Burnup calculation code system COMRAD96
International Nuclear Information System (INIS)
COMRAD was one of the burnup code system developed by JAERI. COMRAD96 is a transfered version of COMRAD to Engineering Work Station. It is divided to several functional modules, 'Cross Section Treatment', 'Generation and Depletion Calculation', and 'Post Process'. It enables us to analyze a burnup problem considering a change of neutron spectrum using UNITBURN. Also it can display the γ Spectrum on a terminal. This report is the general description and user's manual of COMRAD96. (author)
Code system BCG for gamma-ray skyshine calculation
International Nuclear Information System (INIS)
A code system BCG has been developed for calculating conveniently and efficiently gamma-ray skyshine doses using the transport calculation codes ANISN and DOT and the point-kernel calculation codes G-33 and SPAN. To simplify the input forms to the system, the forms for these codes are unified, twelve geometric patterns are introduced to give material regions, and standard data are available as a library. To treat complex arrangements of source and shield, it is further possible to use successively the code such that the results from one code may be used as input data to the same or other code. (author)
SRAC2006: A comprehensive neutronics calculation code system
International Nuclear Information System (INIS)
The SRAC is a code system applicable to neutronics analysis of a variety of reactor types. Since the publication of the second version of the users manual (JAERI-1302) in 1986 for the SRAC system, a number of additions and modifications to the functions and the library data have been made to establish a comprehensive neutronics code system. The current system includes major neutron data libraries (JENDL-3.3, JENDL-3.2, ENDF/B-VII, ENDF/B-VI.8, JEFF-3.1, JEF-2.2, etc.), and integrates five elementary codes for neutron transport and diffusion calculation; PIJ based on the collision probability method applicable to 16 kind of lattice models, SN transport codes ANISN(1D) and TWOTRN(2D), diffusion codes TUD(1D) and CITATION(multi-D). The system also includes an auxiliary code COREBN for multi-dimensional core burn-up calculation. (author)
Methods and computer codes for nuclear systems calculations
Indian Academy of Sciences (India)
B P Kochurov; A P Knyazev; A Yu Kwaretzkheli
2007-02-01
Some numerical methods for reactor cell, sub-critical systems and 3D models of nuclear reactors are presented. The methods are developed for steady states and space–time calculations. Computer code TRIFON solves space-energy problem in (, ) systems of finite height and calculates heterogeneous few-group matrix parameters of reactor cells. These parameters are used as input data in the computer code SHERHAN solving the 3D heterogeneous reactor equation for steady states and 3D space–time neutron processes simulation. Modification of TRIFON was developed for the simulation of space–time processes in sub-critical systems with external sources. An option of SHERHAN code for the system with external sources is under development.
Saphyr: a code system from reactor design to reference calculations
International Nuclear Information System (INIS)
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
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.
Miniature neutron source reactor burnup calculations using IRBURN code system
International Nuclear Information System (INIS)
Highlights: ► Fuel consumption of Iranian MNSR during 15 years of operation has been investigated. ► Calculations have been performed by the IRBURN code. Precision and accuracy of the implemented model has been validated. ► Our study shows the consumption rate of MNSR is about 1%. - Abstract: Fuel consumption of Iranian miniature neutron source reactor (MNSR) during 15 years of operation has been investigated. Reactor core neutronic parameters such as flux and power distributions, control rod worth and effective multiplication factor at BOL and after 15 years of irradiation has been calculated. The Monte Carlo-based depletion code system IRBURN has been used for studying the reactor core neutronic parameters as well as the isotopic inventory of the fuel during burnup. The precision and accuracy of the implemented model has been verified via validation the results for neutronic parameters in the MNSR final safety analysis report. The results show that keff decreases from 1.0034 to 0.9897 and the total U-235 consumption in the core is about 13.669 g after 15 years of operational time. Finally, our studying shows the consumption rate of MNSR is about 1%.
Revised SWAT. The integrated burnup calculation code system
Energy Technology Data Exchange (ETDEWEB)
Suyama, Kenya; Mochizuki, Hiroki [Department of Fuel Cycle Safety Research, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Kiyosumi, Takehide [The Japan Research Institute, Ltd., Tokyo (Japan)
2000-07-01
SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. This report shows an outline and a user's manual of revised SWAT. This revised SWAT includes expansion of functions, increasing supported machines, and correction of several bugs reported from users of previous SWAT. (author)
International Nuclear Information System (INIS)
In a D-T burning fusion reactor, the radioactivity induced by the 14 MeV neutrons causes many problems. It limits personnel access to the reactor during shutdown, generates decay heat and produces radwastes. A code system THIDA had been developed in 1978 to calculate the radioactivity and dose rate around a fusion device. The THIDA system consisted of the followings: one- and two-dimensional discrete ordinates radiation transport codes; induced activity calculation code; three libraries for transmutation and decay chain data, transmutation cross sections and delayed gamma-ray emission data. The present report gives a complete description of THIDA-2, a new advanced version of the THIDA system which has the following major improvements: 1. Capability to treat three-dimensional calculation models by the use of a Monte Carlo transport code. 2. Accurate decay heat calculation following the transport of delayed gamma rays. 3. Simplification of the data input process by the use of free format scheme and closer coupling between the radiation transport codes and the induced activity calculation code. 4. Self-descriptive output format and additional plotter output. 5. Capability to calculate problems requiring larger core memory by the use of variable dimension. (author)
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.
Validation of WIMS-SNAP code systems for calculations in TRIGA-MARK II type reactors
International Nuclear Information System (INIS)
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/d4 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)
Erosion corrosion in power plant piping systems - Calculation code for predicting wall thinning
International Nuclear Information System (INIS)
Extensive experimental and theoretical investigations have been performed to develop a calculation code for wall thinning due to erosion corrosion in power plant piping systems. The so-called WATHEC code can be applied to single-phase water flow as well as to two-phase water/steam flow. Only input data which are available to the operator of the plant are taken into consideration. Together with a continuously updated erosion corrosion data base the calculation code forms one element of a weak point analysis for power plant piping systems which can be applied to minimize material loss due to erosion corrosion, reduce non-destructive testing and curtail monitoring programs for piping systems, recommend life-extending measures. (author). 12 refs, 17 figs
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)
MOSRA-SRAC. Lattice calculation module of the modular code system for nuclear reactor analyses MOSRA
International Nuclear Information System (INIS)
MOSRA-SRAC is a lattice calculation module of the Modular code System for nuclear Reactor Analyses (MOSRA). This module performs the neutron transport calculation for various types of fuel elements including existing light water reactors, research reactors, etc. based on the collision probability method with a set of the 200-group cross-sections generated from the Japanese Evaluated Nuclear Data Library JENDL-4.0. It has also a function of the isotope generation and depletion calculation for up to 234 nuclides in each fuel material in the lattice. In these ways, MOSRA-SRAC prepares the burn-up dependent effective microscopic and macroscopic cross-section data to be used in core calculations. A CD-ROM is attached as an appendix. (J.P.N.)
CEQCSY: a new code for chemical equilibrium calculation in multiphased systems
International Nuclear Information System (INIS)
As part of the CEC Chemval/mirage project, a method is presented for calculating the thermodynamic equilibrium state of a multiphase system, by minimizing its Gibbs free energy constrained by mass balances. Compared to the other algorithms available in the literature, the method has three main characteristics: - the sets of equations corresponding to the conditions of homogeneous and heterogeneous equilibria are simultaneously solved, - a mathematical criterion for bringing a new multicomponent phase in the system is rigorously demontrated. - It enables a detailed representation of the multisite solid solutions with constraints called site closure relation. The code CEQCSY (Chemical Equilibrium in Complex SYstem) uses this formalism, and works with the thermodynamic data base from the EQ3/6 code. This choice makes easier several compared tests with EQ6: quartz dissolution in water, water-atmospheric air equilibrium, theoretical re-equilibrium of seawater, hydrothermal alteration of granite including solid solutions. The test results demonstrate the high efficiency and velocity of the code CEQCSY, when working on equilibrium state of multiphase systems. This high velocity was the aim of this work, in order to couple with thermic, hydrodynamic or mechanic codes
International Nuclear Information System (INIS)
A computer code system for fast calculation of activation and transmutation has been developed. The system consists of a driver code, cross-section libraries, flux libraries, a material library, and a decay library. The code is used to predict transmutations in a Ti-modified 316 stainless steel, a commercial ferritic alloy (HT9), and a V-15%Cr-5%Ti alloy in various magnetic fusion energy (MFE) test facilities and conceptual reactors
Owen, Albert K.
1987-01-01
A computer code was written which utilizes ray tracing techniques to predict the changes in position and geometry of a laser Doppler velocimeter probe volume resulting from refraction effects. The code predicts the position change, changes in beam crossing angle, and the amount of uncrossing that occur when the beams traverse a region with a changed index of refraction, such as a glass window. The code calculates the changes for flat plate, cylinder, general axisymmetric and general surface windows and is currently operational on a VAX 8600 computer system.
International Nuclear Information System (INIS)
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 keff (''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.
Decay heat experiment and validation of calculation code systems for fusion reactor
International Nuclear Information System (INIS)
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)
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
Grebe, A; Lu, T; Mokhov, N; Pronskikh, V
2016-01-01
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)
International Nuclear Information System (INIS)
Highlights: • Benchmark study performed for the neutronic calculations of TRIGA research reactors. • WIMSD-5B/CITATION is the utilized code system along with the WIMSD-IAEA-69 library. • The studied condensed spectra are five and seven energy groups spectra. • Analyzed: lattice parameters, reactivities, CR worth, flux and power distribution. • The lattice and neutronic parameters showed the accuracy of both condensed spectra. - Abstract: The objective of this paper is to assess the suitability and accuracy of the deterministic diffusion method for the neutronic calculations of the TRIGA Mark-III research reactors using the WIMSD/CITATION code system in proposed condensed energy spectra of five and seven energy groups with one and three thermal groups respectively. The utilized cell transport calculations code and core diffusion calculations code are the WIMSD-5B and the CITVAP v3.1 codes respectively, along with the WIMSD-IAEA-69 nuclear data library. Firstly, the assessment goes through analyzing the integral parameters – keff, ρ238, δ235, δ238, and C* – of the TRX and BAPL benchmark lattices and comparison with experimental and previous reference results using other ENDLs at the full energy spectra which show good agreement with the references at both spectra. Secondly, evaluation of the 3D nuclear characteristics of three different cores of the TRR-1/M1 TRIGA Mark-III Thai research reactor at the condensed energy spectra. The results include the excess reactivities of the cores and the worth of selected control rods which were compared with reference Monte Carlo results and experimental values. The results show good agreement with the references at both energy spectra and the better accuracy are attainable in the five energy groups spectrum. The results also include neutron flux distributions which are evaluated for future comparisons with other calculational techniques even they are comparable to reactors and fuels of the same type. 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.)
International Nuclear Information System (INIS)
The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged
Comparison of energy deposition calculations by the LAHET Code System with experimental results
Energy Technology Data Exchange (ETDEWEB)
Beard, C.A.; Lisowski, P.W.; Russell, G.J.; Waters, L.S.
1993-08-01
A comparison was performed between the energy deposition predicted by the LAHET Code System (LCS) with experimental values determined by Belyakov-Bodin et al. for 800, 1000, and 1200 MeV protons on targets composed of lead, bismuth, beryllium, carbon, and aluminum. The lead and bismuth showed agreement within approximately 10% at locations throughout the targets, and the agreement of the total energy deposited over the axial length of the targets ranged from 1% to 25%. For the lead and bismuth cases, the LCS predictions were always greater than the experimental results. For the lighter materials, the agreement at locations throughout the target only agreed within approximately 20%. No definable trend could be determined for the lighter materials since some LCS predictions were greater than the experimental results, some were less than the experimental results, and some showed very good agreement. The total energy deposited over the axial length of the targets was not compared for the lighter materials since it was not explicitly given with the experimental data.
