jeong, J. J.; Lee, S. M
2005-04-15
A new multi-dimensional component has been developed to allow for more flexible 3D capabilities in the system code, MARS. This component can be applied in the Cartesian and cylindrical coordinates. For the development of this model, the 3D convection and diffusion terms are implemented in the momentum and energy equation. And a simple Prandtl's mixing length model is applied for the turbulent viscosity. The developed multi-dimensional component was assessed against five conceptual problems with analytic solution. And some SETs are calculated and compared with experimental data. With this newly developed multi-dimensional flow module, the MARS code can realistic calculate the flow fields in pools such as those occurring in the core, steam generators and IRWST.
Improvements in the Monte Carlo code for simulating 4πβ(PC)–γ coincidence system measurements
Energy Technology Data Exchange (ETDEWEB)
Dias, M.S., E-mail: msdias@ipen.br [Instituto de Pesquisas Energéticas e Nucleares, IPEN-CNEN/SP, Av. Prof. Lineu Prestes 2242, 05508-000 São Paulo, SP (Brazil); Takeda, M.N. [Universidade Santo Amaro, UNISA Rua Prof. Enéas da Siqueira Neto 340, 04829-300 São Paulo, SP (Brazil); Toledo, F.; Brancaccio, F.; Tongu, M.L.O.; Koskinas, M.F. [Instituto de Pesquisas Energéticas e Nucleares, IPEN-CNEN/SP, Av. Prof. Lineu Prestes 2242, 05508-000 São Paulo, SP (Brazil)
2013-01-11
A Monte Carlo simulation code known as ESQUEMA has been developed by the Nuclear Metrology Laboratory (Laboratório de Metrologia Nuclear—LMN) in the Nuclear and Energy Research Institute (Instituto de Pesquisas Energéticas e Nucleares—IPEN) to be used as a benchmark for radionuclide standardization. The early version of this code simulated only β−γ and ec−γ emitters with reasonably high electron and X-ray energies. To extend the code to include other radionuclides and enable the code to be applied to software coincidence counting systems, several improvements have been made and are presented in this work. -- Highlights: ► Improvements to the Monte Carlo code ESQUEMA are described. ► The experimental extrapolation curve was compared to Monte Carlo simulation. ► Eu-152 was standardized by 4π(PC)β-γ coincidence system and compared to Monte Carlo simulation. ► 4π proportional counter gamma-ray efficiency was calculated by MCNPX and compared with experiment. ► X-ray and positron decay emitters were included in the simulation.
DANTSYS: A diffusion accelerated neutral particle transport code system
Energy Technology Data Exchange (ETDEWEB)
Alcouffe, R.E.; Baker, R.S.; Brinkley, F.W.; Marr, D.R.; O`Dell, R.D.; Walters, W.F.
1995-06-01
The DANTSYS code package includes the following transport codes: ONEDANT, TWODANT, TWODANT/GQ, TWOHEX, and THREEDANT. The DANTSYS code package is a modular computer program package designed to solve the time-independent, multigroup discrete ordinates form of the boltzmann transport equation in several different geometries. The modular construction of the package separates the input processing, the transport equation solving, and the post processing (or edit) functions into distinct code modules: the Input Module, one or more Solver Modules, and the Edit Module, respectively. The Input and Edit Modules are very general in nature and are common to all the Solver Modules. The ONEDANT Solver Module contains a one-dimensional (slab, cylinder, and sphere), time-independent transport equation solver using the standard diamond-differencing method for space/angle discretization. Also included in the package are solver Modules named TWODANT, TWODANT/GQ, THREEDANT, and TWOHEX. The TWODANT Solver Module solves the time-independent two-dimensional transport equation using the diamond-differencing method for space/angle discretization. The authors have also introduced an adaptive weighted diamond differencing (AWDD) method for the spatial and angular discretization into TWODANT as an option. The TWOHEX Solver Module solves the time-independent two-dimensional transport equation on an equilateral triangle spatial mesh. The THREEDANT Solver Module solves the time independent, three-dimensional transport equation for XYZ and RZ{Theta} symmetries using both diamond differencing with set-to-zero fixup and the AWDD method. The TWODANT/GQ Solver Module solves the 2-D transport equation in XY and RZ symmetries using a spatial mesh of arbitrary quadrilaterals. The spatial differencing method is based upon the diamond differencing method with set-to-zero fixup with changes to accommodate the generalized spatial meshing.
Energy Technology Data Exchange (ETDEWEB)
Kai, Tetsuya; Maekawa, Fujio; Kasugai, Yoshimi; Takada, Hiroshi; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kosako, Kazuaki [Sumitomo Atomic Energy Industries, Ltd., Tokyo (Japan)
2002-03-01
Reliability assessment for the high energy particle induced radioactivity calculation code DCHAIN-SP 2001 was carried out through analysis of integral activation experiments with 14-MeV neutrons aiming at validating the cross section and decay data revised from previous version. The following three kinds of experiments conducted at the D-T neutron source facility, FNS, in JAERI were employed: (1) the decay gamma-ray measurement experiment for fusion reactor materials, (2) the decay heat measurement experiment for 32 fusion reactor materials, and (3) the integral activation experiment on mercury. It was found that the calculations with DCHAIN-SP 2001 predicted the experimental data for (1) - (3) within several tens of percent. It was concluded that the cross section data below 20 MeV and the associated decay data as well as the calculation algorithm for solving the Beteman equation that was the master equation of DCHAIN-SP were adequate. (author)
Blind Estimation of the Phase and Carrier Frequency Offsets for LDPC-Coded Systems
Directory of Open Access Journals (Sweden)
Houcke Sebastien
2010-01-01
Full Text Available Abstract We consider in this paper the problem of phase offset and Carrier Frequency Offset (CFO estimation for Low-Density Parity-Check (LDPC coded systems. We propose new blind estimation techniques based on the calculation and minimization of functions of the Log-Likelihood Ratios (LLR of the syndrome elements obtained according to the parity check matrix of the error-correcting code. In the first part of this paper, we consider phase offset estimation for a Binary Phase Shift Keying (BPSK modulation and propose a novel estimation technique. Simulation results show that the proposed method is very effective and outperforms many existing algorithms. Then, we modify the estimation criterion so that it can work for higher-order modulations. One interesting feature of the proposed algorithm when applied to high-order modulations is that the phase offset of the channel can be blindly estimated without any ambiguity. In the second part of the paper, we consider the problem of CFO estimation and propose estimation techniques that are based on the same concept as the ones presented for the phase offset estimation. The Mean Squared Error (MSE and Bit Error Rate (BER curves show the efficiency of the proposed estimation techniques.
Near-Capacity Coding for Discrete Multitone Systems with Impulse Noise
Directory of Open Access Journals (Sweden)
Kschischang Frank R
2006-01-01
Full Text Available We consider the design of near-capacity-achieving error-correcting codes for a discrete multitone (DMT system in the presence of both additive white Gaussian noise and impulse noise. Impulse noise is one of the main channel impairments for digital subscriber lines (DSL. One way to combat impulse noise is to detect the presence of the impulses and to declare an erasure when an impulse occurs. In this paper, we propose a coding system based on low-density parity-check (LDPC codes and bit-interleaved coded modulation that is capable of taking advantage of the knowledge of erasures. We show that by carefully choosing the degree distribution of an irregular LDPC code, both the additive noise and the erasures can be handled by a single code, thus eliminating the need for an outer code. Such a system can perform close to the capacity of the channel and for the same redundancy is significantly more immune to the impulse noise than existing methods based on an outer Reed-Solomon (RS code. The proposed method has a lower implementation complexity than the concatenated coding approach.
Low-Complexity Decoding of Block Turbo-Coded System with Antenna Diversity
Directory of Open Access Journals (Sweden)
Chen Yanni
2003-01-01
Full Text Available The goal of this paper is to reduce the decoding complexity of space-time block turbo-coded system with low performance degradation. Two block turbo-coded systems with antenna diversity are considered. These include the simple serial concatenation of error control code with space-time block code, and the recently proposed transmit antenna diversity scheme using forward error correction techniques. It is shown that the former performs better when compared to the latter in terms of bit error rate (BER under the same spectral efficiency (up to 7 dB at the BER of for quasistatic channel with two transmit and two receive antennas. For the former system, a computationally efficient decoding approach is proposed for the soft decoding of space-time block code. Compared to its original maximum likelihood decoding algorithm, it can reduce the computation by up to 70% without any performance degradation. Additionally, for the considered outer code block turbo code, through reduction of test patterns scanned in the Chase algorithm and the alternative computation of its extrinsic information during iterative decoding, extra 0.3 dB to 0.4 dB coding gain is obtained if compared with previous approaches with negligible hardware overhead. The overall decoding complexity is approximately ten times less than that of the near-optimum block turbo decoder with coding gain loss of 0.5 dB at the BER of over AWGN channel.