International Nuclear Information System (INIS)
In Many situations of nuclear system study, it is necessary to know the detailed particle flux in a geometry. Deterministic 1-D and 2-D methods aren't suitable to represent some strong 3-D behavior configurations, for example in cores where the neutron flux varies considerably in the space and Monte Carlo analysis are necessary. The majority of Monte Carlo transport calculation codes, performs time static simulations, in terms of fuel isotopic composition. This work is a initial project to incorporate depletion capability to the MCNP code, by means of a connection with ORIGEN2.1 burnup code. The method to develop the program proposed followed the methodology of other programs used to the same purpose. Essentially, MCNP data library are used to generate one group microscopic cross sections that override default ORIGEN libraries. To verify the actual implemented part, comparisons which MCNPX (version 2.6.0) results were made. The neutron flux and criticality value of core agree. The neutron flux and criticality value of the core agree, especially in beginning of burnup when the influence of fission products are not very considerable. The small difference encountered was probably caused by the difference in the number of isotopes considered in the transport models (89 MCNPX x 25 GB). Next step of this work is to adapt MCNP version 4C to work with a memory higher than its standard value (4MB), in order to allow a greater number of isotopes in the transport model. (author)
International Nuclear Information System (INIS)
Nuclear technology development pointed out the need for a new assessment of the fuel cycle back-end. Treatment and disposal of radioactive wastes arising from nuclear fuel reprocessing is known as one of the problems not yet satisfactorily solved, together with separation process of uranium and plutonium from fission products in highly irradiated fuels. Aim of this work is to present an improvement of the computer code for solvent extraction process calculation previously designed by the authors. The modeling of the extraction system has been modified by introducing a new method for calculating the distribution coefficients. The new correlations were based on deriving empirical functions for not only the apparent equilibrium constants, but also the solvation number. The mathematical model derived for calculating separation performance has been then tested for up to ten components and twelve theoretical stages with minor modifications to the convergence criteria. Suitable correlations for the calculation of the distribution coefficients of Uranium, Plutonium, Nitric Acid and fission products were constructed and used to successfully simulate several experimental conditions. (Author)
International Nuclear Information System (INIS)
EFFI calculates the electromagnetic field and vector potential in coil systems of arbitrary geometry. The coils are made from circular arc and/or straight segments of rectangular cross-section conductor. EFFI can also calculate magnetic flux lines, magnetic force, and inductance. The methods used for the calculations are based on a combination of analytical and numerical integration of the Biot--Savart law for a volume distribution of current. These methods yield accurate field values inside and outside the conductor. All input to EFFI is format-free and is checked for validity before any calculations are done. Any errors detected during the check produce a diagnostic that lists the error, the code's objection to it, and the number of the offending data card. EFFI produces output in both printed and graphical form. Each page of output is labeled with the title of the problem, the time, the computer and date of the run, and the version number and compilation date for EFFI. In addition, each column of numbers on each page is appropriately labeled. Examples from the coil design for the Mirror Fusion Test Facility (MFTF) and a divertor design for a Tokamak reactor are used for illustration
International Nuclear Information System (INIS)
The calculations performed for the Almaraz Unit 2 PWR using the code packages of the Atomic Energy Corporation of South Africa Ltd. are summarized. These calculations were done as part of the IAEA Coordinated Research Programme on In-Core Fuel Management Code Package Validation for LWRs. A brief description of the one-dimensional cross section generation package as well as of the Level II (scoping type) global core calculational package which was used is given. Detailed results are presented in several appendices. 29 figs., 20 tabs., 10 refs
KENO-IV code benchmark calculation, (6)
International Nuclear Information System (INIS)
A series of benchmark tests has been undertaken in JAERI in order to examine the capability of JAERI's criticality safety evaluation system consisting of the Monte Carlo calculation code KENO-IV and the newly developed multigroup constants library MGCL. The present report describes the results of a benchmark test using criticality experiments about Plutonium fuel in various shape. In all, 33 cases of experiments have been calculated for Pu(NO3)4 aqueous solution, Pu metal or PuO2-polystyrene compact in various shape (sphere, cylinder, rectangular parallelepiped). The effective multiplication factors calculated for the 33 cases distribute widely between 0.955 and 1.045 due to wide range of system variables. (author)
KENO-IV code benchmark calculation, (4)
International Nuclear Information System (INIS)
A series of benchmark tests has been undertaken in JAERI in order to examine the capability of JAERI's criticality safety evaluation system consisting of the Monte Carlo calculation code KENO-IV and the newly developed multi-group constants library MGCL. The present paper describes the results of a test using criticality experiments about slab-cylinder system of uranium nitrate solution. In all, 128 cases of experiments have been calculated for the slab-cylinder configuration with and without plexiglass reflector, having the various critical parameters such as the number of cylinders and height of the uranium nitrate solution. It is shown among several important results that the code and library gives a fairly good multiplication factor, that is, k sub(eff) -- 1.0 for heavily reflected cases, whereas k sub(eff) -- 0.91 for the unreflected ones. This suggests the necessity of more advanced treatment of the criticality calculation for the system where neutrons can easily leak out during slowing down process. (author)
International Nuclear Information System (INIS)
Full text: The implementation in the reaction code system EMPIRE-2.19 of an advanced formalism for fission cross-section calculation has been completed. The formalism is based on the optical model for fission and can be applied for nuclei exhibiting double- or triple-humped barrier starting from sub-barrier excitation energies. The optical model for fission, initially developed to describe the resonant structure of the fission cross section at sub-barrier excitation energies due to the vibrational states in the second well of a double-humped fission barrier, was extended to light actinides by including the relations for the transmission coefficients through a complex triple-humped fission barrier. The real part of the fission barrier is parameterised as a function of the nucleus deformation by five smoothly joined parabolas. The imaginary potential is introduced only in the deformation range corresponding to the second well because the tertiary well is supposed to be shallow enough to neglect the damping of class III vibrational states. The transition states are assumed to be rotational states built on vibrational or non-collective band-heads. As the excitation energy increases, the shell effect, which causes the splitting of the outer barrier, diminishes and the outer humps lump into a single one. Therefore, in the present formalism, triple-humped barriers are associated only to the discrete transition states; the contribution of continuum to the fission coefficients is calculated considering a double-humped barrier. The parameters of the second single barrier equivalent with the outer humps are being determined from the condition of equal transmission coefficients. The saddle-point transition states in continuum are described by level densities (BCS below the critical energy and a modified version of Fermi Gas above) accounting for collective enhancements specific to the nuclear shape asymmetry at each saddle point . The neutron cross sections of 232Th in the
The VADMAP code to calculate the SAF of photon
International Nuclear Information System (INIS)
A computer code VADMAP has been developed to calculate the Specific Absorbed Fraction, SAF, of photon. The development of the code is aimed at efficient and systematic preparation of the SAF data files for several different human phantoms in a suitable form as a direct input data file to DOSimetric DAta Calculation system, DOSDAC, which is being developed at Japan Atomic Energy Research Institute, JAERI. This document describes the methodology used in the code, the code structure, user's information including the way of implementing the code on FACOM/M-380, and the performance through calculation and preparation of the SAF data file. In order to show the performance of the code, a set of the SAF values for an adult human phantom was calculated and was organized to prepare the SAF file. Comparing the calculated SAF values with those tabulated in ORNL-5000, the quality of the code was examined. (author)
Coupling CFD code with system code and neutron kinetic code
International Nuclear Information System (INIS)
Highlights: • Coupling interface between CFD code Fluent and system code Athlet was created. • Athlet code is internally coupled with neutron kinetic code Dyn3D. • Explicit coupling of overlapped computational domains was used. • A coupled system of Athlet/Dyn3D+Fluent codes was successfully tested on a real case. - Abstract: The aim of this work was to develop the coupling interface between CFD code Fluent and system code Athlet internally coupled with neutron kinetic code Dyn3D. The coupling interface is intended for simulation of complex transients such as Main Steam Line Break scenarios, which cannot be modeled separately first by system and neutron kinetic code and then by CFD code, because of the feedback between the codes. In the first part of this article, the coupling method is described. Explicit coupling of overlapped computational domains is used in this work. The second part of the article presents a demonstration simulation performed by the coupled system of Athlet/Dyn3D and Fluent. The “Opening a Steam Dump to the Atmosphere” test carried out at the Temelin NPP (VVER-1000) was simulated by the coupled system. In this simulation, the primary and secondary circuits were modeled by Athlet, mixing in downcomer and lower plenum was simulated by Fluent and heat generation in the core was calculated by Dyn3D. The results of the simulation with Athlet/Dyn3D+Fluent were compared with the experimental data and the results from a calculation performed with Athlet/Dyn3D without Fluent
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.
Calculation code MIXSET for Purex process
International Nuclear Information System (INIS)
MIXSET is a FORTRAN IV calculation code for Purex process that simulate the dynamic behavior of solvent extraction processes in mixer-settlers. Two options permit terminating dynamic phase by time or by achieving steady state. These options also permit continuing calculation successively using new inputs from a arbitrary phase. A third option permits artificial rapid close to steady state and a fourth option permits searching optimum input to satisfy both of specification and recovery rate of product. MIXSET handles maximum chemical system of eight components with or without mutual dependence of the distribution of the components. The chemical system in MIXSET includes chemical reactions and/or decaying reaction. Distribution data can be supplied by third-power polynominal equations or tables, and kinetic data by tables or given constants. The fluctuation of the interfacial level height in settler is converted into the flow rate changes of organic and aqueous stream to follow dynamic behavior of extraction process in detail. MIXSET can be applied to flowsheet study, start up and/or shut down procedure study and real time process management in countercurrent solvent extraction processes. (auth.)
International Nuclear Information System (INIS)
Highlights: → The specifications required for the analyses of the destructive assay data taken from irradiated fuel in Ohi-1 and Ohi-2 PWRs were documented in this paper. → These data were analyzed using the SWAT2.1 code, and the calculation results showed good agreement with experimental results. → These destructive assay data are suitable for the benchmarking of the burnup calculation code systems. - Abstract: The isotopic composition of spent nuclear fuels is vital data for studies on the nuclear fuel cycle and reactor physics. The Japan Atomic Energy Agency (JAEA) has been active in obtaining such data for pressurized water reactor (PWR) and boiling water reactor (BWR) fuels, and some data has already been published. These data have been registered with the international Spent Fuel Isotopic Composition Database (SFCOMPO) and widely used as international benchmarks for burnup calculation codes and libraries. In this paper, Assay Data of Spent Nuclear Fuel from two fuel assemblies irradiated in the Ohi-1 and Ohi-2 PWRs in Japan are shown. The destructive assay data from Ohi-2 have already been published. However, these data were not suitable for the benchmarking of calculation codes and libraries because several important specifications and data were not included. This paper summarizes the details of destructive assay data and specifications required for analyses of isotopic composition from Ohi-1 and Ohi-2. For precise burnup analyses, the burnup values of destructive assay samples were re-evaluated in this study. These destructive assay data were analyzed using the SWAT2.1 code, and the calculation results showed good agreement with experimental results. This indicates that the quality of destructive assay data from Ohi-1 and Ohi-2 PWRs is high, and that these destructive assay data are suitable for the benchmarking of burnup calculation code systems.
TEA: A Code Calculating Thermochemical Equilibrium Abundances
Blecic, Jasmina; Harrington, Joseph; Bowman, M. Oliver
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 & Sharp, the free thermochemical equilibrium code Chemical Equilibrium with Applications (CEA), and the example given by Burrows & 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.
TEA: A Code Calculating Thermochemical Equilibrium Abundances
Blecic, Jasmina; Harrington, Joseph; Bowman, M. Oliver
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 & Sharp, the free thermochemical equilibrium code Chemical Equilibrium with Applications (CEA), and the example given by Burrows & 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.
MCOR - Monte Carlo depletion code for reference LWR calculations
Energy Technology Data Exchange (ETDEWEB)
Puente Espel, Federico, E-mail: fup104@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Tippayakul, Chanatip, E-mail: cut110@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Ivanov, Kostadin, E-mail: kni1@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Misu, Stefan, E-mail: Stefan.Misu@areva.com [AREVA, AREVA NP GmbH, Erlangen (Germany)
2011-04-15
Research highlights: > Introduction of a reference Monte Carlo based depletion code with extended capabilities. > Verification and validation results for MCOR. > Utilization of MCOR for benchmarking deterministic lattice physics (spectral) codes. - Abstract: The MCOR (MCnp-kORigen) code system is a Monte Carlo based depletion system for reference fuel assembly and core calculations. The MCOR code is designed as an interfacing code that provides depletion capability to the LANL Monte Carlo code by coupling two codes: MCNP5 with the AREVA NP depletion code, KORIGEN. The physical quality of both codes is unchanged. The MCOR code system has been maintained and continuously enhanced since it was initially developed and validated. The verification of the coupling was made by evaluating the MCOR code against similar sophisticated code systems like MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN. After its validation, the MCOR code has been further improved with important features. The MCOR code presents several valuable capabilities such as: (a) a predictor-corrector depletion algorithm, (b) utilization of KORIGEN as the depletion module, (c) individual depletion calculation of each burnup zone (no burnup zone grouping is required, which is particularly important for the modeling of gadolinium rings), and (d) on-line burnup cross-section generation by the Monte Carlo calculation for 88 isotopes and usage of the KORIGEN libraries for PWR and BWR typical spectra for the remaining isotopes. Besides the just mentioned capabilities, the MCOR code newest enhancements focus on the possibility of executing the MCNP5 calculation in sequential or parallel mode, a user-friendly automatic re-start capability, a modification of the burnup step size evaluation, and a post-processor and test-matrix, just to name the most important. The article describes the capabilities of the MCOR code system; from its design and development to its latest improvements and further ameliorations. Additionally
International Nuclear Information System (INIS)
Highlights: ► To validate the SRAC2006 code system for TRIGA neutronics calculations. ► TRX and BAPL lattices are treated as standard benchmarks for this purpose. ► To compare the calculated results with experiment as well as MCNP values in this study. ► The study demonstrates a good agreement with the experiment and the MCNP results. ► Thus, this analysis reflects the validation study of the SRAC2006 code system. - Abstract: The goal of this study is to present the validation study of the SRAC2006 code system based on evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3 for neutronics analysis of TRIGA Mark-II Research Reactor at AERE, Bangladesh. This study is achieved through the analysis of integral parameters of TRX and BAPL benchmark lattices of thermal reactors. In integral measurements, the thermal reactor lattices TRX-1, TRX-2, BAPL-UO2-1, BAPL-UO2-2 and BAPL-UO2-3 are treated as standard benchmarks for validating/testing the SRAC2006 code system as well as nuclear data libraries. The integral parameters of the said lattices are calculated using the collision probability transport code PIJ of the SRAC2006 code system at room temperature 20 °C based on the above libraries. The calculated integral parameters are compared to the measured values as well as the MCNP values based on the Chinese evaluated nuclear data library CENDL-3.0. It was found that in most cases, the values of integral parameters demonstrate a good agreement with the experiment and the MCNP results. In addition, the group constants in SRAC format for TRX and BAPL lattices in fast and thermal energy range respectively are compared between the above libraries and it was found that the group constants are identical with very insignificant difference. Therefore, this analysis reflects the validation study of the SRAC2006 code system based on evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 and can also be essential to implement further neutronics calculations of
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.)
Shielding calculational system for plutonium
International Nuclear Information System (INIS)
A computer calculational system has been developed and assembled specifically for calculating dose rates in AEC plutonium fabrication facilities. The system consists of two computer codes and all nuclear data necessary for calculation of neutron and gamma dose rates from plutonium. The codes include the multigroup version of the Battelle Monte Carlo code for solution of general neutron and gamma shielding problems and the PUSHLD code for solution of shielding problems where low energy gamma and x-rays are important. The nuclear data consists of built in neutron and gamma yields and spectra for various plutonium compounds, an automatic calculation of age effects and all cross-sections commonly used. Experimental correlations have been performed to verify portions of the calculational system. (23 tables, 7 figs, 16 refs) (U.S.)
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)
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.
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)
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.
International Nuclear Information System (INIS)
The project encompasses the following project tasks and problems: (1) Studies relating to complete failure of the main heat transfer system; (2) Pebble flow; (3) Development of computer codes for detailed calculation of hypothetical accidents; (a) the THERMIX/RZKRIT temperature buildup code (covering a.o. a variation to include exothermal heat sources); (b) the REACT/THERMIX corrosion code (variation taking into account extremely severe air ingress into the primary loop); (c) the GRECO corrosion code (variation for treating extremely severe water ingress into the primary loop); (d) the KIND transients code (for treating extremely fast transients during reactivity incidents. (4) Limiting devices for safety-relevant quantities. (5) Analyses relating to hypothetical accidents. (a) hypothetical air ingress; (b) effects on the fuel particles induced by fast transients. The problems of the various tasks are defined in detail and the main results obtained are explained. The contributions reporting the various project tasks and activities have been prepared for separate retrieval from the database. (orig./HP)
International Nuclear Information System (INIS)
The KDF9/EGDON programme FADDEEV has been written to investigate a technique for the calculation of the matrix of frequency responses G(jw) describing the response of the output vector y from the multivariable differential/algebraic system S to the drive of the system input vector u. S: Ex = Ax + Bu, y = Cx, G(jw) = C(jw E - A )-1 B. The programme uses an algorithm due to Faddeev and has been written with emphasis upon: (a) simplicity of programme structure and computational technique which should enable a user to find his way through the programme fairly easily, and hence facilitate its manipulation as a subroutine in a larger code; (b) rapid computational ability, particularly in systems with fairly large number of inputs and outputs and requiring the evaluation of the frequency responses at a large number of frequencies. Transport or time delays must be converted by the user to Pade or Bode approximations prior to input. Conditions under which the algorithm fails to give accurate results are identified, and methods for increasing the accuracy of the calculations are discussed. The conditions for accurate results using FADDEEV indicate that its application is specialized. (author)
International Nuclear Information System (INIS)
In order to evaluate core characteristics of fast reactors, a computer code system ARCADIAN-FBR has been developed by utilizing the existing analysis codes and the latest nuclear data library JENDL-3.3. The validity of ARCADIAN-FBR was verified by using the experimental data obtained in the MONJU core physics tests. The results of analyses are in good agreement with the experimental data and the applicability of ARCADIAN-FBR for fast reactor core analysis is confirmed. Using ARCADIAN-FBR, the sodium void reactivity worth, which is an important parameter in the safety analysis of fast reactors, was analyzed for MONJU core. 241Pu in the core fuel is transmuted to 241Am due to disintegrations. Therefore, the effect of 241Am accumulation on the sodium void reactivity worth was evaluated for MONJU core. As a result of calculation, it was confirmed that the accumulation of 241Am significantly influences on the sodium void reactivity worth and hence on the safety analysis of sodium-cooled fast reactors. (author)
MOx benchmark calculations by deterministic and Monte Carlo codes
International Nuclear Information System (INIS)
Highlights: ► MOx based depletion calculation. ► Methodology to create continuous energy pseudo cross section for lump of minor fission products. ► Mass inventory comparison between deterministic and Monte Carlo codes. ► Higher deviation was found for several isotopes. - Abstract: A depletion calculation benchmark devoted to MOx fuel is an ongoing objective of the OECD/NEA WPRS following the study of depletion calculation concerning UOx fuels. The objective of the proposed benchmark is to compare existing depletion calculations obtained with various codes and data libraries applied to fuel and back-end cycle configurations. In the present work the deterministic code NEWT/ORIGEN-S of the SCALE6 codes package and the Monte Carlo based code MONTEBURNS2.0 were used to calculate the masses of inventory isotopes. The methodology to apply the MONTEBURNS2.0 to this benchmark is also presented. Then the results from both code were compared.
Accumulative Landings System Code Tables
National Oceanic and Atmospheric Administration, Department of Commerce — Code Tables Used In Landings System. These tables assign meanings to the codes that appear in the data tables. Code tables exist for species, gear, state, county,...