Sempau, J; Wilderman, S J; Bielajew, A F
2000-08-01
A new Monte Carlo (MC) algorithm, the 'dose planning method' (DPM), and its associated computer program for simulating the transport of electrons and photons in radiotherapy class problems employing primary electron beams, is presented. DPM is intended to be a high accuracy MC alternative to the current generation of treatment planning codes which rely on analytical algorithms based on an approximate solution of the photon/electron Boltzmann transport equation. For primary electron beams, DPM is capable of computing 3D dose distributions (in 1 mm3 voxels) which agree to within 1% in dose maximum with widely used and exhaustively benchmarked general-purpose public-domain MC codes in only a fraction of the CPU time. A representative problem, the simulation of 1 million 10 MeV electrons impinging upon a water phantom of 128(3) voxels of 1 mm on a side, can be performed by DPM in roughly 3 min on a modern desktop workstation. DPM achieves this performance by employing transport mechanics and electron multiple scattering distribution functions which have been derived to permit long transport steps (of the order of 5 mm) which can cross heterogeneity boundaries. The underlying algorithm is a 'mixed' class simulation scheme, with differential cross sections for hard inelastic collisions and bremsstrahlung events described in an approximate manner to simplify their sampling. The continuous energy loss approximation is employed for energy losses below some predefined thresholds, and photon transport (including Compton, photoelectric absorption and pair production) is simulated in an analogue manner. The delta-scattering method (Woodcock tracking) is adopted to minimize the computational costs of transporting photons across voxels.
GEDAE-LaB: A Free Software to Calculate the Energy System Contributions during Exercise.
Bertuzzi, Rômulo; Melegati, Jorge; Bueno, Salomão; Ghiarone, Thaysa; Pasqua, Leonardo A; Gáspari, Arthur Fernandes; Lima-Silva, Adriano E; Goldman, Alfredo
2016-01-01
The aim of the current study is to describe the functionality of free software developed for energy system contributions and energy expenditure calculation during exercise, namely GEDAE-LaB. Eleven participants performed the following tests: 1) a maximal cycling incremental test to measure the ventilatory threshold and maximal oxygen uptake (V̇O2max); 2) a cycling workload constant test at moderate domain (90% ventilatory threshold); 3) a cycling workload constant test at severe domain (110% V̇O2max). Oxygen uptake and plasma lactate were measured during the tests. The contributions of the aerobic (AMET), anaerobic lactic (LAMET), and anaerobic alactic (ALMET) systems were calculated based on the oxygen uptake during exercise, the oxygen energy equivalents provided by lactate accumulation, and the fast component of excess post-exercise oxygen consumption, respectively. In order to assess the intra-investigator variation, four different investigators performed the analyses independently using GEDAE-LaB. A direct comparison with commercial software was also provided. All subjects completed 10 min of exercise at moderate domain, while the time to exhaustion at severe domain was 144 ± 65 s. The AMET, LAMET, and ALMET contributions during moderate domain were about 93, 2, and 5%, respectively. The AMET, LAMET, and ALMET contributions during severe domain were about 66, 21, and 13%, respectively. No statistical differences were found between the energy system contributions and energy expenditure obtained by GEDAE-LaB and commercial software for both moderate and severe domains (P > 0.05). The ICC revealed that these estimates were highly reliable among the four investigators for both moderate and severe domains (all ICC ≥ 0.94). These findings suggest that GEDAE-LaB is a free software easily comprehended by users minimally familiarized with adopted procedures for calculations of energetic profile using oxygen uptake and lactate accumulation during exercise. By
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.
Noise suppression system of OCDMA with spectral/spatial 2D hybrid code
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.
Directory of Open Access Journals (Sweden)
Wonkyeong Kim
2015-01-01
Full Text Available A high-leakage core has been known to be a challenging problem not only for a two-step homogenization approach but also for a direct heterogeneous approach. In this paper the DIMPLE S06 core, which is a small high-leakage core, has been analyzed by a direct heterogeneous modeling approach and by a two-step homogenization modeling approach, using contemporary code systems developed for reactor core analysis. The focus of this work is a comprehensive comparative analysis of the conventional approaches and codes with a small core design, DIMPLE S06 critical experiment. The calculation procedure for the two approaches is explicitly presented in this paper. Comprehensive comparative analysis is performed by neutronics parameters: multiplication factor and assembly power distribution. Comparison of two-group homogenized cross sections from each lattice physics codes shows that the generated transport cross section has significant difference according to the transport approximation to treat anisotropic scattering effect. The necessity of the ADF to correct the discontinuity at the assembly interfaces is clearly presented by the flux distributions and the result of two-step approach. Finally, the two approaches show consistent results for all codes, while the comparison with the reference generated by MCNP shows significant error except for another Monte Carlo code, SERPENT2.
EMPIRE-II code-system with RIPL database as a tool for nuclear spectroscopy
Energy Technology Data Exchange (ETDEWEB)
Sin, Mihaela E-mail: msin@pcnet.ro
2004-07-21
The paper presents the modular system of nuclear reaction codes EMPIRE-II in conjunction with RIPL-2 database as a valuable tool in the spin-parity assignment for discrete levels of residual nuclei populated in reactions evolving through the compound nucleus mechanism. Several applications illustrating the method and code's predictive power are presented.
A QR code identification technology in package auto-sorting system
di, Yi-Juan; Shi, Jian-Ping; Mao, Guo-Yong
2017-07-01
Traditional manual sorting operation is not suitable for the development of Chinese logistics. For better sorting packages, a QR code recognition technology is proposed to identify the QR code label on the packages in package auto-sorting system. The experimental results compared with other algorithms in literatures demonstrate that the proposed method is valid and its performance is superior to other algorithms.
Avidan, Alexander; Weissman, Charles; Levin, Phillip D
2015-04-01
Quick response (QR) codes containing anesthesia syllabus data were introduced into an anesthesia information management system. The code was generated automatically at the conclusion of each case and available for resident case logging using a smartphone or tablet. The goal of this study was to evaluate the use and usability/user-friendliness of such system. Resident case logging practices were assessed prior to introducing the QR codes. QR code use and satisfactions amongst residents was reassessed at three and six months. Before QR code introduction only 12/23 (52.2%) residents maintained a case log. Most of the remaining residents (9/23, 39.1%) expected to receive a case list from the anesthesia information management system database at the end of their residency. At three months and six months 17/26 (65.4%) and 15/25 (60.0%) residents, respectively, were using the QR codes. Satisfaction was rated as very good or good. QR codes for residents' case logging with smartphones or tablets were successfully introduced in an anesthesia information management system and used by most residents. QR codes can be successfully implemented into medical practice to support data transfer. Copyright © 2015 Elsevier Ireland Ltd. All rights reserved.
PERFORMANCE ANALYSIS OF OPTICAL CDMA SYSTEM USING VC CODE FAMILY UNDER VARIOUS OPTICAL PARAMETERS
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HASSAN YOUSIF AHMED
2012-06-01
Full Text Available The intent of this paper is to study the performance of spectral-amplitude coding optical code-division multiple-access (OCDMA systems using Vector Combinatorial (VC code under various optical parameters. This code can be constructed by an algebraic way based on Euclidian vectors for any positive integer number. One of the important properties of this code is that the maximum cross-correlation is always one which means that multi-user interference (MUI and phase induced intensity noise are reduced. Transmitter and receiver structures based on unchirped fiber Bragg grating (FBGs using VC code and taking into account effects of the intensity, shot and thermal noise sources is demonstrated. The impact of the fiber distance effects on bit error rate (BER is reported using a commercial optical systems simulator, virtual photonic instrument, VPITM. The VC code is compared mathematically with reported codes which use similar techniques. We analyzed and characterized the fiber link, received power, BER and channel spacing. The performance and optimization of VC code in SAC-OCDMA system is reported. By comparing the theoretical and simulation results taken from VPITM, we have demonstrated that, for a high number of users, even if data rate is higher, the effective power source is adequate when the VC is used. Also it is found that as the channel spacing width goes from very narrow to wider, the BER decreases, best performance occurs at a spacing bandwidth between 0.8 and 1 nm. We have shown that the SAC system utilizing VC code significantly improves the performance compared with the reported codes.