CONDOR: neutronic code for fuel elements calculation with rods
International Nuclear Information System (INIS)
CONDOR neutronic code is used for the calculation of fuel elements formed by fuel rods. The method employed to obtain the neutronic flux is that of collision probabilities in a multigroup scheme on two-dimensional geometry. This code utilizes new calculation algorithms and normalization of such collision probabilities. Burn-up calculations can be made before the alternative of applying variational methods for response flux calculations or those corresponding to collision normalization. (Author)
Xenon poisoning calculation code for miniature neutron source reactor (MNSR)
International Nuclear Information System (INIS)
In line with the actual requirements and based upon the specific characteristics of MNSR, a revised point-reactor model was adopted to model MNSR's xenon poisoning. The corresponding calculation code, MNSRXPCC (Xenon Poisoning Calculation Code for MNSR), was developed and tested by the Shanghai MNSR data
Xenon poisoning calculation code for miniature neutron source reactor (MNSR)
Institute of Scientific and Technical Information of China (English)
无
2001-01-01
In line with the actual requirements and based upon the specific char acteristics of MNSR, a revised point-reactor model was adopted to model MNSR's xenon poisoning. The corresponding calculation code, MNSRXPCC (Xenon Poison ing Calculation Code for MNSR), was developed and tested by the Shanghai MNSR data.
TEA: A Code for Calculating Thermochemical Equilibrium Abundances
Blecic, Jasmina; Harrington, Joseph; Bowman, M. Oliver
2015-01-01
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. (1958) and Eriksson (1971). 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 o...
Usage of burnt fuel isotopic compositions from engineering codes in Monte-Carlo code calculations
Energy Technology Data Exchange (ETDEWEB)
Aleshin, Sergey S.; Gorodkov, Sergey S.; Shcherenko, Anna I. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)
2015-09-15
A burn-up calculation of VVER's cores by Monte-Carlo code is complex process and requires large computational costs. This fact makes Monte-Carlo codes usage complicated for project and operating calculations. Previously prepared isotopic compositions are proposed to use for the Monte-Carlo code (MCU) calculations of different states of VVER's core with burnt fuel. Isotopic compositions are proposed to calculate by an approximation method. The approximation method is based on usage of a spectral functionality and reference isotopic compositions, that are calculated by engineering codes (TVS-M, PERMAK-A). The multiplication factors and power distributions of FA and VVER with infinite height are calculated in this work by the Monte-Carlo code MCU using earlier prepared isotopic compositions. The MCU calculation data were compared with the data which were obtained by engineering codes.
Calculation code PULCO for Purex process in pulsed column
International Nuclear Information System (INIS)
The calculation code PULCO, which can simulate the Purex process using a pulsed column as an extractor, has been developed. The PULCO is based on the fundamental concept of mass transfer that the mass transfer within a pulsed column occurs through the interface of liquid drops and continuous phase fluid, and is the calculation code different from conventional ones, by which various phenomena such as the generation of liquid drops, their rising and falling, and the unification of liquid drops actually occurring in a pulsed column are exactly reflected and can be correctly simulated. In the PULCO, the actually measured values of the fundamental quantities representing the extraction behavior of liquid drops in a pulsed column are incorporated, such as the mass transfer coefficient of each component, the diameter and velocity of liquid drops in a pulsed column, the holdup of dispersed phase, and axial turbulent flow diffusion coefficient. The verification of the results calculated with the PULCO was carried out by installing a pulsed column of 50 mm inside diameter and 2 m length with 40 plate stage in a glove box for unirradiated uranium-plutonium mixed system. The results of the calculation and test were in good agreement, and the validity of the PULCO was confirmed. (Kako, I.)
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
Development and validation of a nodal code for core calculation
International Nuclear Information System (INIS)
The code RHENO solves the multigroup three-dimensional diffusion equation using a nodal method of polynomial expansion.A comparative study has been made between this code and present internationals nodal diffusion codes, resulting that the RHENO is up to date.The RHENO has been integrated to a calculation line and has been extend to make burnup calculations.Two methods for pin power reconstruction were developed: modulation and imbedded. The modulation method has been implemented in a program, while the implementation of the imbedded method will be concluded shortly.The validation carried out (that includes experimental data of a MPR) show very good results and calculation efficiency
TEA: A Code for Calculating Thermochemical Equilibrium Abundances
Blecic, Jasmina; Bowman, M Oliver
2015-01-01
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. (1958) and Eriksson (1971). 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 & Sharp (1999), the free thermochemical equilibrium code CEA (Chemical Equilibrium with Applications), and the example given by White et al. (1958). 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 ...
Burnup calculations using serpent code in accelerator driven thorium reactors
International Nuclear Information System (INIS)
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 232Th and mixed 233U 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.)
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.
Calculation codes in radiation protection, radiation physics and dosimetry
International Nuclear Information System (INIS)
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.)
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)
Description of the CAREM Reactor Neutronic Calculation Codes
International Nuclear Information System (INIS)
In this work is described the neutronic calculation line used to design the CAREM reactor.A description of the codes used and the interfaces between the different programs are presented.Both, the normal calculation line and the alternative or verification calculation line are included.The calculation line used to obtain the kinetics parameters (effective delayed-neutron fraction and prompt-neutron lifetime) is also included
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.)
International Nuclear Information System (INIS)
A fast running computer code SHETEMP has been developed for analysis of reactivity initiated accidents under constant core cooling conditions such as coolant temperature and heat transfer coefficient on fuel rods. This code can predict core power and fuel temperature behaviours. A control rod movement can be taken into account in power control system. The objective of the code is to provide fast running capability with easy handling of the code required for audit and design calculations where a large number of calculations are performed for parameter surveys during short time period. The fast running capability of the code was realized by neglection of fluid flow calculation. The computer code SHETEMP was made up by extracting and conglomerating routines for reactor kinetics and heat conduction in the transient reactor thermal-hydraulic analysis code ALARM-P1, and by combining newly developed routines for reactor power control system. As ALARM-P1, SHETEMP solves point reactor kinetics equations by the modified Runge-Kutta method and one-dimensional transient heat conduction equations for slab and cylindrical geometries by the Crank-Nicholson methods. The model for reactor power control system takes into account effects of PID regulator and control rod drive mechanism. In order to check errors in programming of the code, calculated results by SHETEMP were compared with analytic solution. Based on the comparisons, the appropriateness of the programming was verified. Also, through a sample calculation for typical modelling, it was concluded that the code could satisfy the fast running capability required for audit and design calculations. This report will be described as a code manual of SHETEMP. It contains descriptions on a sample problem, code structure, input data specifications and usage of the code, in addition to analytical models and results of code verification calculations. (author)
CONSUL code package application for LMFR core calculations
Energy Technology Data Exchange (ETDEWEB)
Chibinyaev, A.V.; Teplov, P.S.; Frolova, M.V. [RNC ' Kurchatovskiy institute' , Kurchatov sq.1, Moscow (Russian Federation)
2008-07-01
CONSUL code package designed for the calculation of reactor core characteristics has been developed at the beginning of 90's. The calculation of nuclear reactor core characteristics is carried out on the basis of correlated neutron, isotope and temperature distributions. The code package has been generally used for LWR core characteristics calculations. At present CONSUL code package was adapted to calculate liquid metal fast reactors (LMFR). The comparisons with IAEA computational test 'Evaluation of benchmark calculations on a fast power reactor core with near zero sodium void effect' and BN-1800 testing calculations are presented in the paper. The IAEA benchmark core is based on the innovative core concept with sodium plenum above the core BN-800. BN-1800 core is the next development step which is foreseen for the Russian fast reactor concept. The comparison of the operational parameters has shown good agreement and confirms the possibility of CONSUL code package application for LMFR core calculation. (authors)
Validation of the Monteburns code for criticality calculation of TRIGA reactors
Energy Technology Data Exchange (ETDEWEB)
Dalle, Hugo Moura [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil); Jeraj, Robert [Jozef Stafan Institute, Ljubljana (Slovenia)
2002-07-01
Use of Monte Carlo methods in burnup calculations of nuclear fuel has become practical due to increased speed of computers. Monteburns is an automated computational tool that links the Monte Carlo code MCNP with the burnup and decay code ORIGEN2.1. This code system was used to simulate a criticality benchmark experiment with burned fuel on a TRIGA Mark II research reactor. Two core configurations were simulated and k{sub eff} values calculated. The comparison between the calculated and experimental values shows good agreement, which indicates that the MCNP/Monteburns/ORIGEN2.1 system gives reliable results for neutronic simulations of TRIGA reactors. (author)
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.
The MCEF code for nuclear evaporation and fission calculations
Energy Technology Data Exchange (ETDEWEB)
Deppman, A.; Pina, S.R. de; Likhachev, V.P.; Mesa, J. [Sao Paulo Univ., SP (Brazil). Inst. de Fisica; Tavares, O.A.P.; Duarte, S.B.; Oliveira, E.C. de [Centro Brasileiro de Pesquisas Fisicas (CBPF), Rio de Janeiro, RJ (Brazil); Arruda-Neto, J.D.T. [Universidade Santo Amaro (UNISA), SP (Brazil); Rodriguez, O. [Instituto Superior de Ciencias y Tecnologia Nucleares, La Habana (Cuba); Goncalves, M. [Instituto de Radioprotecao e Dosimetria (IRD), Rio de Janeiro, RJ (Brazil)
2001-11-01
We present an object oriented algorithm, written in the Java programming language, which performs a Monte Carlo calculation of the evaporation-fission process taking place inside an excited nucleus. We show that this nuclear physics problem is very suited for the object oriented programming by constructing two simple objects: one that handles all nuclear properties and another that takes care of the nuclear reaction. The MCEF code was used to calculate important results for nuclear reactions, and here we show examples of possible uses for this code. (author)
The MCEF code for nuclear evaporation and fission calculations
International Nuclear Information System (INIS)
We present an object oriented algorithm, written in the Java programming language, which performs a Monte Carlo calculation of the evaporation-fission process taking place inside an excited nucleus. We show that this nuclear physics problem is very suited for the object oriented programming by constructing two simple objects: one that handles all nuclear properties and another that takes care of the nuclear reaction. The MCEF code was used to calculate important results for nuclear reactions, and here we show examples of possible uses for this code. (author)
Elements of algebraic coding systems
Cardoso da Rocha, Jr, Valdemar
2014-01-01
Elements of Algebraic Coding Systems is an introductory textto algebraic coding theory. In the first chapter, you'll gain insideknowledge of coding fundamentals, which is essential for a deeperunderstanding of state-of-the-art coding systems.This book is a quick reference for those who are unfamiliar withthis topic, as well as for use with specific applications such as cryptographyand communication. Linear error-correcting block codesthrough elementary principles span eleven chapters of the text.Cyclic codes, some finite field algebra, Goppa codes, algebraic decodingalgorithms, and applications in public-key cryptography andsecret-key cryptography are discussed, including problems and solutionsat the end of each chapter. Three appendices cover the Gilbertbound 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 finitefield Fourier transform and the Euclidean algorithm for polynomials.
Kumar, A.; Graves, R. A., Jr.; Weilmuenster, K. J.
1980-01-01
A vectorized code, EQUIL, was developed for calculating the equilibrium chemistry of a reacting gas mixture on the Control Data STAR-100 computer. The code provides species mole fractions, mass fractions, and thermodynamic and transport properties of the mixture for given temperature, pressure, and elemental mass fractions. The code is set up for the electrons H, He, C, O, N system of elements. In all, 24 chemical species are included.
Blind Recognition Algorithm of Turbo Codes for Communication Intelligence Systems
Directory of Open Access Journals (Sweden)
Ali Naseri
2011-11-01
Full Text Available Turbo codes are widely used in land and space radio communication systems, and because of complexity of structure, are custom in military communication systems. In electronic warfare, COMINT systems make attempt to recognize codes by blind ways. In this Paper, the algorithm is proposed for blind recognition of turbo code parameters like code kind, code-word length, code rate, length of interleaver and delay blocks number of convolution code. The algorithm calculations volume is0.5L3+1.25L, therefore it is suitable for real time systems.
Decay heat calculation an international nuclear code comparison
International Nuclear Information System (INIS)
The results of an international code comparison on decay heat are presented and discussed. Participants from more than ten laboratories calculated, using the same input data, decay heat for thirteen cooling times between 1 and 1013 sec. Two irradiation cases were proposed: fission pulse and 3x107 seconds of irradiation of 235U fuel. The results are analysed and compared. This inter-comparison shows that, if the same input data are given, most of the codes give very similar results for the decay heat and consequently also for the fission product contribution
Progress on burnup calculation methods coupling Monte Carlo and depletion codes
Energy Technology Data Exchange (ETDEWEB)
Leszczynski, Francisco [Comision Nacional de Energia Atomica, San Carlos de Bariloche, RN (Argentina). Centro Atomico Bariloche]. E-mail: lesinki@cab.cnea.gob.ar
2005-07-01
Several methods of burnup calculations coupling Monte Carlo and depletion codes that were investigated and applied for the author last years are described. here. Some benchmark results and future possibilities are analyzed also. The methods are: depletion calculations at cell level with WIMS or other cell codes, and use of the resulting concentrations of fission products, poisons and actinides on Monte Carlo calculation for fixed burnup distributions obtained from diffusion codes; same as the first but using a method o coupling Monte Carlo (MCNP) and a depletion code (ORIGEN) at a cell level for obtaining the concentrations of nuclides, to be used on full reactor calculation with Monte Carlo code; and full calculation of the system with Monte Carlo and depletion codes, on several steps. All these methods were used for different problems for research reactors and some comparisons with experimental results of regular lattices were performed. On this work, a resume of all these works is presented and discussion of advantages and problems found are included. Also, a brief description of the methods adopted and MCQ system for coupling MCNP and ORIGEN codes is included. (author)
Blind Recognition Algorithm of Turbo Codes for Communication Intelligence Systems
Ali Naseri; Omid Azmoon; Samad Fazeli
2011-01-01
Turbo codes are widely used in land and space radio communication systems, and because of complexity of structure, are custom in military communication systems. In electronic warfare, COMINT systems make attempt to recognize codes by blind ways. In this Paper, the algorithm is proposed for blind recognition of turbo code parameters like code kind, code-word length, code rate, length of interleaver and delay blocks number of convolution code. The algorithm calculations volume is0.5L3+1.25L, th...
TEMP: a computer code to calculate fuel pin temperatures during a transient
International Nuclear Information System (INIS)
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
How Accurately can we Calculate Thermal Systems?
Energy Technology Data Exchange (ETDEWEB)
Cullen, D; Blomquist, R N; Dean, C; Heinrichs, D; Kalugin, M A; Lee, M; Lee, Y; MacFarlan, R; Nagaya, Y; Trkov, A
2004-04-20
I would like to determine how accurately a variety of neutron transport code packages (code and cross section libraries) can calculate simple integral parameters, such as K{sub eff}, for systems that are sensitive to thermal neutron scattering. Since we will only consider theoretical systems, we cannot really determine absolute accuracy compared to any real system. Therefore rather than accuracy, it would be more precise to say that I would like to determine the spread in answers that we obtain from a variety of code packages. This spread should serve as an excellent indicator of how accurately we can really model and calculate such systems today. Hopefully, eventually this will lead to improvements in both our codes and the thermal scattering models that they use in the future. In order to accomplish this I propose a number of extremely simple systems that involve thermal neutron scattering that can be easily modeled and calculated by a variety of neutron transport codes. These are theoretical systems designed to emphasize the effects of thermal scattering, since that is what we are interested in studying. I have attempted to keep these systems very simple, and yet at the same time they include most, if not all, of the important thermal scattering effects encountered in a large, water-moderated, uranium fueled thermal system, i.e., our typical thermal reactors.
International Nuclear Information System (INIS)
The development of a high-accuracy reactor benchmark analysis capability is described. This capability has been incorporated into a revised and extended version of the lattice analysis program HAMMER. Previous analyses using the HAMMER program required the introduction of correction factors obtained from more rigorous treatments of various effects such as resonance capture and neutron leakage. The present version of the program will remove the ambiguities associated with the introduction of such correction factors by optionally performing the more rigorous calculations internally or by automating the correctional procedure
Analytical stress tensor and pressure calculations with the CRYSTAL code
Doll, Klaus
2010-01-01
Abstract The calculation of the stress tensor and related properties and its implementation in the CRYSTAL code are described. The stress tensor is obtained from the earlier implemented analytical gradients with respect to the cell parameters. Subsequently, the pressure and enthalpy is computed, and a test concerning the pressure driven phase transition in KI is used as an illustration. Finally, the possibility of applying external pressure is implemented. The ...
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.
International Nuclear Information System (INIS)
A tokamak systems code capable of modeling experimental test reactors has been developed and is described in this document. The code, named TETRA (for Tokamak Engineering Test Reactor Analysis), consists of a series of modules, each describing a tokamak system or component, controlled by an optimizer/driver. This code development was a national effort in that the modules were contributed by members of the fusion community and integrated into a code by the Fusion Engineering Design Center. The code has been checked out on the Cray computers at the National Magnetic Fusion Energy Computing Center and has satisfactorily simulated the Tokamak Ignition/Burn Experimental Reactor II (TIBER) design. A feature of this code is the ability to perform optimization studies through the use of a numerical software package, which iterates prescribed variables to satisfy a set of prescribed equations or constraints. This code will be used to perform sensitivity studies for the proposed International Thermonuclear Experimental Reactor (ITER). 22 figs., 29 tabs
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)
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).
Comparison of computer code calculations with FEBA test data
International Nuclear Information System (INIS)
The FEBA forced feed reflood experiments included base line tests with unblocked geometry. The experiments consisted of separate effect tests on a full-length 5x5 rod bundle. Experimental cladding temperatures and heat transfer coefficients of FEBA test No. 216 are compared with the analytical data postcalculated utilizing the SSYST-3 computer code. The comparison indicates a satisfactory matching of the peak cladding temperatures, quench times and heat transfer coefficients for nearly all axial positions. This agreement was made possible by the use of an artificially adjusted value of the empirical code input parameter in the heat transfer for the dispersed flow regime. A limited comparison of test data and calculations using the RELAP4/MOD6 transient analysis code are also included. In this case the input data for the water entrainment fraction and the liquid weighting factor in the heat transfer for the dispersed flow regime were adjusted to match the experimental data. On the other hand, no fitting of the input parameters was made for the COBRA-TF calculations which are included in the data comparison. (orig.)
Comparison of code calculations with experiments on containment response during LOCA conditions
International Nuclear Information System (INIS)
A series of experiments were performed on a one-tenth scale model of PHWR containment, incorporating pressure suppression system. The pressure-temperature transients in the model containment observed during simulated LOCA (Loss of Coolant) blowdown conditions were compared against calculated results form computer code PACSR, for purposes of verification of the code. Comparison of results indicated that calculated values of peak pressure in various compartment were significantly higher than observed ones. This disagreement was attributed mainly to modelling for energy absorption from containment atmosphere to structural surfaces, this effect being particularly important in a scaled down model. Good agreement between calculation and experiment was obtained after heat transfer correlation for energy absorption on surfaces were modified in the code. The study demonstrates the conservatism of the results from the code. (author). 6 refs., 1 tab., 9 figs
Tokamak plasma power balance calculation code (TPC code) outline and operation manual
International Nuclear Information System (INIS)
This report is a detailed description on the TPC code, that calculates the power balance of a tokamak plasma according to the ITER guidelines. The TPC code works on a personal computer (Macintosh or J-3100/ IBM-PC). Using input data such as the plasma shape, toroidal magnetic field, plasma current, electron temperature, electron density, impurities and heating power, TPC code can determine the operation point of the fusion reactor (Ion temperature is assumed to be equal to the electron temperature). Supplied flux (Volt · sec) and burn time are also estimated by coil design parameters. Calculated energy confinement time is compared with various L-mode scaling laws and the confinement enhancement factor (H-factor) is evaluated. Divertor heat load is predicted by using simple scaling models (constant-χ, Bohm-type-χ and JT-60U empirical scaling models). Frequently used data can be stored in a 'device file' and used as the default values. TPC code can generate 2-D mesh data and the POPCON plot is drawn by a contour line plotting program (CONPLT). The operation manual about CONPLT code is also described. (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.
Adjoint Monte Carlo techniques and codes for organ dose calculations
International Nuclear Information System (INIS)
Adjoint Monte Carlo simulations can be effectively used for the estimation of doses in small targets when the sources are extended in large volumes or surfaces. The main features of two computer codes for calculating doses at free points or in organs of an anthropomorphic phantom are described. In the first program (REBEL-3) natural gamma-emitting sources are contained in the walls of a dwelling room; in the second one (POKER-CAMP) the user can specify arbitrary gamma sources with different spatial distributions in the environment: in (or on the surface of) the ground and in the air. 3 figures
Code system for fast reactor neutronics analysis
International Nuclear Information System (INIS)
A code system for analysis of fast reactor neutronics has been developed for the purpose of handy use and error reduction. The JOINT code produces the input data file to be used in the neutronics calculation code and also prepares the cross section library file with an assigned format. The effective cross sections are saved in the PDS file with an unified format. At the present stage, this code system includes the following codes; SLAROM, ESELEM5, EXPANDA-G for the production of effective cross sections and CITATION-FBR, ANISN-JR, TWOTRAN2, PHENIX, 3DB, MORSE, CIPER and SNPERT. In the course of the development, some utility programs and service programs have been additionaly developed. These are used for access of PDS file, edit of the cross sections and graphic display. Included in this report are a description of input data format of the JOINT and other programs, and of the function of each subroutine and utility programs. The usage of PDS file is also explained. In Appendix A, the input formats are described for the revised version of the CIPER code. (author)
Development of the Joyo MK-II core bowing reactivity calculation code
Energy Technology Data Exchange (ETDEWEB)
Tabuchi, Shiro; Torimaru, Tadahiko; Yoshida, Akihiro; Aoyama, Takafumi [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center
1999-09-01
The study on the passive safety test by using the Experimental Fast Reactor Joyo has been performed to demonstrate the inherent safety of fast breeder reactors. In this study, emphasis was placed on the improvement on the accuracy of the feedback reactivity analysis. As a bowing reactivity might play a significant roll in ATWS analysis because of its effectively short time constant and relatively large magnitude, an emphasis was placed upon the evaluation of the analysis precision of bowing reactivity. Taking into account of the refueling and irradiation history of the individual core component, the core bowing behavior in Joyo has been analyzed by using the MK-II core management code system MAGI, the interface code TETRAS which interpolate neutron flux and coolant temperature at the position of wrapper tube, and the core bowing calculation code BEACON. Calculation accuracy of above mentioned system was evaluated through the comparison of calculated and measured permanent distortion of subassemblies. In 1996, core bowing reactivity was calculated by AURORA code using the above calculated bowing behavior of individual core component as input. But because an approximate two dimensional material reactivity worth map was utilized in AURORA, it was made clear that some amount of error caused by extrapolation could not be neglected. Therefore calculation code ARCHCOM (Analysis of Reactivity Change due to Core Mechanics) which utilize three dimensional material reactivity worth map as input was developed for the Joyo MK-II core bowing reactivity calculation. This code reduces above mentioned extrapolation error that used to be occurred at isolated core component, such as control rod or irradiation rig and at the interface region between fuel and reflector which had sharp bowing reactivity worth gradient. (author)
Directory of Open Access Journals (Sweden)
Asghar Mesbahi
2015-09-01
Full Text Available Introduction Radiotherapy with small fields is used widely in newly developed techniques. Additionally, dose calculation accuracy of treatment planning systems in small fields plays a crucial role in treatment outcome. In the present study, dose calculation accuracy of two commercial treatment planning systems was evaluated against Monte Carlo method. Materials and Methods Siemens Once or linear accelerator was simulated, using MCNPX Monte Carlo code, according to manufacturer’s instructions. Three analytical algorithms for dose calculation including full scatter convolution (FSC in TiGRT, along with convolution and superposition in XiO system were evaluated for a small solid liver tumor. This solid tumor with a diameter of 1.8 cm was evaluated in a thorax phantom, and calculations were performed for different field sizes (1×1, 2×2, 3×3 and4×4 cm2. The results obtained in these treatment planning systems were compared with calculations by MC method (regarded as the most reliable method. Results For FSC and convolution algorithm, comparison with MC calculations indicated dose overestimations of up to 120%and 25% inside the lung and tumor, respectively in 1×1 cm2field size, using an 18 MV photon beam. Regarding superposition, a close agreement was seen with MC simulation in all studied field sizes. Conclusion The obtained results showed that FSC and convolution algorithm significantly overestimated doses of the lung and solid tumor; therefore, significant errors could arise in treatment plans of lung region, thus affecting the treatment outcomes. Therefore, use of MC-based methods and super position is recommended for lung treatments, using small fields and beamlets.
Calculation code evaluating the confinement of a nuclear facility in case of fires
Energy Technology Data Exchange (ETDEWEB)
Laborde, J.C.; Prevost, C.; Vendel, J. [and others
1995-02-01
Accident events involving fire are quite frequent and could have a severe effect on the safety of nuclear facilities. As confinement must be maintained, the ventilation and filtration systems have to be designed to limit radioactive release to the environment. To determine and analyse the consequences of a fire on the contamination confinement, IPSN, COGEMA and SGN are participating in development of a calculation code based on introduction, in the SIMEVENT ventilation code, of various models associated to fire risk and mass transfer in the ventilation networks. This calculation code results from the coupling of the SIMEVENT code with several models describing the temperature in a room resulting of a fire, the temperatures along the ventilation ducts, the contamination transfers through out the ventilation equipments (ducts, dampers, valves, air cleaning systems) and the High Efficiency Particulate Air (HEPA) filters clogging. The paper proposed presents the current level of progress in development of this calculation code. It describes, in particular, the empirical model used for the clogging of HEPA filters by the aerosols derived from the combustion of standard materials used in the nuclear industry. It describes, also, the specific models used to take into account the mass transfers and resulting from the basic mechanisms of aerosols physics. In addition, an assessment of this code is given using the example of a simple laboratory installation.
Fragmentation calculation by intranuclear-cascade-evaporation code
Energy Technology Data Exchange (ETDEWEB)
Shigyo, Nobuhiro; Iga, Kiminori; Ishibashi, Kenji [Kyushu Univ., Fukuoka (Japan)
1997-03-01
High Energy Transport Code (HETC) based on the intranuclear-cascade-evaporation model is modified for calculating the fragmentation cross section. For the intranuclear-cascade process, nucleon-nucleon cross sections are used for collision computation; effective in-medium-corrected cross sections are adopted instead of the original free-nucleon collision. The exciton model is adopted for improvement of backward nucleon-emission cross section for low-energy nucleon-incident events. The fragmentation reaction is incorporated into the original HETC as a subroutine set by the use of the systematics of the reaction. The modified HETC (HETC-3STEP/FRG) reproduces experimental fragment yields to a reasonable degree. (author)
Module type plant system dynamics analysis code (MSG-COPD). Code manual
International Nuclear Information System (INIS)
MSG-COPD is a module type plant system dynamics analysis code which involves a multi-dimensional thermal-hydraulics calculation module to analyze pool type of fast breeder reactors. Explanations of each module and the methods for the input data are described in this code manual. (author)
Development of neutral transport lattice code DENT-2D and benchmark calculation
Energy Technology Data Exchange (ETDEWEB)
Kim, K. S.; Kim, H. Y.; Ji, S. K. [KAERI, Taejon (Korea, Republic of)
2002-05-01
We developed new transport lattice code called DENT-2D (Deterministic Neutral Particle Transport Code in 2-D imensional Space)primarily to generate few- group constants for the reactor physics analysis diffusion codes. This code is designed to be coupled with KAERI reactor analysis nodal code, MASTER [1] ,to complete the design system package. CASMO-3 and HELIOS have been used in generating the few- group constant for MASTER. Currently DENT-2D includes only neutron particle transport calculation in 2-dimensional Cartesian geometry. The characteristics method is adopted for the spatial discretization, which is advantageous for the treatment of the complicated geometry structure and the highly anisotropic scattering. The subgroup method is used for the resonance treatment. B1 approximation has been used to obtain the criticality spectrum considering the leakage effect in the real core situation. The exponential matrix method has been used for the depletion calculation. The results of benchmark calculations show that the prediction capability of DENT-2D is comparable to the other lattice codes such as HELIOS and CASMO-3.
Kumawat, Soma; Ravi Kumar, M.
2016-07-01
Double Weight (DW) code family is one of the coding schemes proposed for Spectral Amplitude Coding-Optical Code Division Multiple Access (SAC-OCDMA) systems. Modified Double Weight (MDW) code for even weights and Enhanced Double Weight (EDW) code for odd weights are two algorithms extending the use of DW code for SAC-OCDMA systems. The above mentioned codes use mapping technique to provide codes for higher number of users. A new generalized algorithm to construct EDW and MDW like codes without mapping for any weight greater than 2 is proposed. A single code construction algorithm gives same length increment, Bit Error Rate (BER) calculation and other properties for all weights greater than 2. Algorithm first constructs a generalized basic matrix which is repeated in a different way to produce the codes for all users (different from mapping). The generalized code is analysed for BER using balanced detection and direct detection techniques.
Computer access security code system
Collins, Earl R., Jr. (Inventor)
1990-01-01
A security code system for controlling access to computer and computer-controlled entry situations comprises a plurality of subsets of alpha-numeric characters disposed in random order in matrices of at least two dimensions forming theoretical rectangles, cubes, etc., such that when access is desired, at least one pair of previously unused character subsets not found in the same row or column of the matrix is chosen at random and transmitted by the computer. The proper response to gain access is transmittal of subsets which complete the rectangle, and/or a parallelepiped whose opposite corners were defined by first groups of code. Once used, subsets are not used again to absolutely defeat unauthorized access by eavesdropping, and the like.
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.
Energy Technology Data Exchange (ETDEWEB)
Proskuryakov, K.N.; Bogomazov, D.N.; Poliakov, N. [Moscow Power Engineering Institute (Technical University), Moscow (Russian Federation)
2007-07-01
The new special module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation is worked out. The Russian computer code Rainbow has been selected for joint use with a developed module. This code system provides the possibility of EFOCP (Eigen Frequencies of Oscillations of the Coolant Pressure) calculations in any coolant acoustical elements of primary circuits of NPP. EFOCP values have been calculated for transient and for stationary operating. The calculated results for nominal operating were compared with results of measured EFOCP. For example, this comparison was provided for the system: 'pressurizer + surge line' of a WWER-1000 reactor. The calculated result 0.58 Hz practically coincides with the result of measurement (0.6 Hz). The EFOCP variations in transients are also shown. The presented results are intended to be useful for NPP vibration-acoustical certification. There are no serious difficulties for using this module with other computer codes.
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.
GOBLIN computer code. Comparison between calculations and TLTA small break test
International Nuclear Information System (INIS)
GOBLIN calcuations have been performed for two simulation tests of the boiling water reactor (BWR) small break loss-of-coolant accidents (LOCAs) which were conducted in the two loop test apparatus (TLTA). The first test investigated the small break with nondegraded emergency core coolant (ECC) systems and the second test studied the same small break but with degraded ECC systems in which the high pressure core spray (HPCS) was assumed unavailable. Very good agreement between test data and calculations is achieved. The second test is the most challenging from code comparison point of view and the code prediction of the complicated mass distribution pattern which changes with time is very satisfactory. In the first test and to some extent late in the second test multidimensional subchannel effects are evident in the core bundle region. These are not and cannot be reproduced by the code since the bundle model of GOBLIN is strictly one-dimensional. (Author)
Comparison of computer codes for calculating dynamic loads in wind turbines
Spera, D. A.
1978-01-01
The development of computer codes for calculating dynamic loads in horizontal axis wind turbines was examined, and a brief overview of each code was given. The performance of individual codes was compared against two sets of test data measured on a 100 KW Mod-0 wind turbine. All codes are aeroelastic and include loads which are gravitational, inertial and aerodynamic in origin.