Watanabe, Junpei; Ishikawa, Hiroaki; Arouette, Xavier; Matsumoto, Yasuaki; Miki, Norihisa
2012-06-01
In this paper, we present a vibrational Braille code display with large-displacement micro-electro-mechanical systems (MEMS) actuator arrays. Tactile receptors are more sensitive to vibrational stimuli than to static ones. Therefore, when each cell of the Braille code vibrates at optimal frequencies, subjects can recognize the codes more efficiently. We fabricated a vibrational Braille code display that used actuators consisting of piezoelectric actuators and a hydraulic displacement amplification mechanism (HDAM) as cells. The HDAM that encapsulated incompressible liquids in microchambers with two flexible polymer membranes could amplify the displacement of the MEMS actuator. We investigated the voltage required for subjects to recognize Braille codes when each cell, i.e., the large-displacement MEMS actuator, vibrated at various frequencies. Lower voltages were required at vibration frequencies higher than 50 Hz than at vibration frequencies lower than 50 Hz, which verified that the proposed vibrational Braille code display is efficient by successfully exploiting the characteristics of human tactile receptors.
A good performance watermarking LDPC code used in high-speed optical fiber communication system
Zhang, Wenbo; Li, Chao; Zhang, Xiaoguang; Xi, Lixia; Tang, Xianfeng; He, Wenxue
2015-07-01
A watermarking LDPC code, which is a strategy designed to improve the performance of the traditional LDPC code, was introduced. By inserting some pre-defined watermarking bits into original LDPC code, we can obtain a more correct estimation about the noise level in the fiber channel. Then we use them to modify the probability distribution function (PDF) used in the initial process of belief propagation (BP) decoding algorithm. This algorithm was tested in a 128 Gb/s PDM-DQPSK optical communication system and results showed that the watermarking LDPC code had a better tolerances to polarization mode dispersion (PMD) and nonlinearity than that of traditional LDPC code. Also, by losing about 2.4% of redundancy for watermarking bits, the decoding efficiency of the watermarking LDPC code is about twice of the traditional one.
The Marriage of Residential Energy Codes and Rating Systems: Conflict Resolution or Just Conflict?
Energy Technology Data Exchange (ETDEWEB)
Taylor, Zachary T.; Mendon, Vrushali V.
2014-08-21
After three decades of coexistence at a distance, model residential energy codes and residential energy rating systems have come together in the 2015 International Energy Conservation Code. At the October, 2013, International Code Council’s Public Comment Hearing, a new compliance path based on an Energy Rating Index was added to the IECC. Although not specifically named in the code, RESNET’s HERS rating system is the likely candidate Index for most jurisdictions. While HERS has been a mainstay in various beyond-code programs for many years, its direct incorporation into the most popular model energy code raises questions about the equivalence of a HERS-based compliance path and the traditional IECC performance compliance path, especially because the two approaches use different efficiency metrics, are governed by different simulation rules, and have different scopes with regard to energy impacting house features. A detailed simulation analysis of more than 15,000 house configurations reveals a very large range of HERS Index values that achieve equivalence with the IECC’s performance path. This paper summarizes the results of that analysis and evaluates those results against the specific Energy Rating Index values required by the 2015 IECC. Based on the home characteristics most likely to result in disparities between HERS-based compliance and performance path compliance, potential impacts on the compliance process, state and local adoption of the new code, energy efficiency in the next generation of homes subject to this new code, and future evolution of model code formats are discussed.
Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures
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Alessandro Petruzzi
2008-01-01
Full Text Available In the last four decades, large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Whereas the first system codes, developed at the beginning of the 1970s, utilized the homogenous equilibrium model with three balance equations to describe the two-phase flow, nowadays the more advanced system codes are based on the so-called “two-fluid model” with separation of the water and vapor phases, resulting in systems with at least six balance equations. The wide experimental campaign, constituted by the integral and separate effect tests, conducted under the umbrella of the OECD/CSNI was at the basis of the development and validation of the thermal-hydraulic system codes by which they have reached the present high degree of maturity. However, notwithstanding the huge amounts of financial and human resources invested, the results predicted by the code are still affected by errors whose origins can be attributed to several reasons as model deficiencies, approximations in the numerical solution, nodalization effects, and imperfect knowledge of boundary and initial conditions. In this context, the existence of qualified procedures for a consistent application of qualified thermal-hydraulic system code is necessary and implies the drawing up of specific criteria through which the code-user, the nodalization, and finally the transient results are qualified.
Pintado, O. I.; Santillán, L.; Marquetti, M. E.
All images obtained with a telescope are distorted by the instrument. This distorsion is known as instrumental profile or instrumental broadening. The deformations in the spectra could introduce large errors in the determination of different parameters, especially in those dependent on the spectral lines shapes, such as chemical abundances, winds, microturbulence, etc. To correct this distortion, in some cases, the spectral lines are convolved with a Gaussian function and in others the lines are widened with a fixed value. Some codes used to calculate synthetic spectra, as SYNTHE, include this corrections. We present results obtained for the spectrograph REOSC and EBASIM of CASLEO.
Beta Testing of CFD Code for the Analysis of Combustion Systems
Yee, Emma; Wey, Thomas
2015-01-01
A preliminary version of OpenNCC was tested to assess its accuracy in generating steady-state temperature fields for combustion systems at atmospheric conditions using three-dimensional tetrahedral meshes. Meshes were generated from a CAD model of a single-element lean-direct injection combustor, and the latest version of OpenNCC was used to calculate combustor temperature fields. OpenNCC was shown to be capable of generating sustainable reacting flames using a tetrahedral mesh, and the subsequent results were compared to experimental results. While nonreacting flow results closely matched experimental results, a significant discrepancy was present between the code's reacting flow results and experimental results. When wide air circulation regions with high velocities were present in the model, this appeared to create inaccurately high temperature fields. Conversely, low recirculation velocities caused low temperature profiles. These observations will aid in future modification of OpenNCC reacting flow input parameters to improve the accuracy of calculated temperature fields.
Energy Technology Data Exchange (ETDEWEB)
Pietrzak, Robert [Department of Nuclear Physics and Its Applications, Institute of Physics, University of Silesia, Katowice (Poland); Konefał, Adam, E-mail: adam.konefal@us.edu.pl [Department of Nuclear Physics and Its Applications, Institute of Physics, University of Silesia, Katowice (Poland); Sokół, Maria; Orlef, Andrzej [Department of Medical Physics, Maria Sklodowska-Curie Memorial Cancer Center, Institute of Oncology, Gliwice (Poland)
2016-08-01
The success of proton therapy depends strongly on the precision of treatment planning. Dose distribution in biological tissue may be obtained from Monte Carlo simulations using various scientific codes making it possible to perform very accurate calculations. However, there are many factors affecting the accuracy of modeling. One of them is a structure of objects called bins registering a dose. In this work the influence of bin structure on the dose distributions was examined. The MCNPX code calculations of Bragg curve for the 60 MeV proton beam were done in two ways: using simple logical detectors being the volumes determined in water, and using a precise model of ionization chamber used in clinical dosimetry. The results of the simulations were verified experimentally in the water phantom with Marcus ionization chamber. The average local dose difference between the measured relative doses in the water phantom and those calculated by means of the logical detectors was 1.4% at first 25 mm, whereas in the full depth range this difference was 1.6% for the maximum uncertainty in the calculations less than 2.4% and for the maximum measuring error of 1%. In case of the relative doses calculated with the use of the ionization chamber model this average difference was somewhat greater, being 2.3% at depths up to 25 mm and 2.4% in the full range of depths for the maximum uncertainty in the calculations of 3%. In the dose calculations the ionization chamber model does not offer any additional advantages over the logical detectors. The results provided by both models are similar and in good agreement with the measurements, however, the logical detector approach is a more time-effective method. - Highlights: • Influence of the bin structure on the proton dose distributions was examined for the MC simulations. • The considered relative proton dose distributions in water correspond to the clinical application. • MC simulations performed with the logical detectors and the
Cardinality enhancement utilizing Sequential Algorithm (SeQ) code in OCDMA system
Fazlina, C. A. S.; Rashidi, C. B. M.; Rahman, A. K.; Aljunid, S. A.