Expansion of the CHR bone code system
International Nuclear Information System (INIS)
This report describes the coding system used in the Center for Human Radiobiology (CHR) to identify individual bones and portions of bones of a complete skeletal system. It includes illustrations of various bones and bone segments with their respective code numbers. Codes are also presented for bone groups and for nonbone materials
A guide to the AUS modular neutronics code system
International Nuclear Information System (INIS)
A general description is given of the AUS modular neutronics code system, which may be used for calculations of a very wide range of fission reactors, fusion blankets and other neutron applications. The present system has cross-section libraries derived from ENDF/B-IV and includes modules which provide for lattice calculations, one-dimensional transport calculations, and one, two, and three-dimensional diffusion calculations, burnup calculations and the flexible editing of results. Details of all system aspects of AUS are provided but the major individual modules are only outlined. Sufficient information is given to enable other modules to be added to the system
Development of tokamak reactor system analysis code NEW-TORSAC
Kasai, Masao; Ida, Toshio; Nishikawa, Masana; Kameari, Akihisa; Nishio, Satoshi; Tone, Tatsuzo
1987-07-01
A systems analysis code named NEW-TORSAC (TOkamak Reactor Systems Analysis Code) has been developed by modifying the TORSAC which had been already developed by us. The NEW-TORSAC is available for tokamak reactor designs and evaluations from experimental machines to commercial reactor plants. It has functions to design tokamaks automatically from plasma parameter setting to determining configurations of reactor equipments and calculating main characteristics parameters of auxiliary systems and the capital costs. In the case of analyzing tokamak reactor plants, the code can calculate busbar energy costs. In addition to numerical output, some output of this code such as a reactor configuration, plasma equilibrium, electro-magnetic forces, etc., are graphically displayed. The code has been successfully applied to the scoping studies of the next generation machines and commercial reactor plants.
14 CFR 234.8 - Calculation of on-time performance codes.
2010-01-01
... (AVIATION PROCEEDINGS) ECONOMIC REGULATIONS AIRLINE SERVICE QUALITY PERFORMANCE REPORTS § 234.8 Calculation of on-time performance codes. (a) Each reporting carrier shall calculate an on-time performance code... 14 Aeronautics and Space 4 2010-01-01 2010-01-01 false Calculation of on-time performance...
Tandem Mirror Reactor Systems Code (Version I)
International Nuclear Information System (INIS)
A computer code was developed to model a Tandem Mirror Reactor. Ths is the first Tandem Mirror Reactor model to couple, in detail, the highly linked physics, magnetics, and neutronic analysis into a single code. This report describes the code architecture, provides a summary description of the modules comprising the code, and includes an example execution of the Tandem Mirror Reactor Systems Code. Results from this code for two sensitivity studies are also included. These studies are: (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power; and (2) to determine the impact of reactor power level on cost
Characterizing Video Coding Computing in Conference Systems
Tuquerres, G.
2000-01-01
In this paper, a number of coding operations is provided for computing continuous data streams, in particular, video streams. A coding capability of the operations is expressed by a pyramidal structure in which coding processes and requirements of a distributed information system are represented. Th
Energy Technology Data Exchange (ETDEWEB)
1979-09-01
A methodology for analyzing the economic impact of WECS on a utility is described in Volume I of this report. The methodology requires extrapolating both historical utility load data and historical wind power into a year of analysis; calculating the total amount of funds made available in that year, as a result of the inclusion of wind power in the utility mix; and then estimating the present value of the total funds made available to the utility over the life of the WECS. To apply the methodology to a specific case, it was necessary to develop various computer programs. The following sections in this report list the programs developed for this study, briefly summarize their contents, and explain how they are used. Wherever possible, a typical input/output file is shown.
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.
NUFACE: An interface code for the calculation of nuclear responses
International Nuclear Information System (INIS)
The NUFACE interface code computes nuclear responses for use in the nuclear analysis of a given tokamak reactor design. The NUFACE code operates on the neutron and gamma fluxes provided by the one-dimensional neutral-particle transport code ONEDANT. Zonewise and zone-boundary responses are computed to obtain both zone-integrated values and maximum surface values. Information on each material mixture within a zone and on each element or isotope constituent of each material is computed. This feature allows for a detailed analysis of the reactor whereby one can easily identify the fractional contribution to the response of interest from each material and each element or isotope. 4 refs., 4 figs., 3 tabs
NUFACE: An interface code for the calculation of nuclear responses
Energy Technology Data Exchange (ETDEWEB)
Henderson, D.L. (Oak Ridge National Lab., TN (USA)); Gomes, I.C. (Tennessee Univ., Knoxville, TN (USA))
1990-01-01
The NUFACE interface code computes nuclear responses for use in the nuclear analysis of a given tokamak reactor design. The NUFACE code operates on the neutron and gamma fluxes provided by the one-dimensional neutral-particle transport code ONEDANT. Zonewise and zone-boundary responses are computed to obtain both zone-integrated values and maximum surface values. Information on each material mixture within a zone and on each element or isotope constituent of each material is computed. This feature allows for a detailed analysis of the reactor whereby one can easily identify the fractional contribution to the response of interest from each material and each element or isotope. 4 refs., 4 figs., 3 tabs.
Code system to compute radiation dose in human phantoms
International Nuclear Information System (INIS)
Monte Carlo photon transport code and a code using Monte Carlo integration of a point kernel have been revised to incorporate human phantom models for an adult female, juveniles of various ages, and a pregnant female at the end of the first trimester of pregnancy, in addition to the adult male used earlier. An analysis code has been developed for deriving recommended values of specific absorbed fractions of photon energy. The computer code system and calculational method are described, emphasizing recent improvements in methods
Accuracy evaluation of pin exposure calculations in current LWR core design codes
International Nuclear Information System (INIS)
The deregulated utility environment and better utilization of fuel assemblies in nuclear power plants has allowed designers to burn fuel assemblies to maximum allowable exposures. Any uncertainties, associated with the technical approach and numerical methods used to perform pin exposure calculations may cause either peak power exposure to exceed the Nuclear Regulatory Commission (NRC) exposure limit or lead to excessive conservatism and thus inefficient fuel utilization. In this work, a Monte Carlo based coupled depletion code (MCNP5/ORIGEN-S) is utilized to provide reference solutions in order to assess the accuracy of pin power and pin exposure reconstruction methods in the current commercial and licensed three-dimensional (3D) nodal Light Water Reactor (LWR) core design codes. The developed at Pennsylvania State University (PSU) MCNP5/ORIGEN-S coupled depletion code system was validated using measured data from the PSU TRIGA research reactor critical experiments. A number of test cases (modeling benchmarks) representative of LWRs were developed starting from the least complex model towards more complicated and more realistic models. The accuracy evaluation of the pin reconstruction methods was performed by using the CASMO-4 and SIMULATE-3 codes as the representative of current commercial LWR core design systems. Two-dimensional (2D) transport calculations with the TRITON module from the SCALE5 package were employed to produce the spectrum averaged cross-section libraries as a function of burnup for ORIGEN-S calculations. The burnup dependent cross-section libraries are specifically generated for each lattice configuration type. For the MCNP5 calculations continuous cross-section libraries for different isotopes at hot operating temperatures are generated and subsequently utilized. Realistic lattice configurations of the GE13 BWR fuel assemblies (unrodded and rodded) depleted under operating conditions were studied in this research because of their heterogeneous
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.
Development of tokamak reactor systems analysis code 'TORSAC'
International Nuclear Information System (INIS)
This report describes Tokamak Reactor Systems Analysis Code ''TORSAC'' which has been developed in order to assess the impact of the design choises on reactor systems and to improve tokamak designs in wide parameter range. This computer code has following functions. (1) Systematic sensitivity analysis for a set of given design parameters, (2) Cost calculation of a new reactor concept designed automatically as a result of systematic sensitivity analysis. (author)
A New, Efficient Stellar Evolution Code for Calculating Complete Evolutionary Tracks
Kovetz, Attay; Prialnik, Dina
2008-01-01
We present a new stellar evolution code and a set of results, demonstrating its capability at calculating full evolutionary tracks for a wide range of masses and metallicities. The code is fast and efficient, and is capable of following through all evolutionary phases, without interruption or human intervention. It is meant to be used also in the context of modeling the evolution of dense stellar systems, for performing live calculations for both normal star models and merger-products. The code is based on a fully implicit, adaptive-grid numerical scheme that solves simultaneously for structure, mesh and chemical composition. Full details are given for the treatment of convection, equation of state, opacity, nuclear reactions and mass loss. Results of evolutionary calculations are shown for a solar model that matches the characteristics of the present sun to an accuracy of better than 1%; a $1 \\Msun$ model for a wide range of metallicities; a series of models of stellar populations I and II, for the mass rang...
MUXS: a code to generate multigroup cross sections for sputtering calculations
International Nuclear Information System (INIS)
This report documents MUXS, a computer code to generate multigroup cross sections for charged particle transport problems. Cross sections generated by MUXS can be used in many multigroup transport codes, with minor modifications to these codes, to calculate sputtering yields, reflection coefficients, penetration distances, etc
MUXS: a code to generate multigroup cross sections for sputtering calculations
Energy Technology Data Exchange (ETDEWEB)
Hoffman, T.J.; Robinson, M.T.; Dodds, H.L. Jr.
1982-10-01
This report documents MUXS, a computer code to generate multigroup cross sections for charged particle transport problems. Cross sections generated by MUXS can be used in many multigroup transport codes, with minor modifications to these codes, to calculate sputtering yields, reflection coefficients, penetration distances, etc.
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)
SIMULATE-3K linkage with reactor systems codes
International Nuclear Information System (INIS)
SIMULATE-3K is Studsvik Scandpower's best-estimate three-dimensional core kinetics code. SIMULATE-3K has been coupled to several best-estimate reactor systems codes including, RELAP5-3D, RELAP5-3.3, TRACE V5.0, and RETRAN-3D. The coupled codes can be applied to existing reactors and to advanced reactor designs. The S3K linkage to each of the systems codes is a direct, explicit coupling of the two codes on a synchronous time-step basis. The coupling provides an execution method for the S3K three-dimensional neutronic model using the Nuclear Steam Supply System (NSSS) boundary conditions calculated by the systems code. Also, it allows the S3K calculated total core power and core power distributions to drive the system model core. Detailed calculations from the component codes result in a methodology for analyzing limiting transients such as steam line breaks, rod drops/ejections, and ATWS scenarios. These transient events require detailed three- dimensional core data and information about the behavior of NSSS components. A coupled analysis of these transients is important because the core behavior is closely tied to the NSSS system. For example, to capture the timing and characteristics of the important thermal-hydraulic phenomena and/or operations events, such as valve closures, safety injection, or control system interactions, requires a detailed plant model. The Peach Bottom 2 turbine trip transient is used to assess the accuracy of the coupled code calculations. Comparisons of the important plant parameters to results from RELAP5-3D, RELAP5-3.3, and TRACE V5.0 calculations are shown and discussed. The MSLB benchmark is also used to demonstrate the capabilities of the coupled code systems. Comparisons of the calculated reactor power to the reference data are shown can discussed. The comparisons demonstrate the applicability of S3K, either standalone or coupled with a system analysis code, to properly model system response during accident scenarios. (author)
Energy Technology Data Exchange (ETDEWEB)
Tobita, Masahiro; Matsui, Yoshinori [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment
2003-03-01
Prediction of irradiation temperature is one of the important issues in the design of the capsule for irradiation test. Many kinds of capsules with complex structure have been designed for recent irradiation requests, and three-dimensional (3D) temperature calculation becomes inevitable for the evaluation of irradiation temperature. For such 3D calculation, however, many works are usually needed for input data preparation, and a lot of time and resources are necessary for parametric studies in the design. To improve such situation, JAERI introduced 3D-FEM (finite element method) code NISA (Numerically Integrated elements for System Analysis) and developed several subprograms, which enabled to support input preparation works in the capsule design. The 3D temperature calculation of the capsule are able to carried out in much easier way by the help of the subprograms, and specific features in the irradiation tests such as non-uniform gamma heating in the capsule, becomes to be considered. (author)
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.
The JAERI code system for evaluation of BWR ECCS performance
International Nuclear Information System (INIS)
Development of respective computer code system of BWR and PWR for evaluation of ECCS has been conducted since 1973 considering the differences of the reactor cooling system, core structure and ECCS. The first version of the BWR code system, of which developmental work started earlier than that of the PWR, has been completed. The BWR code system is designed to provide computational tools to analyze all phases of LOCAs and to evaluate the performance of the ECCS including an ''Evaluation Model (EM)'' feature in compliance with the requirements of the current Japanese Evaluation Guideline of ECCS. The BWR code system could be used for licensing purpose, i.e. for ECCS performance evaluation or audit calculations to cross-examine the methods and results of applicants or vendors. The BWR code system presented in this report comprises several computer codes, each of which analyzes a particular phase of a LOCA or a system blowdown depending on a range of LOCAs, i.e. large and small breaks in a variety of locations in the reactor system. The system includes ALARM-B1, HYDY-B1 and THYDE-B1 for analysis of the system blowdown for various break sizes, THYDE-B-REFLOOD for analysis of the reflood phase and SCORCH-B2 for the calculation of the fuel assembl hot plane temperature. When the multiple codes are used to analyze a broad range of LOCA as stated above, it is very important to evaluate the adequacy and consistency between the codes used to cover an entire break spectrum. The system consistency together with the system performance are discussed for a large commercial BWR. (author)
Calculating system reliability with SRFYDO
Energy Technology Data Exchange (ETDEWEB)
Morzinski, Jerome [Los Alamos National Laboratory; Anderson - Cook, Christine M [Los Alamos National Laboratory; Klamann, Richard M [Los Alamos National Laboratory
2010-01-01
SRFYDO is a process for estimating reliability of complex systems. Using information from all applicable sources, including full-system (flight) data, component test data, and expert (engineering) judgment, SRFYDO produces reliability estimates and predictions. It is appropriate for series systems with possibly several versions of the system which share some common components. It models reliability as a function of age and up to 2 other lifecycle (usage) covariates. Initial output from its Exploratory Data Analysis mode consists of plots and numerical summaries so that the user can check data entry and model assumptions, and help determine a final form for the system model. The System Reliability mode runs a complete reliability calculation using Bayesian methodology. This mode produces results that estimate reliability at the component, sub-system, and system level. The results include estimates of uncertainty, and can predict reliability at some not-too-distant time in the future. This paper presents an overview of the underlying statistical model for the analysis, discusses model assumptions, and demonstrates usage of SRFYDO.
Stepšys, A.; Mickevicius, S.; Germanas, D.; Kalinauskas, R. K.
2014-11-01
This new version of the HOTB program for calculation of the three and four particle harmonic oscillator transformation brackets provides some enhancements and corrections to the earlier version (Germanas et al., 2010) [1]. In particular, new version allows calculations of harmonic oscillator transformation brackets be performed in parallel using MPI parallel communication standard. Moreover, higher precision of intermediate calculations using GNU Quadruple Precision and arbitrary precision library FMLib [2] is done. A package of Fortran code is presented. Calculation time of large matrices can be significantly reduced using effective parallel code. Use of Higher Precision methods in intermediate calculations increases the stability of algorithms and extends the validity of used algorithms for larger input values. Catalogue identifier: AEFQ_v4_0 Program summary URL: http://cpc.cs.qub.ac.uk/summaries/AEFQ_v4_0.html Program obtainable from: CPC Program Library, Queen’s University of Belfast, N. Ireland Licensing provisions: GNU General Public License, version 3 Number of lines in programs, including test data, etc.: 1711 Number of bytes in distributed programs, including test data, etc.: 11667 Distribution format: tar.gz Program language used: FORTRAN 90 with MPI extensions for parallelism Computer: Any computer with FORTRAN 90 compiler Operating system: Windows, Linux, FreeBSD, True64 Unix Has the code been vectorized of parallelized?: Yes, parallelism using MPI extensions. Number of CPUs used: up to 999 RAM(per CPU core): Depending on allocated binomial and trinomial matrices and use of precision; at least 500 MB Catalogue identifier of previous version: AEFQ_v1_0 Journal reference of previous version: Comput. Phys. Comm. 181, Issue 2, (2010) 420-425 Does the new version supersede the previous version? Yes Nature of problem: Calculation of matrices of three-particle harmonic oscillator brackets (3HOB) and four-particle harmonic oscillator brackets (4HOB) in a more
Development of the next generation reactor analysis code system, MARBLE
International Nuclear Information System (INIS)
A next generation reactor analysis code system, MARBLE, has been developed. MARBLE is a successor of the fast reactor neutronics analysis code systems, JOINT-FR and SAGEP-FR (conventional systems), which were developed for so-called JUPITER standard analysis methods. MARBLE has the equivalent analysis capability to the conventional system because MARBLE can utilize sub-codes included in the conventional system without any change. On the other hand, burnup analysis functionality for power reactors is improved compared with the conventional system by introducing models on fuel exchange treatment and control rod operation and so on. In addition, MARBLE has newly developed solvers and some new features of burnup calculation by the Krylov sub-space method and nuclear design accuracy evaluation by the extended bias factor method. In the development of MARBLE, the object oriented technology was adopted from the view-point of improvement of the software quality such as flexibility, expansibility, facilitation of the verification by the modularization and assistance of co-development. And, software structure called the two-layer system consisting of scripting language and system development language was applied. As a result, MARBLE is not an independent analysis code system which simply receives input and returns output, but an assembly of components for building an analysis code system (i.e. framework). Furthermore, MARBLE provides some pre-built analysis code systems such as the fast reactor neutronics analysis code system. SCHEME, which corresponds to the conventional code and the fast reactor burnup analysis code system, ORPHEUS. (author)
SAMDIST: A computer code for calculating statistical distributions for R-matrix resonance parameters
Energy Technology Data Exchange (ETDEWEB)
Leal, L.C.; Larson, N.M.
1995-09-01
The SAMDIST computer code has been developed to calculate distribution of resonance parameters of the Reich-Moore R-matrix type. The program assumes the parameters are in the format compatible with that of the multilevel R-matrix code SAMMY. SAMDIST calculates the energy-level spacing distribution, the resonance width distribution, and the long-range correlation of the energy levels. Results of these calculations are presented in both graphic and tabular forms.
A Comparative Study of the Code Calculations for Local Flow Blockages in the KALIMER-150 Core
Energy Technology Data Exchange (ETDEWEB)
Chang, Won Pyo; Ha, Ki Suk; Lee, Yong Bum
2009-09-15
A sub-channel blockage may be caused by ingression of damaged fuel debris or foreign obstacles into a core fuel subassembly for a liquid metal reactor(LMR) due to its geometrical compactness of the core design. Local coolant temperature could rise during the incident and it might eventually lead to the degradation of the fuel rods. An analysis computer code is obviously needed not only to assure the safe design of the core, but also to design an effective monitoring system to prevent it from propagating to a serious consequence. The code, therefore, must be capable of representing the thermal-hydraulic phenomena anticipated during the incident reasonably enough to be used to evaluate fuel rod intactness. Most of the technically leading countries for LMR have developed and are using the codes for sub-channel blockage analyses, Korea couldn't afford the resources sponsoring the develop of such a code past years. It was realized later that such an analysis code would ultimately be a prerequisite in the future licensing process of KALIMER as its conceptual design was being elaborated. Since those advanced countries had been reluctant to transfer such the codes, MATRA-LMR/FB had to be developed independently in Korea. It is a revised version of the existing MATRA-LMR code which was aimed for the core sub-channel analysis of LMRs. Some of its models have been improved so appropriately to be able to analyze the sub-channel blockages. Nevertheless, an experiment relevant to the sub-channel blockages had never been conducted for investigating the phenomenon in Korea. Further more, very few experimental data are available on published papers or reports world wide. Under this circumstance, a study has been made as an effort to evaluate the prediction capability of the MATRA-LMR/FB code by comparing the calculation results of the SABRE code, which had already been applied to EFR design. In result, a discrepancy has been observed in a case, but an overall agreement has
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
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...... on the basis of kinetics. However, many of the most interesting non-natural chemical reactions which could potentially be catalysed by enzymes, are thermodynamically unfavourable and are thus limited by the equilibrium position of the reaction. A good example is the enzyme ω-transaminase, which catalyses...... be altered by coupling with other reactions. For instance, in the case of ω-transaminase, such a coupling could be with alanine dehydrogenase. Herein, the aim of this work is to identify thermodynamic bottlenecks within a multi-enzyme process, using group contribution method to calculate the Gibbs free...
Radiative Transfer Code: Application to the calculation of PAR
Indian Academy of Sciences (India)
D Emmanuel; D Phillippe; C Malik
2000-12-01
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 column. We focus on the visible spectrum of the solar radiation defined as the Photosynthetically Active Radiation (PAR). We developed a model (Chami et al. 1997) to simulate the behavior of the solar beam in the atmosphere and the ocean. We first describe the theoretical basis of the code and the method we used to solve the radiative transfer equation (RTE): the successive orders of scattering (SO). The second part deals with a sensitivity study of the PAR just above and below the sea surface for various atmospheric conditions. In a cloudy sky, we computed a ratio between vector fluxes just above the sea surface and spherical fluxes just beneath the sea surface. When the optical thickness of the cloud increases this ratio remains constant and around 1.29. This parameter is convenient to convert vector flux at the sea surface as retrieved from satellite to PAR. Subsequently, we show how solar radiation as vector flux rather than PAR leads to an underestimate of the primary production up to 40% for extreme cases.
International Nuclear Information System (INIS)
The authors report calculations performed using the MNCP and PENELOPE codes to determine the Hp(3)/K air conversion coefficient which allows the Hp(3) dose equivalent to be determined from the measured value of the kerma in the air. They report the definition of the phantom, a 20 cm diameter and 20 cm high cylinder which is considered as representative of a head. Calculations are performed for an energy range corresponding to interventional radiology or cardiology (20 keV-110 keV). Results obtained with both codes are compared
Quasiparticle GW calculations within the GPAW electronic structure code
DEFF Research Database (Denmark)
Hüser, Falco
properties are to a large extent governed by the physics on the atomic scale, that means pure quantum mechanics. For many decades, Density Functional Theory has been the computational method of choice, since it provides a fairly easy and yet accurate way of determining electronic structures and related...... is considered, which can be regarded as the lowest level of the GW approximation. This thesis documents the implementation of the G0W0 approximation in GPAW. It serves two purposes: First, it can be read as a manual by anyone who is interested in doing GW calculations with GPAW. All features and requirements...
Development of leak rate calculation model and code in piping
International Nuclear Information System (INIS)
Background: With the development of fracture mechanics, Leak-Before-Break (LBB) is widely used in nuclear power plant piping design. Purpose: In order to support the application of LBB, leak rate through crack need to be calculated. Methods: In this text, an analytical flow model is developed based on homogeneous non-equilibrium model, a computer program is also developed based on this analytical model. Results: Comparison between the results from the above program and test results shows that deviations of calc. results to test results are within ±50%. Conclusions: Conclusions can be got that this analytical model and computer program meet the requirement of engineering application. (authors)
Preliminary Development of Thermal Power Calculation Code H-Power for a Supercritical Water Reactor
Directory of Open Access Journals (Sweden)
Fan Zhang
2014-01-01
Full Text Available SCWR (Supercritical Water Reactor is one of the promising Generation IV nuclear systems, which has higher thermal power efficiency than current pressurized water reactor. It is necessary to perform the thermal equilibrium and thermal power calculation for the conceptual design and further monitoring and calibration of the SCWR. One visual software named H-Power was developed to calculate thermal power and its uncertainty of SCWR, in which the advanced IAPWS-IF97 industrial formulation was used to calculate the thermodynamic properties of water and steam. The ISO-5167-4: 2003 standard was incorporated in the code as the basis of orifice plate to compute the flow rate. New heat balance model and uncertainty estimate have also been included in the code. In order to validate H-Power, an assessment was carried out by using data published by US and Qinshan Phase II. The results showed that H-Power was able to estimate the thermal power of SCWR.
Energy Technology Data Exchange (ETDEWEB)
Chersola, Davide [GeNERG – DIME/TEC, University of Genova, via all’Opera Pia 15/a, 16145 Genova (Italy); INFN, via Dodecaneso 33, 16146 Genova (Italy); Lomonaco, Guglielmo, E-mail: guglielmo.lomonaco@unige.it [GeNERG – DIME/TEC, University of Genova, via all’Opera Pia 15/a, 16145 Genova (Italy); INFN, via Dodecaneso 33, 16146 Genova (Italy); Marotta, Riccardo [GeNERG – DIME/TEC, University of Genova, via all’Opera Pia 15/a, 16145 Genova (Italy); INFN, via Dodecaneso 33, 16146 Genova (Italy); Mazzini, Guido [Centrum výzkumu Řež (Research Centre Rez), Husinec-Rez, cp. 130, 25068 Rez (Czech Republic)
2014-07-01
Highlights: • MC codes are widely adopted to analyze nuclear facilities, including GEN-IV reactors. • Burnup calculations are an efficient tool to test neutronic Monte Carlo codes. • In this comparison the used codes show some differences but a good agreement exists. - Abstract: This paper presents the comparison between two Monte Carlo based burnup codes: SERPENT and MONTEBURNS. Monte Carlo codes are fully and worldwide adopted to perform analyses on nuclear facilities, also in the frame of Generation IV advanced reactors simulations. Thus, faster and most powerful calculation codes are needed with the aim to analyze complex geometries and specific neutronic behaviors. Burnup calculations are an efficient tool to test neutronic Monte Carlo codes: indeed these calculations couple transport and depletion procedures, so that neutronic reactor behavior can be simulated in its totality. Comparisons have been performed on a configuration representing the Allegro MOX 75 MW{sub th} reactor proposed by the European GoFastR (Gas-cooled Fast Reactor) Project in the frame of the 7th Euratom Framework Program. Although in burnup and criticality comparisons the codes used in simulations show different calculation times and some differences in amounts and in precision (in term of statistical errors), a reasonably good agreement between them exists.
Code of hybrid calculation for the study of heat exchangers
International Nuclear Information System (INIS)
A series integration method has been chosen, machine time being similar to space, the computer storing the solutions passed to allow for an approximation at the finished differences of the timed derivatives. Space distributions of the functions of state (enthalpy and temperature) of the two fluids are calculated by analogical integration in the direction of the circulation of the fluids. This is made possible by the disconnection of the two fluids thanks to a method of prediction-correction for the equation of the wall. The great rapidity of the analogical integrations and the absence of iterations make the calculation in actual time possible (assuming that 200 points are taken for the storage of the solutions passed, the time required for a full calculation is 0.4 second). It is, therefore, possible to integrate this simulation in an actual (for example Nuclear Station) or simulated loop. The exchanger may or may not be with counter-current, with or without change of phase. Thermodynamic tables with two variables are introduced into the computer, interpolation is effected in analogical and this allows, inter alia, enthalpy, temperature, pressure and volumic mass to be perfectly in phase one with the other. The work bears most particularly on the accuracy and stability of simulation of the timed derivatives which are obtained by difference between an analogical function and the same function stored in the computer at the preceding step. Particular emphasis is made to bear on the operation of the model, a monitor program controlling the various phases of utilization in function of the type of exchanger, experiments to be made; a program of control of the results enables the operator to obtain the results either immediately in printed form, curve tracer or oscilloscope, put to scale, or in deferred a more elaborate processing of the results. The same program allows for the simulation of any exchanger (different fluid, variable geometry, etc.) or any number of
FOOD: an interactive code to calculate internal radiation doses from contaminated food products
International Nuclear Information System (INIS)
An interactive code, FOOD, has been written in BASIC for the UNIVAC 1108 to facilitate calculation of internal radiation doses to man from radionuclides in food products. In the dose model, vegetation may be contaminated by either air or irrigation water containing radionuclides. The model considers two mechanisms for radionuclide contamination of vegetation: direct deposition on leaves and uptake from soil through the root system. The user may select up to 14 food categories with corresponding consumption rates, growing periods and either irrigation rates or atmospheric deposition rates. These foods include various kinds of produce, grains and animal products. At present, doses may be calculated for the skin, total body and five internal organs from 190 radionuclides. Dose summaries can be displayed at the local terminal. Further details on percent contribution to dose by nuclide and by food type are available from an auxiliary high-speed printer. This output also includes estimated radionuclide concentrations in soil, plants and animal products
International Nuclear Information System (INIS)
In the frame of a project on scientific-technical co-operation funded by BMBF/BMWi, the program code DYN3D and the coupled code ATHLET-DYN3D have been transferred to the Institute for Nuclear Research and Nuclear Energy (INRNE) Sofia. The coupled code represents an implementation of the 3D core model DYN3D developed by FZR into the GRS thermal-hydraulics code system ATHLET. For the purpose of validation of these codes, a measurement data base about a start-up experiment obtained at the unit 6 of Kozloduy NPP (VVER-1000/V-320) has been generated. The results of performed validation calculations were compared with measurement values from the data base. A simplified model for estimation of cross flow mixing between fuel assemblies has been implemented into the program code DYN3D by Bulgarian experts. Using this cross flow model, transient processes with asymmetrical boundary conditions can be analysed more realistic. The validation of the implemented model were performed with help of comparison calculations between modified DYD3D code and thermal-hydraulics code COBRA-4I, and also on the base of the collected measurement data from Kozloduy NPP. (orig.)
Chromaticity calculations and code comparisons for x-ray lithography source XLS and SXLS rings
International Nuclear Information System (INIS)
This note presents the chromaticity calculations and code comparison results for the (x-ray lithography source) XLS (Chasman Green, XUV Cosy lattice) and (2 magnet 4T) SXLS lattices, with the standard beam optic codes, including programs SYNCH88.5, MAD6, PATRICIA88.4, PATPET88.2, DIMAD, BETA, and MARYLIE. This analysis is a part of our ongoing accelerator physics code studies. 4 figs., 10 tabs
Chromaticity calculations and code comparisons for x-ray lithography source XLS and SXLS rings
Energy Technology Data Exchange (ETDEWEB)
Parsa, Z.
1988-06-16
This note presents the chromaticity calculations and code comparison results for the (x-ray lithography source) XLS (Chasman Green, XUV Cosy lattice) and (2 magnet 4T) SXLS lattices, with the standard beam optic codes, including programs SYNCH88.5, MAD6, PATRICIA88.4, PATPET88.2, DIMAD, BETA, and MARYLIE. This analysis is a part of our ongoing accelerator physics code studies. 4 figs., 10 tabs.
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.
SFR whole core burnup calculations with TRIPOLI-4 Monte Carlo code
International Nuclear Information System (INIS)
Under the Working Party on Scientific Issues of Reactor Systems (WPRS) of the OECD/NEA, an international collaboration benchmark was recently established on the neutronic analysis of four Sodium-cooled Fast Reactor (SFR) concepts of the Generation- IV nuclear energy systems. As the whole core Monte Carlo depletion calculation is one of the essential challenges of current reactor physics studies, the continuous-energy TRIPOLI-4 Monte Carlo transport code was firstly used in this study to perform whole core 3D neutronic calculations for these four SFR cores. Two medium size (1000 MWt) and two large size (3600 MWt) SFR of GEN-IV systems were analyzed. The medium size SFR concepts are from the Advanced Burner Reactors (ABR). The large size SFR concepts are from the self-breeding reactors. The TRIPOLI-4 depletion calculations were made with MOX and metallic U-Pu-Zr fuels for the ABR cores and with MOX and Carbide (U,Pu)C fuels for the self-breeding cores. The whole core reactor physics parameters calculations were performed for the BOEC and EOEC (Beginning and End of Equilibrium Cycle) conditions. This paper summarizes the TRIPOLI-4 calculation results of Keff, βeff, sodium void worth, Doppler constant, control rod worth, and core power distributions for the BOEC and EOEC conditions. The one-cycle depletion calculation results of the core inventory of U and TRU (Pu, Am, Cm, and Np) are also analyzed, after 328.5 days depletion irradiation for the ABR cores, 410 days for the large MOX core, and 500 days for the large carbide core. (author)
Core design calculations of IRIS reactor using modified CORD-2 code package
International Nuclear Information System (INIS)
Core design calculations, with thermal-hydraulic feedback, for the first cycle of the IRIS reactor were performed using the modified CORD-2 code package. WIMSD-5B code is applied for cell and cluster calculations with two different 69-group data libraries (ENDF/BVI rev. 5 and JEF-2.2), while the nodal code GNOMER is used for diffusion calculations. The objective of the calculation was to address basic core design problems for innovative reactors with long fuel cycle. The results were compared to our results obtained with CORD-2 before the modification and to preliminary results obtained with CASMO code for a similar problem without thermal-hydraulic feedback.(author)
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.