2017-11-01
Optical Code Division Multiple Access (OCDMA) has been important with increasing demand for high capacity and speed for communication in optical networks because of OCDMA technique high efficiency that can be achieved, hence fibre bandwidth is fully used. In this paper we will focus on Sequential Algorithm (SeQ) code with AND detection technique using Optisystem design tool. The result revealed SeQ code capable to eliminate Multiple Access Interference (MAI) and improve Bit Error Rate (BER), Phase Induced Intensity Noise (PIIN) and orthogonally between users in the system. From the results, SeQ shows good performance of BER and capable to accommodate 190 numbers of simultaneous users contrast with existing code. Thus, SeQ code have enhanced the system about 36% and 111% of FCC and DCS code. In addition, SeQ have good BER performance 10-25 at 155 Mbps in comparison with 622 Mbps, 1 Gbps and 2 Gbps bit rate. From the plot graph, 155 Mbps bit rate is suitable enough speed for FTTH and LAN networks. Resolution can be made based on the superior performance of SeQ code. Thus, these codes will give an opportunity in OCDMA system for better quality of service in an optical access network for future generation's usage
Nurses' attitudes toward the use of the bar-coding medication administration system.
Marini, Sana Daya; Hasman, Arie; Huijer, Huda Abu-Saad; Dimassi, Hani
2010-01-01
This study determines nurses' attitudes toward bar-coding medication administration system use. Some of the factors underlying the successful use of bar-coding medication administration systems that are viewed as a connotative indicator of users' attitudes were used to gather data that describe the attitudinal basis for system adoption and use decisions in terms of subjective satisfaction. Only 67 nurses in the United States had the chance to respond to the e-questionnaire posted on the CARING list server for the months of June and July 2007. Participants rated their satisfaction with bar-coding medication administration system use based on system functionality, usability, and its positive/negative impact on the nursing practice. Results showed, to some extent, positive attitude, but the image profile draws attention to nurses' concerns for improving certain system characteristics. The high bar-coding medication administration system skills revealed a more negative perception of the system by the nursing staff. The reasons underlying dissatisfaction with bar-coding medication administration use by skillful users are an important source of knowledge that can be helpful for system development as well as system deployment. As a result, strengthening bar-coding medication administration system usability by magnifying its ability to eliminate medication errors and the contributing factors, maximizing system functionality by ascertaining its power as an extra eye in the medication administration process, and impacting the clinical nursing practice positively by being helpful to nurses, speeding up the medication administration process, and being user-friendly can offer a congenial settings for establishing positive attitude toward system use, which in turn leads to successful bar-coding medication administration system use.
Improving performance of DS-CDMA systems using chaotic complex Bernoulli spreading codes
Farzan Sabahi, Mohammad; Dehghanfard, Ali
2014-12-01
The most important goal of spreading spectrum communication system is to protect communication signals against interference and exploitation of information by unintended listeners. In fact, low probability of detection and low probability of intercept are two important parameters to increase the performance of the system. In Direct Sequence Code Division Multiple Access (DS-CDMA) systems, these properties are achieved by multiplying the data information in spreading sequences. Chaotic sequences, with their particular properties, have numerous applications in constructing spreading codes. Using one-dimensional Bernoulli chaotic sequence as spreading code is proposed in literature previously. The main feature of this sequence is its negative auto-correlation at lag of 1, which with proper design, leads to increase in efficiency of the communication system based on these codes. On the other hand, employing the complex chaotic sequences as spreading sequence also has been discussed in several papers. In this paper, use of two-dimensional Bernoulli chaotic sequences is proposed as spreading codes. The performance of a multi-user synchronous and asynchronous DS-CDMA system will be evaluated by applying these sequences under Additive White Gaussian Noise (AWGN) and fading channel. Simulation results indicate improvement of the performance in comparison with conventional spreading codes like Gold codes as well as similar complex chaotic spreading sequences. Similar to one-dimensional Bernoulli chaotic sequences, the proposed sequences also have negative auto-correlation. Besides, construction of complex sequences with lower average cross-correlation is possible with the proposed method.
Development of a system of computer codes for severe accident analysis and its applications
Energy Technology Data Exchange (ETDEWEB)
Jang, S. H.; Chun, S. W.; Jang, H. S. and others [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)
1993-01-15
As a continuing study for the development of a system of computer codes to analyze severe accidents which had been performed last year, major focuses were on the aspect of application of the developed code systems. As the first step, two most commonly used code packages other than STCP, i.e., MELCOR of NRC and MAAP of IDCOR were reviewed to compare the models that they used. Next, important heat transfer phenomena were surveyed as severe accident progressed. Particularly, debris bed coolability and molten core-concrete interaction were selected as sample models, and they were studied extensively. The recent theoretical works and experiments for these phenomena were surveyed, and also the relevant models adopted by major code packages were compared and assessed. Based on the results obtained in this study, it is expected to be able to take into account these phenomenological uncertainties when one uses the severe accident code packages for probabilistic safety assessments or accident management programs.
Maximum Likelihood Blind Channel Estimation for Space-Time Coding Systems
Directory of Open Access Journals (Sweden)
Hakan A. Çırpan
2002-05-01
Full Text Available Sophisticated signal processing techniques have to be developed for capacity enhancement of future wireless communication systems. In recent years, space-time coding is proposed to provide significant capacity gains over the traditional communication systems in fading wireless channels. Space-time codes are obtained by combining channel coding, modulation, transmit diversity, and optional receive diversity in order to provide diversity at the receiver and coding gain without sacrificing the bandwidth. In this paper, we consider the problem of blind estimation of space-time coded signals along with the channel parameters. Both conditional and unconditional maximum likelihood approaches are developed and iterative solutions are proposed. The conditional maximum likelihood algorithm is based on iterative least squares with projection whereas the unconditional maximum likelihood approach is developed by means of finite state Markov process modelling. The performance analysis issues of the proposed methods are studied. Finally, some simulation results are presented.
Energy Technology Data Exchange (ETDEWEB)
Jaeger, Wadim
2011-12-19
The system code TRACE is the latest development of the U.S. Nuclear Regulatory Commission (US NRC). TRACE, developed for the analysis of operational conditions, transients and accidents of light water reactors (LWR), is a best-estimate code with two fluid, six equation models for mass, energy, and momentum conservation, and related closure models. Since TRACE is mainly applied to LWR specific issues, the validation process related to innovative nuclear systems (liquid metal cooled systems, systems operated with supercritical water, etc.) is very limited, almost not existing. In this work, essential contribution to the validation of TRACE related to lead and lead alloy cooled systems as well as systems operated with supercritical water is provided in a consistent and corporate way. In a first step, model discrepancies of the TRACE source code were removed. This inconsistencies caused the wrong prediction of the thermo physical properties of supercritical water and lead bismuth eutectic, and hence the incorrect prediction of heat transfer relevant characteristic numbers like Reynolds or Prandtl number. In addition to the correction of the models to predict these quantities, models describing the thermo physical properties of lead and Diphyl THT (synthetic heat transfer medium) were implemented. Several experiments and numerical benchmarks were used to validate the modified TRACE version. These experiments, mainly focused on wall-to-fluid heat transfer, revealed that not only the thermo physical properties are afflicted with inconsistencies but also the heat transfer models. The models for the heat transfer to liquid metals were enhanced in a way that the code can now distinguish between pipe and bundle flow by using the right correlation. The heat transfer to supercritical water was not existing in TRACE up to now. Completely new routines were implemented to overcome that issue. The comparison of the calculations to the experiments showed, on one hand, the necessity
Application of EPMA data for the development of the code systems TRANSURANUS and ALEPH.