Application of Monte Carlo code EGS4 to calculate gamma exposure buildup factors
International Nuclear Information System (INIS)
Exposure buildup factors up to 40 mean free paths ranging from 0.015 MeV to 15 MeV photon energy were calculated by using the Monte Carlo simulation code EGS4 for ordinary concrete. The calculation involves PHOTX cross section library, a point isotropic source, infinite uniform medium model and a particle splitting method and considers the Bremsstrahlung, fluorescent effect, correlative (Rayleigh) scatter. The results were compared with the relevant data. Results show that the data of the buildup factors calculated by the Monte Carlo code EGS4 was reliable. The Monte Carlo method can be used widely to calculate gamma-ray exposure buildup factors. (authors)
International Nuclear Information System (INIS)
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
Code Formal Verification of Operation System
Directory of Open Access Journals (Sweden)
Yu Zhang
2010-12-01
Full Text Available with the increasing pressure on non-function attributes (security, safety and reliability requirements of an operation system, high–confidence operation system is becoming more important. Formal verification is the only known way to guarantee that a system is free of programming errors. We research on formal verification of operation system kernel in system code level and take theorem proving and model checking as the main technical methods to resolve the key techniques of verifying operation system kernel in C code level. We present a case study to the verification of real-world C systems code derived from an implementation of μC/OS – II in the end.
A general dimensional neutron diffusion calculation code: ADC
International Nuclear Information System (INIS)
A FORTRAN computer program ADC is developed for the FACOM 230-75 computer to be capable of solving eigenvalue problems of neutron diffusion equation in one, two and three spatial dimensions. The available coordinate systems are orthogonal (X), (X,Y), (X,Y,Z) and cylindrical (R,Z), (R,THETA), (R,THETA,Z). The outer boundary condition for the neutron flux can be chosen to be symmetric, zero flux or log-derivative condition. The present program can be used also for obtaining the adjoint flux. (author)
Design of a physical format coding system
Hu, Beibei; Pei, Jing; Zhang, Qicheng; Liu, Hailong; Tang, Yi
2008-12-01
A novel design of physical format coding system (PFCS) is presented based on Multi-level read-only memory disc (ML ROM) in order to solve the problem of low efficiency and long period of disc testing during system development. The PFCS is composed of four units, which are 'Encode', 'Add Noise', 'Decode', 'Error Rate', and 'Information'. It is developed with MFC under the environment of VC++ 6.0, and capable to visually simulate the procedure of data processing for ML ROM. This system can also be used for developing other optical disc storage system or similar channel coding system.
Calculation of the CAREM reactor with the HUEMUL-PUMA-THERMIT chain of codes
International Nuclear Information System (INIS)
The purpose of the work was the evaluation of the the CAREM 25 reactor core, using a chain of codes (HUEMUL-PUMA-THERMIT) different to the one used in the original design (CONDOR-CITVAP-THERMIT). First, we performed a partial validation of the our codes in lattices similar to CAREM and reproduced a benchmark for simulation of gadolinium burnup. The results were considered satisfactory for this stage of the project. Then, we calculated the core along the normal operating equilibrium cycle and in hot and cold shut-down conditions. The main outcome of our evaluation confirms the general behaviour of the reference calculations except in one important point referring to the cold shut down. In this condition, the failure of one single rod of bank number 13 of the shut down system, leaves the core in a supercritical state at the beginning of the cycle and this anomaly persists during almost a third of the overall cycle. A new design of the core is proposed with minor modifications of the reference one, without introducing new types of fuel elements and keeping the same fuel management scheme. This new core fulfills all the design requirements. (author)
Energy Technology Data Exchange (ETDEWEB)
Young, P.G. [Los Alamos National Lab., NM (United States); Chadwick, M.B. [Lawrence Livermore National Lab., CA (United States)
1994-06-01
A number of modifications have been made to the reaction theory code GNASH in order the accuracy of calculations at incident particle energies up to 200 MeV. Direct reaction a level density models appropriate for higher energy calculations are now used in the code, and most importantly, improved preequilibrium models have been incorporated into the code system. The code has been used to calculate proton-induced reactions on {sup 90}Zr and {sup 208}Pb for the International Code and Model Intercomparison for Intermediate Energy Reactions organized by the NEA. Calculations were performed with GNASH at incident proton energies of 25, 45, 80, and 160 mev using both the exciton model and Feshbach-Kerman-Koonin theory for the preequilibrium component. The models and procedures used in the GNASH calculations with the exciton model are described here. The results are compared to experimental data and to results from the GNASH calculations with Feshbach-Kerman-Koonin preequilibrium theory.
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.
Calculation of equilibria at elevated temperatures using the MINTEQ geochemical code
Energy Technology Data Exchange (ETDEWEB)
Smith, R.W.
1988-12-01
Coefficients and equations for calculating mineral hydrolysis constants, solubility products and formation constants for 60 minerals and 57 aqueous species in the 13 component thermodynamic system K/sub 2/O-Na/sub 2/O-CaO-MgO-FeO-Al/sub 2/O/sub 3/-SiO/sub 2/-CO/sub 2/-H/sub 2/O-HF-HCl-H/sub 2/S-H/sub 2/SO/sub 4/ are presented in a format suitable for inclusion in the MINTEQ computer code. The temperature functions presented for minerals are based on the MINTEQ data base at 25/degree/C and the integration of analytical heat capacity power functions. This approach ensures that the temperature functions join smoothly with the low-temperature data base. A new subroutine, DEBYE, was added to MINTEQ that is used to calculate the theoretical Debye-Hueckel parameters A and B as a function of temperature. In addition, this subroutine also calculates a universal value of the extended Debye-Hueckel parameter, b/sub i/, as a function of temperature. The coefficients and equations provide the capability to use MINTEQ to more accurately calculate water/rock equilibrium for temperatures of up to 250/degree/C, and in dilute, low-sulfate, near neutral groundwaters to 300/degree/C. 52 refs., 1 fig., 6 tabs.
Boltzmann-Fokker-Planck calculations using standard discrete-ordinates codes
International Nuclear Information System (INIS)
The Boltzmann-Fokker-Planck (BFP) equation can be used to describe both neutral and charged-particle transport. Over the past several years, the author and several collaborators have developed methods for representing Fokker-Planck operators with standard multigroup-Legendre cross-section data. When these data are input to a standard S/sub n/ code such as ONETRAN, the code actually solves the Boltzmann-Fokker-Planck equation rather than the Boltzmann equation. This is achieved wihout any modification to the S/sub n/ codes. Because BFP calculations can be more demanding from a numerical viewpoint than standard neutronics calculations, we have found it useful to implement new quadrature methods ad convergence acceleration methods in the standard discrete-ordinates code, ONETRAN. We discuss our BFP cross-section representation techniques, our improved quadrature and acceleration techniques, and present results from BFP coupled electron-photon transport calculations performed with ONETRAN. 19 refs., 7 figs
Code Formal Verification of Operation System
Yu Zhang; Yunwei Dong; Huo Hong; Fan Zhang
2010-01-01
with the increasing pressure on non-function attributes (security, safety and reliability) requirements of an operation system, high–confidence operation system is becoming more important. Formal verification is the only known way to guarantee that a system is free of programming errors. We research on formal verification of operation system kernel in system code level and take theorem proving and model checking as the main technical methods to resolve the key techniques of verifying operatio...
Manual of Nucost 1.0 - code for calculation of nuclear power generation costs
International Nuclear Information System (INIS)
Nucost is a computer code developed at CDTN to perform cost calculation of electric power generated in PWR nuclear power plants, based on present worth cost method. The Nucost version 1.0 performs calculations of nuclear fuel cost cycle by cycle during the time life of the power plant. That calculation is performed with enough details permitting optimization and minimization. The code is also a tool to aid reload projects and economic operation of PWR reactors. This manual presents a description of Nucost version 1.0, instruction to enter data preparation and description of the Nucost output. (M.I.)
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.
Principles of the reactor code system RHEIN
International Nuclear Information System (INIS)
A description is given of the principles of the reactor code system RHEIN which is applied in connection with a BESM6-type computer. In transfering data between the components of the system external storage is used. The programme passage is controlled by the input data. (author)
Modular ORIGEN-S for multi-physics code systems
Energy Technology Data Exchange (ETDEWEB)
Yesilyurt, Gokhan; Clarno, Kevin T.; Gauld, Ian C., E-mail: yesilyurtg@ornl.gov, E-mail: clarnokt@ornl.gov, E-mail: gauldi@ornl.gov [Oak Ridge National Laboratory, TN (United States); Galloway, Jack, E-mail: jack@galloways.net [Los Alamos National Laboratory, Los Alamos, NM (United States)
2011-07-01
The ORIGEN-S code in the SCALE 6.0 nuclear analysis code suite is a well-validated tool to calculate the time-dependent concentrations of nuclides due to isotopic depletion, decay, and transmutation for many systems in a wide range of time scales. Application areas include nuclear reactor and spent fuel storage analyses, burnup credit evaluations, decay heat calculations, and environmental assessments. Although simple to use within the SCALE 6.0 code system, especially with the ORIGEN-ARP graphical user interface, it is generally complex to use as a component within an externally developed code suite because of its tight coupling within the infrastructure of the larger SCALE 6.0 system. The ORIGEN2 code, which has been widely integrated within other simulation suites, is no longer maintained by Oak Ridge National Laboratory (ORNL), has obsolete data, and has a relatively small validation database. Therefore, a modular version of the SCALE/ORIGEN-S code was developed to simplify its integration with other software packages to allow multi-physics nuclear code systems to easily incorporate the well-validated isotopic depletion, decay, and transmutation capability to perform realistic nuclear reactor and fuel simulations. SCALE/ORIGEN-S was extensively restructured to develop a modular version that allows direct access to the matrix solvers embedded in the code. Problem initialization and the solver were segregated to provide a simple application program interface and fewer input/output operations for the multi-physics nuclear code systems. Furthermore, new interfaces were implemented to access and modify the ORIGEN-S input variables and nuclear cross-section data through external drivers. Three example drivers were implemented, in the C, C++, and Fortran 90 programming languages, to demonstrate the modular use of the new capability. This modular version of SCALE/ORIGEN-S has been embedded within several multi-physics software development projects at ORNL, including
Calculation Of Extraction Optics For Ion System With Plazma Emitter
Frolov, B A
2004-01-01
The 2-D code for simulating of ion optics system of positive ion extraction from a plasma source is described. Example calculation of 100 kV optics for the extraction ion IHEP gun is presented. The trajectories of particles and emittance plots are resulted. The aberrations influ-ence strongly on ion optics for considered geometry.
Methods, algorithms and computer codes for calculation of electron-impact excitation parameters
Bogdanovich, P; Stonys, D
2015-01-01
We describe the computer codes, developed at Vilnius University, for the calculation of electron-impact excitation cross sections, collision strengths, and excitation rates in the plane-wave Born approximation. These codes utilize the multireference atomic wavefunctions which are also adopted to calculate radiative transition parameters of complex many-electron ions. This leads to consistent data sets suitable in plasma modelling codes. Two versions of electron scattering codes are considered in the present work, both of them employing configuration interaction method for inclusion of correlation effects and Breit-Pauli approximation to account for relativistic effects. These versions differ only by one-electron radial orbitals, where the first one employs the non-relativistic numerical radial orbitals, while another version uses the quasirelativistic radial orbitals. The accuracy of produced results is assessed by comparing radiative transition and electron-impact excitation data for neutral hydrogen, helium...
Energy Technology Data Exchange (ETDEWEB)
Vitruk, S.G.; Korsun, A.S. [Moscow Engineering Physics Institute (Russian Federation); Ushakov, P.A. [Institute of Physics and Power Engineering, Obninsk (R)] [and others
1995-09-01
The multilevel mathematical model of neutron thermal hydrodynamic processes in a passive safety core without assemblies duct walls and appropriate computer code SKETCH, consisted of thermal hydrodynamic module THEHYCO-3DT and neutron one, are described. A new effective discretization technique for energy, momentum and mass conservation equations is applied in hexagonal - z geometry. The model adequacy and applicability are presented. The results of the calculations show that the model and the computer code could be used in conceptual design of advanced reactors.
A FORTRAN computer code for calculating flows in multiple-blade-element cascades
Mcfarland, E. R.
1985-01-01
A solution technique has been developed for solving the multiple-blade-element, surface-of-revolution, blade-to-blade flow problem in turbomachinery. The calculation solves approximate flow equations which include the effects of compressibility, radius change, blade-row rotation, and variable stream sheet thickness. An integral equation solution (i.e., panel method) is used to solve the equations. A description of the computer code and computer code input is given in this report.
A computer code for calculations in the algebraic collective model of the atomic nucleus
Welsh, T A
2016-01-01
A Maple code is presented for algebraic collective model (ACM) calculations. The ACM is an algebraic version of the Bohr model of the atomic nucleus, in which all required matrix elements are derived by exploiting the model's SU(1,1) x SO(5) dynamical group. This, in particular, obviates the use of coefficients of fractional parentage. This paper reviews the mathematical formulation of the ACM, and serves as a manual for the code. The code makes use of expressions for matrix elements derived elsewhere and newly derived matrix elements of the operators [pi x q x pi]_0 and [pi x pi]_{LM}, where q_M are the model's quadrupole moments, and pi_N are corresponding conjugate momenta (-2>=M,N<=2). The code also provides ready access to SO(3)-reduced SO(5) Clebsch-Gordan coefficients through data files provided with the code.
International Nuclear Information System (INIS)
The CITHAN computer code was developed at IPEN (Instituto de Pesquisas Energeticas e Nucleares) to link the HAMMER computer code with a fuel depletion routine and to provide neutron cross sections to be read with the appropriate format of the CITATION code. The problem arised due to the efforts to addapt the new version denomined HAMMER-TECHION with the routine refered. The HAMMER-TECHION computer code was elaborated by Haifa Institute, Israel within a project with EPRI. This version is at CNEN to be used in multigroup constant generation for neutron diffusion calculation in the scope of the new methodology to be adopted by CNEN. The theoretical formulation of CITHAM computer code, tests and modificatins are described. (Author)
ANIGAM: a computer code for the automatic calculation of nuclear group data
International Nuclear Information System (INIS)
The computer code ANIGAM consists mainly of the well-known programmes GAM-I and ANISN as well as of a subroutine which reads the THERMOS cross section library and prepares it for ANISN. ANIGAM has been written for the automatic calculation of microscopic and macroscopic cross sections of light water reactor fuel assemblies. In a single computer run both were calculated, the cross sections representative for fuel assemblies in reactor core calculations and the cross sections of each cell type of a fuel assembly. The calculated data were delivered to EXTERMINATOR and CITATION for following diffusion or burn up calculations by an auxiliary programme. This report contains a detailed description of the computer codes and methods used in ANIGAM, a description of the subroutines, of the OVERLAY structure and an input and output description. (oririg.)
An Efficient Group Key Management Using Code for Key Calculation for Simultaneous Join/Leave: CKCS
Directory of Open Access Journals (Sweden)
Melisa Hajyvahabzadeh
2012-08-01
Full Text Available This paper presents an efficient group key management protocol, CKCS (Code for Key Calculation in Simultaneous join/leave for simultaneous join/leave in secure multicast. This protocol is based on logical key hierarchy. In this protocol, when new members join the group simultaneously, server sends only thegroup key for those new members. Then, current members and new members calculate the necessary keys by node codes and one-way hash function. A node code is a random number which is assigned to each key to help users calculate the necessary keys. Again, at leave, the server just sends the new group key to remaining members. The results show that CKCS reduces computational and communication overhead, and also message size in simultaneous join/leave.