Haeck, Wim; Verboomen, Bernard; Schubert, Arndt; Van Uffelen, Paul
2007-06-01
In the present article, electron probe microanalysis data for Pu and Nd is being used for validating the predictions of the radial power profile in a nuclear fuel rod at an ultrahigh burn-up of 95 and 102 MWd/kgHM. As such the validation of both the new Monte Carlo burn-up code ALEPH and the simpler TUBRNP model of the fuel rod performance code TRANSURANUS has been extended. The analysis of the absolute concentrations and individual isotopes also indicates potential improvements in the predictive capabilities of the simple TUBRNP model, based on the one-group cross sections inferred from the neutron transport calculations in the ALEPH code. This is a first important step toward extending the application range of the fuel rod performance code to burn-up values projected in nuclear power rods based on current trends.
CODING, ANALOG SYSTEMS), INFORMATION THEORY, DATA TRANSMISSION SYSTEMS , TRANSMITTER RECEIVERS, WHITE NOISE, PROBABILITY, ERRORS, PROBABILITY DENSITY FUNCTIONS, DIFFERENTIAL EQUATIONS, SET THEORY, COMPUTER PROGRAMS
Energy Technology Data Exchange (ETDEWEB)
Takahashi, Tomoyuki [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst; Takeda, Seiji; Kimura, Hideo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2001-01-01
It is indicated that some types of radioactive material generating from the development and utilization of nuclear energy do not need to be subject regulatory control because they can only give rise to trivial radiation hazards. The process to remove such materials from regulatory control is called as 'clearance'. The corresponding levels of the concentration of radionuclides are called as 'clearance levels'. In the Nuclear Safety Commission's discussion, the deterministic approach was applied to derive the clearance levels, which are the concentrations of radionuclides in a cleared material equivalent to an individual dose criterion. Basically, realistic parameter values were selected for it. If the realistic values could not be defined, reasonably conservative values were selected. Additionally, the stochastic approaches were performed to validate the results which were obtained by the deterministic calculations. We have developed a computer code system PASCLR (Probabilistic Assessment code System for derivation of Clearance Levels of Radioactive materials) by using the Monte Carlo technique for carrying out the stochastic calculations. This report describes the structure and user information for execution of PASCLR code. (author)
National Research Council Canada - National Science Library
Fabio Bignucolo; Alberto Cerretti; Massimiliano Coppo; Andrea Savio; Roberto Turri
2017-01-01
...), with negative impact on the safety of medium voltage (MV) and low voltage (LV) systems. With the scope of preserving the main network stability, international and national grid connection codes have been updated recently...
Validation Study of CODES Dragonfly Network Model with Theta Cray XC System
Energy Technology Data Exchange (ETDEWEB)
Mubarak, Misbah [Argonne National Lab. (ANL), Argonne, IL (United States); Ross, Robert B. [Argonne National Lab. (ANL), Argonne, IL (United States)
2017-05-31
This technical report describes the experiments performed to validate the MPI performance measurements reported by the CODES dragonfly network simulation with the Theta Cray XC system at the Argonne Leadership Computing Facility (ALCF).
Modulated Coding for Digital Communication Systems under ISI/Multipath Fading and Jamming
National Research Council Canada - National Science Library
Xia, Xiang-Gen
2002-01-01
.... The main achievements include the joint turbo and modulated coding for ISI channels, precoded and vector OFDM systems with single transmit antennas and matrix modulation schemes in broadband wireless...
A systematic literature review of automated clinical coding and classification systems.
Stanfill, Mary H; Williams, Margaret; Fenton, Susan H; Jenders, Robert A; Hersh, William R
2010-01-01
Clinical coding and classification processes transform natural language descriptions in clinical text into data that can subsequently be used for clinical care, research, and other purposes. This systematic literature review examined studies that evaluated all types of automated coding and classification systems to determine the performance of such systems. Studies indexed in Medline or other relevant databases prior to March 2009 were considered. The 113 studies included in this review show that automated tools exist for a variety of coding and classification purposes, focus on various healthcare specialties, and handle a wide variety of clinical document types. Automated coding and classification systems themselves are not generalizable, nor are the results of the studies evaluating them. Published research shows these systems hold promise, but these data must be considered in context, with performance relative to the complexity of the task and the desired outcome.
Sandia Engineering Analysis Code Access System v. 2.0.1
Energy Technology Data Exchange (ETDEWEB)
2017-10-30
The Sandia Engineering Analysis Code Access System (SEACAS) is a suite of preprocessing, post processing, translation, visualization, and utility applications supporting finite element analysis software using the Exodus database file format.
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.
Jeongeun, Kim; Sukwha, Kim; Kyusup, Han; Kyungsoon, Lee
2012-07-01
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. 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. 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. 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.
Directory of Open Access Journals (Sweden)
Dzhalandinov A.
2016-01-01
Full Text Available Usually the synthesis of two-dimensional and one-dimensional discrete ordinate calculations is used to evaluate neutron fluence on VVER-1000 reactor pressure vessel (RPV for prognosis of radiation embrittlement. But there are some cases when this approach is not applicable. For example the latest projects of VVER-1000 have upgraded surveillance program. Containers with surveillance specimens are located on the inner surface of RPV with fast neutron flux maximum. Therefore, the synthesis approach is not suitable enough for calculation of local disturbance of neutron field in RPV inner surface behind the surveillance specimens because of their complicated and heterogeneous structure. In some cases the VVER-1000 core loading consists of fuel assemblies with different fuel height and the applicability of synthesis approach is also ambiguous for these fuel cycles. Also, the synthesis approach is not enough correct for the neutron fluence estimation at the RPV area above core top. Because of these reasons only the 3D neutron transport codes seem to be satisfactory for calculation of neutron fluence on the VVER-1000 RPV. The direct 3D calculations are also recommended by modern regulations.
Energy Technology Data Exchange (ETDEWEB)
Jimenez, J.; Herrero, J. J.; Cuervo, D.; Aragones, J. M., E-mail: jimenez@din.upm.e [Universidad Politecnica de Madrid, Departamento de Ingenieria Nuclear, Jose Gutierrez Abascal No. 2, 28006 Madrid (Spain)
2010-10-15
Nowadays coupled 3-dimensional neutron kinetics and thermal-hydraulic core calculations are performed by applying a radial average channel approach using a meshing of one quarter of assembly in the best case. This approach does not take into account the subchannels effects due to the averaging of the physical fields and the loose of heterogeneity in the thermal-hydraulic model. Therefore the models do not have enough resolution to predict those subchannels effects which are important for the fuel design safety margins, because it is in the local scale, where we can search the hottest pellet or the maximum heat flux. The Polytechnic University of Madrid advanced multi-scale neutron-kinetics and thermal-hydraulics methodologies being implemented in COBAYA3 include domain decomposition by alternate core dissections for the local 3-dimensional fine-mesh scale problems (pin cells/subchannels) and an analytical nodal diffusion solver for the coarse mesh scale coupled with the thermal-hydraulic using a model of one channel per assembly or per quarter of assembly. In this work, we address the domain decomposition by the alternate core dissections methodology applied to solve coupled 3-dimensional neutronic-thermal-hydraulic problems at the fine-mesh scale. The neutronic-thermal-hydraulic coupling at the cell-subchannel scale allows the treatment of the effects of the detailed thermal-hydraulic feedbacks on cross-sections, thus resulting in better estimates of the local safety margins at the pin level. (Author)
Characterizing a Proton Beam Scanning System for Monte Carlo Dose Calculation in Patients
Grassberger, C; Lomax, Tony; Paganetti, H
2015-01-01
The presented work has two goals. First, to demonstrate the feasibility of accurately characterizing a proton radiation field at treatment head exit for Monte Carlo dose calculation of active scanning patient treatments. Second, to show that this characterization can be done based on measured depth dose curves and spot size alone, without consideration of the exact treatment head delivery system. This is demonstrated through calibration of a Monte Carlo code to the specific beam lines of two institutions, Massachusetts General Hospital (MGH) and Paul Scherrer Institute (PSI). Comparison of simulations modeling the full treatment head at MGH to ones employing a parameterized phase space of protons at treatment head exit reveals the adequacy of the method for patient simulations. The secondary particle production in the treatment head is typically below 0.2% of primary fluence, except for low–energy electrons (protons), whose contribution to skin dose is negligible. However, there is significant difference between the two methods in the low-dose penumbra, making full treatment head simulations necessary to study out-of field effects such as secondary cancer induction. To calibrate the Monte Carlo code to measurements in a water phantom, we use an analytical Bragg peak model to extract the range-dependent energy spread at the two institutions, as this quantity is usually not available through measurements. Comparison of the measured with the simulated depth dose curves demonstrates agreement within 0.5mm over the entire energy range. Subsequently, we simulate three patient treatments with varying anatomical complexity (liver, head and neck and lung) to give an example how this approach can be employed to investigate site-specific discrepancies between treatment planning system and Monte Carlo simulations. PMID:25549079
Calculation of Industrial Power Systems Containing Induction Motors
Directory of Open Access Journals (Sweden)
Gheorghe Hazi
2014-09-01
Full Text Available The current paper proposes two methods and algorithms for determining the operating regimes of industrial electrical networks which include induction motors. The two methods presented are based on specific principles for calculating electrical networks: Newton-Raphson and Backward-Forward for iteratively determining currents and voltages. The particularity of this paper is how the driven load influences the determination of the motors operating regimes. For the industrial machines driven by motors we take into account the characteristic of the resistant torque depending on speed. In this way, at the electrical busbars to which motors are connected, the active and the reactive power absorbed are calculated as a function of voltage as opposed to a regular consumer busbar. The algorithms for the two methods are presented. Finally, a numerical study for a test network is realized and the convergence is analyzed.
Spreading Code Assignment Strategies for MIMO-CDMA Systems Operating in Frequency-Selective Channels
Directory of Open Access Journals (Sweden)
Claude D'Amours
2009-01-01
Full Text Available Code Division Multiple Access (CDMA and multiple input multiple output- (MIMO- CDMA systems suffer from multiple access interference (MAI which limits the spectral efficiency of these systems. By making these systems more power efficient, we can increase the overall spectral efficiency. This can be achieved through the use of improved modulation and coding techniques. Conventional MIMO-CDMA systems use fixed spreading code assignments. By strategically selecting the spreading codes as a function of the data to be transmitted, we can achieve coding gain and introduce additional degrees of freedom in the decision variables at the output of the matched filters. In this paper, we examine the bit error rate performance of parity bit-selected spreading and permutation spreading under different wireless channel conditions. A suboptimal detection technique based on maximum likelihood detection is proposed for these systems operating in frequency selective channels. Simulation results demonstrate that these code assignment techniques provide an improvement in performance in terms of bit error rate (BER while providing increased spectral efficiency compared to the conventional system. Moreover, the proposed strategies are more robust to channel estimation errors as well as spatial correlation.
Spreading Code Assignment Strategies for MIMO-CDMA Systems Operating in Frequency-Selective Channels
Directory of Open Access Journals (Sweden)
Dahmane AdelOmar
2009-01-01
Full Text Available Abstract Code Division Multiple Access (CDMA and multiple input multiple output- (MIMO- CDMA systems suffer from multiple access interference (MAI which limits the spectral efficiency of these systems. By making these systems more power efficient, we can increase the overall spectral efficiency. This can be achieved through the use of improved modulation and coding techniques. Conventional MIMO-CDMA systems use fixed spreading code assignments. By strategically selecting the spreading codes as a function of the data to be transmitted, we can achieve coding gain and introduce additional degrees of freedom in the decision variables at the output of the matched filters. In this paper, we examine the bit error rate performance of parity bit-selected spreading and permutation spreading under different wireless channel conditions. A suboptimal detection technique based on maximum likelihood detection is proposed for these systems operating in frequency selective channels. Simulation results demonstrate that these code assignment techniques provide an improvement in performance in terms of bit error rate (BER while providing increased spectral efficiency compared to the conventional system. Moreover, the proposed strategies are more robust to channel estimation errors as well as spatial correlation.
A review on synchronous CDMA systems: optimum overloaded codes, channel capacity, and power control
Directory of Open Access Journals (Sweden)
Hosseini Seyed Amirhossein
2011-01-01
Full Text Available Abstract This paper is a tutorial review on important issues related to code-division multiple-access (CDMA systems such as channel capacity, power control, and optimum codes; specifically, we consider optimum overloaded codes that achieve errorless transmission in the absence of noise for the binary and nonbinary cases. A survey of lower and upper bounds for the sum channel capacity of such systems is given in the presence and absence of channel noise. The asymptotic results for the channel capacity are also investigated. The channel capacity, errorless transmission codes, and power estimation for near-far effects are also explored. The emphasis of this tutorial review is on the overloaded CDMA systems.
Coding and Billing in Surgical Education: A Systems-Based Practice Education Program.
Ghaderi, Kimeya F; Schmidt, Scott T; Drolet, Brian C
Despite increased emphasis on systems-based practice through the Accreditation Council for Graduate Medical Education core competencies, few studies have examined what surgical residents know about coding and billing. We sought to create and measure the effectiveness of a multifaceted approach to improving resident knowledge and performance of documenting and coding outpatient encounters. We identified knowledge gaps and barriers to documentation and coding in the outpatient setting. We implemented a series of educational and workflow interventions with a group of 12 residents in a surgical clinic at a tertiary care center. To measure the effect of this program, we compared billing codes for 1 year before intervention (FY2012) to prospectively collected data from the postintervention period (FY2013). All related documentation and coding were verified by study-blinded auditors. Interventions took place at the outpatient surgical clinic at Rhode Island Hospital, a tertiary-care center. A cohort of 12 plastic surgery residents ranging from postgraduate year 2 through postgraduate year 6 participated in the interventional sequence. A total of 1285 patient encounters in the preintervention group were compared with 1170 encounters in the postintervention group. Using evaluation and management codes (E&M) as a measure of documentation and coding, we demonstrated a significant and durable increase in billing with supporting clinical documentation after the intervention. For established patient visits, the monthly average E&M code level increased from 2.14 to 3.05 (p educational and workflow interventions, which improved resident coding and billing of outpatient clinic encounters. Using externally audited coding data, we demonstrate significantly increased rates of higher complexity E&M coding in a stable patient population based on improved documentation and billing awareness by the residents. Copyright © 2017 Association of Program Directors in Surgery. Published by
Spatiotemporal Coding of Individual Chemicals by the Gustatory System
National Research Council Canada - National Science Library
Reiter, Sam; Campillo Rodriguez, Chelsey; Sun, Kui; Stopfer, Mark
2015-01-01
.... However, using new methods for delivering tastant chemicals and making electrophysiological recordings from the tractable gustatory system of the moth Manduca sexta, we found chemical-specific information is as follows: (1...
Hierarchical sparse coding in the sensory system of Caenorhabditis elegans
Zaslaver, Alon; Liani, Idan; Shtangel, Oshrat; Ginzburg, Shira; Yee, Lisa; Sternberg, Paul W.
2015-01-01
Animals with compact sensory systems face an encoding problem where a small number of sensory neurons are required to encode information about its surrounding complex environment. Using Caenorhabditis elegans worms as a model, we ask how chemical stimuli are encoded by a small and highly connected sensory system. We first generated a comprehensive library of transgenic worms where each animal expresses a genetically encoded calcium indicator in individual sensory neurons....
Structural Calculations for Amorphous Systems Using Structural Diffusion Model
2001-06-01
Dalg,9, S. Dalg&, N. Talip , I. Orug Department of Physics, Trakya University, 22030 Edirne, TURKEY We present the results of calculations of the...or the deduced pair distribution function 538 S. Dalg&, S. Dalg&, N. Talip , I. Orug g(r). Both these properties are one-dimensional constructs...S(k) are shown in figure(1-a). We note that the differences between the 540 S. Dalg3q, S. Dalg&, N. Talip , I. Orue experimental S(k) in a-Fe not very
Analysis of PPM-CDMA and OPPM-CDMA communication systems with new optical code
Liu, F.; Ghafouri-Shiraz, H.
2005-11-01
A novel type of optical spreading sequences, named the 'new-Modified Prime Code (nMPC)', is proposed for use in synchronous direct-detection optical code-division multiple-access (CDMA) systems which employ both pulse position modulation (PPM) and overlapping pulse position modulation (OPPM) schemes. The upper bounds on the bit error rate (BER) for nMPC used in PPM-CDMA systems are derived and compared with the respective systems, using a modified prime code (MPC) and a padded modified prime code (PMPC). The nMPC is further applied to the OPPM-CDMA system and the system with a proposed interference cancellation scheme. Our results show that under the same conditions the PPM-CDMA system performances are more improved with the use of nMPC than with the two other traditional codes. Moreover, they show that the system performances are significantly enhanced by the proposed interference reduction methods, if the nMPC is used in the OPPM-CDMA systems.
Harvey, India; Bolgan, Samuela; Mosca, Daniel; McLean, Colin; Rusconi, Elena
2016-01-01
Studies on hacking have typically focused on motivational aspects and general personality traits of the individuals who engage in hacking; little systematic research has been conducted on predispositions that may be associated not only with the choice to pursue a hacking career but also with performance in either naïve or expert populations. Here, we test the hypotheses that two traits that are typically enhanced in autism spectrum disorders-attention to detail and systemizing-may be positively related to both the choice of pursuing a career in information security and skilled performance in a prototypical hacking task (i.e., crypto-analysis or code-breaking). A group of naïve participants and of ethical hackers completed the Autism Spectrum Quotient, including an attention to detail scale, and the Systemizing Quotient (Baron-Cohen et al., 2001, 2003). They were also tested with behavioral tasks involving code-breaking and a control task involving security X-ray image interpretation. Hackers reported significantly higher systemizing and attention to detail than non-hackers. We found a positive relation between self-reported systemizing (but not attention to detail) and code-breaking skills in both hackers and non-hackers, whereas attention to detail (but not systemizing) was related with performance in the X-ray screening task in both groups, as previously reported with naïve participants (Rusconi et al., 2015). We discuss the theoretical and translational implications of our findings.
The Effects of Bi-directional Refueling Scheme through CANDUCS-DEFENS Code System in a CANDU Reactor
Energy Technology Data Exchange (ETDEWEB)
Ryu, Eun Hyun; Song, Yong Mann; Park, Joo Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2013-05-15
Because the bi-directional refueling scheme gives the flattening of the power distribution of the core and enhancement of thermal-hydraulic conditions, it has been used as the standard refueling scheme in every CANDU-type reactor. In this research, the differences between the mono-directional refueling (MDR) and the bidirectional refueling (BDR) are discussed quantitatively with a multiplication factor and power distribution. In addition, by comparing the results from the CANDUCS-DEFENS code system with the RFSP results, the soundness of the CANDUCS code for a cross section generator will be verified. In this research, the results of RFSP and DEFENS codes are compared with each other for cases of MDR and BDR. For both cases, the multiplication factor error is extremely small, but the power errors are larger than those of mathematical initial core calculation. For the predictions of locations of max. rel. power err., max. channel power and max. bundle power, the DEFENS code prediction is matched well with the result of RFSP except for the prediction of location of max. bundle power for the BDR case. However, this is not a problem because the power difference between bundle 6 and 7 is negligible for the BDR case. In Fig. 2, Fig. 4, and Table X, it can be verified that the center of power distribution is moved to the centre of the core.
Andrade, Xavier; Strubbe, David; De Giovannini, Umberto; Larsen, Ask Hjorth; Oliveira, Micael J. T.; Alberdi-Rodriguez, Joseba; Varas, Alejandro; Theophilou, Iris; Helbig, Nicole; Verstraete, Matthieu J.; Stella, Lorenzo; Nogueira, Fernando; Aspuru-Guzik, Alán; Castro, Alberto; Marques, Miguel A. L.; Rubio, Angel
Real-space grids are a powerful alternative for the simulation of electronic systems. One of the main advantages of the approach is the flexibility and simplicity of working directly in real space where the different fields are discretized on a grid, combined with competitive numerical performance and great potential for parallelization. These properties constitute a great advantage at the time of implementing and testing new physical models. Based on our experience with the Octopus code, in this article we discuss how the real-space approach has allowed for the recent development of new ideas for the simulation of electronic systems. Among these applications are approaches to calculate response properties, modeling of photoemission, optimal control of quantum systems, simulation of plasmonic systems, and the exact solution of the Schr\\"odinger equation for low-dimensionality systems.
Pressure Profile in the experimental area of FCC-hh and FCC-ee calculated by an analytical code
Aichinger, Ida
2017-01-01
Ultra high vacuum in the beam pipe is a basic requirement for the Future Circular Colliders (FCC). The dimension of the FCC and the high energy of the particles will make this requirement challenging. Simulations that predict the vacuum quality due to material and beam induced effects will allow to evaluate different designs and to choose an optimal solution. The mathematical model behind the simulations will be shown. Four coupled differential equations describe the mass conservation of the residual gas particles in the beam pipe. The sinks include all kind of distributed and local pumping. The sources are caused by synchrotron radiation, electron clouds, thermal outgassing and ion-induced desorption. The equation system is solved by an analytical method. This requires a transformation to first order equations for which a general valid solution exists. Adding a particular solution and the inclusion of appropriate boundary conditions define the solution function. The big advantage here is that an analytical...
Code orange: Towards transformational leadership of emergency management systems.
Caro, Denis H J
2015-09-01
The 21(st) century calls upon health leaders to recognize and respond to emerging threats and systemic emergency management challenges through transformative processes inherent in the LEADS in a caring environment framework. Using a grounded theory approach, this qualitative study explores key informant perspectives of leaders in emergency management across Canada on pressing needs for relevant systemic transformation. The emerging model points to eight specific attributes of transformational leadership central to emergency management and suggests that contextualization of health leadership is of particular import. © 2015 The Canadian College of Health Leaders.
A Spanish version for the new ERA-EDTA coding system for primary renal disease.
Zurriaga, Óscar; López-Briones, Carmen; Martín Escobar, Eduardo; Saracho-Rotaeche, Ramón; Moina Eguren, Íñigo; Pallardó Mateu, Luis; Abad Díez, José María; Sánchez Miret, José Ignacio
2015-01-01
The European Renal Association and the European Dialysis and Transplant Association (ERA-EDTA) have issued an English-language new coding system for primary kidney disease (PKD) aimed at solving the problems that were identified in the list of "Primary renal diagnoses" that has been in use for over 40 years. In the context of Registro Español de Enfermos Renales (Spanish Registry of Renal Patients, [REER]), the need for a translation and adaptation of terms, definitions and notes for the new ERA-EDTA codes was perceived in order to help those who have Spanish as their working language when using such codes. Bilingual nephrologists contributed a professional translation and were involved in a terminological adaptation process, which included a number of phases to contrast translation outputs. Codes, paragraphs, definitions and diagnostic criteria were reviewed and agreements and disagreements aroused for each term were labelled. Finally, the version that was accepted by a majority of reviewers was agreed. A wide agreement was reached in the first review phase, with only 5 points of discrepancy remaining, which were agreed on in the final phase. Translation and adaptation into Spanish represent an improvement that will help to introduce and use the new coding system for PKD, as it can help reducing the time devoted to coding and also the period of adaptation of health workers to the new codes. Copyright © 2015 The Authors. Published by Elsevier España, S.L.U. All rights reserved.
Energy Technology Data Exchange (ETDEWEB)
White, Morgan C. [Univ. of Florida, Gainesville, FL (United States)
2000-07-01
The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V&V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second
Directory of Open Access Journals (Sweden)
Mohammad Mirzaie
2012-09-01
Full Text Available Background: The 131I radioisotope is used for diagnosis and treatment of hyperthyroidism and thyroid cancer. In optimized Iodine therapy, a specific dose must be reached to the thyroid gland with minimum radiation to the cervical spine, cervical vertebrae, neck tissue, subcutaneous fat and skin. Dose measurement inside the alive organ is difficult therefore the aim of this research was dose calculation in the organs by MCNPX code. Materials and Methods: First of all, the input file for MCNPX code has been prepared to calculate F6 and F8 tallies for ellipsoidal thyroid lobes with long axes is tow times of short axes which the 131I is distributed uniformly inside the lobes. Then the code has been run for F6 and F8 tallies for variation of lobe volume from 1 to 25 milliliters. From the output file of tally F6, the gamma absorbed dose in ellipsoidal thyroid, spinal neck, neck bone, neck tissue, subcutaneous fat layer and skin for the volume lobe variation from 1 ml to 25 ml have been derived and the graphs are drew. As well as, form the output of F8 tally the absorbed energy of beta in thyroid and soft tissue of neck is obtained and listed in the table and then absorbed dose of bate has been calculated. Results: The results of this research show that for constant activity in thyroid, the absorbed dose of gamma decreases about 88.3% in thyroid, 6.9% at soft tissue, 19.3% in adipose layer and 17.4% in skin, but it increases 32.1% in spinal of neck and 32.3% in neck bone when the lobe volume varied from 1 to 25 milliliters. For the same situation, the beta absorbed dose decreases 95.9% in thyroid and 64.2% in soft tissue. Conclusion: For the constant activity in thyroid by increasing the thyroid volume, absorbed dose of gamma in thyroid and soft tissue of neck, adipose layer under the skin and skin of neck decreased, but it increased at spinal of neck and neck bone. Also, by increasing of the lobe volume in constant activity, the beta absorbed dose
Improving 3D-Turbo Code's BER Performance with a BICM System over Rayleigh Fading Channel
Directory of Open Access Journals (Sweden)
R. Yao
2016-12-01
Full Text Available Classical Turbo code suffers from high error floor due to its small Minimum Hamming Distance (MHD. Newly-proposed 3D-Turbo code can effectively increase the MHD and achieve a lower error floor by adding a rate-1 post encoder. In 3D-Turbo codes, part of the parity bits from the classical Turbo encoder are further encoded through the post encoder. In this paper, a novel Bit-Interleaved Coded Modulation (BICM system is proposed by combining rotated mapping Quadrature Amplitude Modulation (QAM and 3D-Turbo code to improve the Bit Error Rate (BER performance of 3D-Turbo code over Raleigh fading channel. A key-bit protection scheme and a Two-Dimension (2D iterative soft demodulating-decoding algorithm are developed for the proposed BICM system. Simulation results show that the proposed system can obtain about 0.8-1.0 dB gain at BER of 10^{-6}, compared with the existing BICM system with Gray mapping QAM.
MATHEMATICAL MODELING OF HYDRAULIC CALCULATIONS IN DESIGNING FHC SYSTEMS
Directory of Open Access Journals (Sweden)
V. K. Gromov
2014-01-01
Full Text Available The law of Hydrant Systems control is formulated in the article. The method of construction of Hydrant Sustems mathematical models in the linearly-nodal form is presented. The distribution of liquid in a pipeline network is simulated by a system of equations describing Kirchhoff’s first and second laws in the nodal form.
Vision Aided Inertial Navigation System Augmented with a Coded Aperture
2011-03-24
21 Figure 2-4 Stereopsis Example...camera may be known precisely. Knowledge of this vector allows stereopsis techniques to be employed. With stereopsis , the angle from the focal point...for a MA V are accurate for significantly shorter time periods than those more commonly used for larger systems [9]. Aiding an INS using stereopsis
Load calculations of radiant cooling systems for sizing the plant
DEFF Research Database (Denmark)
Bourdakis, Eleftherios; Kazanci, Ongun Berk; Olesen, Bjarne W.
2015-01-01
% of the maximum cooling load. It was concluded that all tested systems were able to provide an acceptable thermal environment even when the 50% of the maximum cooling load was used. From all the simulated systems the one that performed the best under both control principles was the ESCS ceiling system. Finally......The aim of this study was, by using a building simulation software, to prove that a radiant cooling system should not be sized based on the maximum cooling load but at a lower value. For that reason six radiant cooling models were simulated with two control principles using 100%, 70% and 50...... it was proved that ventilation systems should be sized based on the maximum cooling load....
Performance of Turbo Interference Cancellation Receivers in Space-Time Block Coded DS-CDMA Systems
Directory of Open Access Journals (Sweden)
Emmanuel Oluremi Bejide
2008-07-01
Full Text Available We investigate the performance of turbo interference cancellation receivers in the space time block coded (STBC direct-sequence code division multiple access (DS-CDMA system. Depending on the concatenation scheme used, we divide these receivers into the partitioned approach (PA and the iterative approach (IA receivers. The performance of both the PA and IA receivers is evaluated in Rayleigh fading channels for the uplink scenario. Numerical results show that the MMSE front-end turbo space-time iterative approach receiver (IA effectively combats the mixture of MAI and intersymbol interference (ISI. To further investigate the possible achievable data rates in the turbo interference cancellation receivers, we introduce the puncturing of the turbo code through the use of rate compatible punctured turbo codes (RCPTCs. Simulation results suggest that combining interference cancellation, turbo decoding, STBC, and RCPTC can significantly improve the achievable data rates for a synchronous DS-CDMA system for the uplink in Rayleigh flat fading channels.
Revisiting the dose calculation methodologies in European decision support systems
DEFF Research Database (Denmark)
Andersson, Kasper Grann; Roos, Per; Hou, Xiaolin
2012-01-01
The paper presents examples of current needs for improvement and extended applicability of the European decision support systems. The systems were originally created for prediction of the radiological consequences of accidents at nuclear installations. They could however also be of great value...... in connection with management of the consequences of other types of contaminating incidents, including ‘dirty bomb’ explosions. This would require a number of new modelling features and parametric changes. Also for nuclear power plant preparedness a number of revisions of the decision support systems are called...
Auto Code Generation for Simulink-Based Attitude Determination Control System
MolinaFraticelli, Jose Carlos
2012-01-01
This paper details the work done to auto generate C code from a Simulink-Based Attitude Determination Control System (ADCS) to be used in target platforms. NASA Marshall Engineers have developed an ADCS Simulink simulation to be used as a component for the flight software of a satellite. This generated code can be used for carrying out Hardware in the loop testing of components for a satellite in a convenient manner with easily tunable parameters. Due to the nature of the embedded hardware components such as microcontrollers, this simulation code cannot be used directly, as it is, on the target platform and must first be converted into C code; this process is known as auto code generation. In order to generate C code from this simulation; it must be modified to follow specific standards set in place by the auto code generation process. Some of these modifications include changing certain simulation models into their atomic representations which can bring new complications into the simulation. The execution order of these models can change based on these modifications. Great care must be taken in order to maintain a working simulation that can also be used for auto code generation. After modifying the ADCS simulation for the auto code generation process, it is shown that the difference between the output data of the former and that of the latter is between acceptable bounds. Thus, it can be said that the process is a success since all the output requirements are met. Based on these results, it can be argued that this generated C code can be effectively used by any desired platform as long as it follows the specific memory requirements established in the Simulink Model.
Coding for MIMO-OFDM in future wireless systems
Ahmed, Bannour
2015-01-01
This book introduces the reader to the MIMO-OFDM system, in Rayleigh frequency selective-channels. Orthogonal frequency division multiplexing (OFDM) has been adopted in the wireless local-area network standards IEEE 802.11a due to its high spectral efficiency and ability to deal with frequency selective fading. The combination of OFDM with spectral efficient multiple antenna techniques makes the OFDM a good candidate to overcome the frequency selective problems.
Yan, Feng; Zhang, Xuejun
2009-09-14
In this paper the invariance of modulation transfer function (MTF), which describes the insensitivity to perturbation of MTF, is defined to be the evaluating criterion of the wavefront coding system. The rapid optimization of wavefront coding system based on the MTF invariance is proposed by means of introducing the mathematical program Matlab to normal optical design process. The interface called MZDDE between Matlab and Zemax is applied to realize the fast data exchanging and merit function calculating. The genetic algorithm tool (GA) in Matlab is introduced to the optimizing process, which accelerates the converging efficiency considerably. The MTF invariance of optimized system drops to 0.0119 while that of original system is larger than 0.018. If the all the fields of view is taken into consideration, the MTF invariance of optimized system and original system is less than 0.015 and larger than 0.020 respectively. It is proven that the optimization of the unusual optical system with special property can be executed conveniently and rapidly with the help of external program and dynamic data exchange.
Calculation of characteristics of torsionally vibrating mechatronic system
A. Buchacz
2007-01-01
Purpose: of this paper is the application of the approximate method to solve the task of assigning the frequencymodalanalysis and characteristics of a mechatronic system.Design/methodology/approach: was the formulated and solved as a problem in the form of a set of differentialequations of motion and state equations of the considered mechatronic model of an object. To obtain thesolution, Galerkin’s method was used. The discussed torsionally vibrating mechanical system is a continuousbar of ci...