Calculations of reactor-accident consequences, Version 2. CRAC2: computer code user's guide
International Nuclear Information System (INIS)
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
The FLUFF code for calculating finned surface heat transfer -description and user's guide
International Nuclear Information System (INIS)
FLUFF is a computer code for calculating heat transfer from finned surfaces by convection and radiation. It can also represent heat transfer by radiation to a partially emitting and absorbing medium within the fin cavity. The FLUFF code is useful not only for studying the behaviour of finned surfaces but also for deriving heat fluxes which can be applied as boundary conditions to other heat transfer codes. In this way models of bodies with finned surfaces may be greatly simplified since the fins need not be explicitly represented. (author)
VVER-440 Ex-Core Neutron Transport Calculations by MCNP-5 Code and Comparison with Experiment
Energy Technology Data Exchange (ETDEWEB)
Borodkin, Pavel; Khrennikov, Nikolay [Scientific and Engineering Centre for Nuclear and Radiation Safety (SEC NRS) Malaya Krasnoselskaya ul., 2/8, bld. 5, 107140 Moscow (Russian Federation)
2008-07-01
Ex-core neutron transport calculations are needed to evaluate radiation loading parameters (neutron fluence, fluence rate and spectra) on the in-vessel equipment, reactor pressure vessel (RPV) and support constructions of VVER type reactors. Due to these parameters are used for reactor equipment life-time assessment, neutron transport calculations should be carried out by precise and reliable calculation methods. In case of RPVs, especially, of first generation VVER-440s, the neutron fluence plays a key role in the prediction of RPV lifetime. Main part of VVER ex-core neutron transport calculations are performed by deterministic and Monte-Carlo methods. This paper deals with precise calculations of the Russian first generation VVER-440 by MCNP-5 code. The purpose of this work was an application of this code for expert calculations, verification of results by comparison with deterministic calculations and validation by neutron activation measured data. Deterministic discrete ordinates DORT code, widely used for RPV neutron dosimetry and many times tested by experiments, was used for comparison analyses. Ex-vessel neutron activation measurements at the VVER-440 NPP have provided space (in azimuth and height directions) and neutron energy (different activation reactions) distributions data for experimental (E) validation of calculated results. Calculational intercomparison (DORT vs. MCNP-5) and comparison with measured values (MCNP-5 and DORT vs. E) have shown agreement within 10-15% for different space points and reaction rates. The paper submits a discussion of results and makes conclusions about practice use of MCNP-5 code for ex-core neutron transport calculations in expert analysis. (authors)
Nexus: A modular workflow management system for quantum simulation codes
Krogel, Jaron T.
2016-01-01
The management of simulation workflows represents a significant task for the individual computational researcher. Automation of the required tasks involved in simulation work can decrease the overall time to solution and reduce sources of human error. A new simulation workflow management system, Nexus, is presented to address these issues. Nexus is capable of automated job management on workstations and resources at several major supercomputing centers. Its modular design allows many quantum simulation codes to be supported within the same framework. Current support includes quantum Monte Carlo calculations with QMCPACK, density functional theory calculations with Quantum Espresso or VASP, and quantum chemical calculations with GAMESS. Users can compose workflows through a transparent, text-based interface, resembling the input file of a typical simulation code. A usage example is provided to illustrate the process.
Development of BERMUDA: a radiation transport code system, 1
International Nuclear Information System (INIS)
A radiation transport code system BERMUDA has been developed for one-, two- and three-dimensional geometries. The time-independent transport equation is numerically solved using a direct integration method in a multigroup model, to obtain spatial, angular and energy distributions of neutron, gamma rays or adjoint neutron flux. As to group constants, a library with an any structure of energy groups is capable to be produced from a data base JSSTDL, or by a processing code PROF-GROUCH-G/B, selecting objective nuclear data through a retrieval system EDFSRS. Validity of the present code system has been tested by analyzing the shielding benchmark experiments. The test has shown that accurate results are obtainable with this system especially in deep penetration calculation. Described are the devised calculation method and the results of validity tests. Input data specification, job control languages and output data are also described as a user's manual for the following four neutron transport codes: BERMUDA-1DN : sphere, slab(S20), BERMUDA-2DN : cylinder (S8), BERMUDA-2DN-S16 : cylinder (S16), and BERMUDA-3DN : rectangular parallelpiped (S8). (J.P.N.)
Energy Technology Data Exchange (ETDEWEB)
Perfetti, C.; Martin, W. [Univ. of Michigan, Dept. of Nuclear Engineering and Radiological Sciences, 2355 Bonisteel Boulevard, Ann Arbor, MI 48109-2104 (United States); Rearden, B.; Williams, M. [Oak Ridge National Laboratory, Reactor and Nuclear Systems Div., Bldg. 5700, P.O. Box 2008, Oak Ridge, TN 37831-6170 (United States)
2012-07-01
Three methods for calculating continuous-energy eigenvalue sensitivity coefficients were developed and implemented into the Shift Monte Carlo code within the SCALE code package. The methods were used for two small-scale test problems and were evaluated in terms of speed, accuracy, efficiency, and memory requirements. A promising new method for calculating eigenvalue sensitivity coefficients, known as the CLUTCH method, was developed and produced accurate sensitivity coefficients with figures of merit that were several orders of magnitude larger than those from existing methods. (authors)
A Photovoltaic System Payback Calculator.
Energy Technology Data Exchange (ETDEWEB)
Riley, Daniel; Fleming, Jeffrey; Gallegos, Gerald R.
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.
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.
LOLA-SYSTEM, JEN-UPM PWR Fuel Management System Burnup Code System
International Nuclear Information System (INIS)
1 - Description of program or function: The LOLA-SYSTEM is a part of the JEN-UPM code package for PWR fuel management, scope or design calculations. It is a code package for core burnup calculations using nodal theory based on a FLARE type code. The LOLA-SYSTEM includes four modules: the first one (MELON-3) generates the constants of the K-inf and M2 correlations to be input into SIMULA-3. It needs the K-inf and M2 fuel assembly values at different conditions of moderator temperature, Boron concentration, burnup, etc., which are provided by MARIA fuel assembly calculations. The main module (SIMULA-3) is the core burnup calculation code in three dimensions and one group of energy. It normally uses a geometrical representation of one node per fuel assembly or per quarter of fuel assembly. It has included a thermal hydraulic feedback on flow and voids and criticality searches on boron concentration and control rods insertion. The CONCON code makes the calculation of the albedo, transport factors, K-inf and M2 correction factors to be input into SIMULA-3. The calculation is made in the XY transversal plane. The CONAXI code is similar to CONCON, but in the axial direction. 2 - Method of solution: MELON-3 makes a mean squares fit of K-inf and M2 values at different conditions in order to determine the constants of the feedback correlations. SIMULA-3 uses a modified one-group nodal theory, with a new transport kernel that provides the same node interface leakages as a fine mesh diffusion calculation. CONCON and CONAXI determine the transport and correction factors, as well as the albedo, to be input into SIMULA-3. They are determined by a method of leakages equivalent to the detailed diffusion calculation of CARMEN or VENTURE; these factors also include the heterogeneity effects inside the node. 3 - Restrictions on the complexity of the problem: Number of axial nodes less than or equal 34. Number of material types less than or equal 30. Number of fuel assembly types less
Bar-code automated waste tracking system
International Nuclear Information System (INIS)
The Bar-Code Automated Waste Tracking System was designed to be a site-Specific program with a general purpose application for transportability to other facilities. The system is user-friendly, totally automated, and incorporates the use of a drive-up window that is close to the areas dealing in container preparation, delivery, pickup, and disposal. The system features ''stop-and-go'' operation rather than a long, tedious, error-prone manual entry. The system is designed for automation but allows operators to concentrate on proper handling of waste while maintaining manual entry of data as a backup. A large wall plaque filled with bar-code labels is used to input specific details about any movement of waste
Bar-code automated waste tracking system
Energy Technology Data Exchange (ETDEWEB)
Hull, T.E.
1994-10-01
The Bar-Code Automated Waste Tracking System was designed to be a site-Specific program with a general purpose application for transportability to other facilities. The system is user-friendly, totally automated, and incorporates the use of a drive-up window that is close to the areas dealing in container preparation, delivery, pickup, and disposal. The system features ``stop-and-go`` operation rather than a long, tedious, error-prone manual entry. The system is designed for automation but allows operators to concentrate on proper handling of waste while maintaining manual entry of data as a backup. A large wall plaque filled with bar-code labels is used to input specific details about any movement of waste.
Energy Technology Data Exchange (ETDEWEB)
Pavlovichev, A.M.
2001-06-19
The report presents calculation results of isotopic composition of irradiated fuel performed for the Quad Cities-1 reactor bundle with UO{sub 2} and MOX fuel. The MCU-REA code was used for calculations. The code is developed in Kurchatov Institute, Russia. The MCU-REA results are compared with the experimental data and HELIOS code results.
Validation of the COBRA code for dry out power calculation in CANDU type advanced fuels
International Nuclear Information System (INIS)
Stern Laboratories perform a full scale CHF testing of the CANFLEX bundle under AECL request. This experiment is modeled with the COBRA IV HW code to verify it's capacity for the dry out power calculation . Good results were obtained: errors below 10 % with respect to all data measured and 1 % for standard operating conditions in CANDU reactors range . This calculations were repeated for the CNEA advanced fuel CARA obtaining the same performance as the CANFLEX fuel. (author)
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.
WOLF: a computer code package for the calculation of ion beam trajectories
Energy Technology Data Exchange (ETDEWEB)
Vogel, D.L.
1985-10-01
The WOLF code solves POISSON'S equation within a user-defined problem boundary of arbitrary shape. The code is compatible with ANSI FORTRAN and uses a two-dimensional Cartesian coordinate geometry represented on a triangular lattice. The vacuum electric fields and equipotential lines are calculated for the input problem. The use may then introduce a series of emitters from which particles of different charge-to-mass ratios and initial energies can originate. These non-relativistic particles will then be traced by WOLF through the user-defined region. Effects of ion and electron space charge are included in the calculation. A subprogram PISA forms part of this code and enables optimization of various aspects of the problem. The WOLF package also allows detailed graphics analysis of the computed results to be performed.
A Students Attendance System Using QR Code
Directory of Open Access Journals (Sweden)
Fadi Masalha
2014-01-01
Full Text Available Smartphones are becoming more preferred companions to users than desktops or notebooks. Knowing that smartphones are most popular with users at the age around 26, using smartphones to speed up the process of taking attendance by university instructors would save lecturing time and hence enhance the educational process. This paper proposes a system that is based on a QR code, which is being displayed for students during or at the beginning of each lecture. The students will need to scan the code in order to confirm their attendance. The paper explains the high level implementation details of the proposed system. It also discusses how the system verifies student identity to eliminate false registrations.
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.
The solution of the LEU and MOX WWER-1000 calculation benchmark with the CARATE - multicell code
International Nuclear Information System (INIS)
Preparations for disposition of weapons grade plutonium in WWER-1000 reactors are in progress. Benchmark: Defined by the Kurchatov Institute (S. Bychkov, M. Kalugin, A. Lazarenko) to assess the applicability of computer codes for weapons grade MOX assembly calculations. Framework: 'Task force on reactor-based plutonium disposition' of OECD Nuclear Energy Agency. (Authors)
CPS: a continuous-point-source computer code for plume dispersion and deposition calculations
Energy Technology Data Exchange (ETDEWEB)
Peterson, K.R.; Crawford, T.V.; Lawson, L.A.
1976-05-21
The continuous-point-source computer code calculates concentrations and surface deposition of radioactive and chemical pollutants at distances from 0.1 to 100 km, assuming a Gaussian plume. The basic input is atmospheric stability category and wind speed, but a number of refinements are also included.
International Nuclear Information System (INIS)
The general purpose code BALTORO was written for coupling the three-dimensional Monte-Carlo /MC/ with the one-dimensional Discrete Ordinates /DO/ radiation transport calculations. The quantity of a radiation-induced /neutrons or gamma-rays/ nuclear effect or the score from a radiation-yielding nuclear effect can be analysed in this way. (author)
Sub-channel analysis by RELAP5 system code
Energy Technology Data Exchange (ETDEWEB)
Alessandro Petruzzi; Anis Bousbia Salah [DIMNP, Universit y of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy); Francesco D' Auria [DIMNP, Universit y of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy)
2005-07-01
Full text of publication follows: Recent progress in computer technology has increased the possibilities for code calculations in predicting realistically transient scenarios in nuclear power plants. Several attempts have been engaged in order to enlarge the domain for code applications, and to allow best estimate core simulation including interaction effects between neutronics and thermal-hydraulics. In this context, Relap5/Mod3.3 system thermalhydraulic code was used as a sub-channel code for the simulation of the low-pressure boil off experiment No 5002 of Neptun test facility. The experiment constitutes one of the separate effects test (SET) in the OECD/CSNI matrix for thermalhydraulic code validation related to phase separation and vertical flow 'with or without mixture level'. The drying out of the heated elements is expect to occur at very low coolant flow rates, low pressure (about 1.1 bar) and low power level (24.6 kW). The main aim of the activity discussed in the paper is to develop a 'nodalization technology' for accurately modeling the sub-channel grade void distribution problem and in the same way to assess the degree of success in using the Relap5 system code as a sub-channel code for the analysis of local quantities during transients in nuclear reactors. All thermal-hydraulic parameters, such as the collapsed liquid level, critical heat flux time occurrence and heaters surface temperature have been predicted with reasonable accuracy. A series of sensitivity analyses were also performed in order to assess the code prediction capabilities. More accurate results have been obtained considering the surface to surface radiation heat transfer model, as well as more cross flow nodes between the test section rods. The overall analysis confirms the possibility of using the Relap5/Mod3.3 system thermal-hydraulic code as sub-channel code to predict the evolution of relevant local quantities measured during 'relevant' experiments
IPEN/MB-01 heavy reflector benchmark calculations using Serpent code
International Nuclear Information System (INIS)
A series of critical experiments with water-moderated square-pitched lattices with low-enriched uranium fuel rods was conducted at the IPEN/MB-01 research reactor facility, in 2005. Later, this data become some benchmarks. In one of these experiments the west face of the reactor core was covered with a set of thin SS-304 plates to simulate a heavy reflector as used in the EPR reactor (LEU-COMP-HERM-043). The plates are 3 mm thick and their width and axial length were large enough to cover one whole side of the active core of the reactor. The critical configurations were found as a function of the number of plates. Fuel rods containing UO2 with uranium enriched to 4.3% 235U were arranged in specific geometric configurations to be as close as possible to the critical state. In this work, these benchmark configurations with heavy reflectors were modeled using the Serpent Monte Carlo Code. Serpent uses a universe-based geometry model, which allows the description of practically any three-dimensional fuel or reactor configuration. Neutron transport is based on a combination of surface-to-surface ray-tracing and the Woodcock delta-tracking method. Woodcock method is many times faster than ray-tracing, so compared to MCNP code, Serpent code can bring huge gains in processing time of reactor calculations and reaction rate calculations. The results of these calculations were compared with experimental data and calculations with codes MCNP5 and SCALE6 (KENO-VI) using ENDF/B-VII.0 as cross-section input data. The codes performances are compared in terms of CPU calculation time and agreement with experimental data. Additional y, sensitivity on keff of Serpent woodcock threshold parameter was analyzed. (author)
EAI-oriented information classification code system in manufacturing enterprises
Institute of Scientific and Technical Information of China (English)
Junbiao WANG; Hu DENG; Jianjun JIANG; Binghong YANG; Bailing WANG
2008-01-01
Although the traditional information classifi-cation coding system in manufacturing enterprises (MEs) emphasizes the construction of code standards, it lacks the management of the code creation, code data transmission and so on. According to the demands of enterprise application integration (EAI) in manufacturing enter-prises, an enterprise application integration oriented information classification code system (EAIO-ICCS) is proposed. EAIO-ICCS expands the connotation of the information classification code system and assures the identity of the codes in manufacturing enterprises with unified management of codes at the view of its lifecycle.
Comparison: RELAP5-3D systems analysis code and fluent CFD code momentum equation formulations
International Nuclear Information System (INIS)
Recently the Idaho National Engineering and Environmental Laboratory (INEEL), in conjunction with Fluent Corporation, have developed a new analysis tool by coupling the Fluent computational fluid dynamics (CFD) code to the RELAP5-3D advanced thermal-hydraulic analysis code. This tool enables researchers to perform detailed, two- or three-dimensional analyses using Fluent's CFD capability while the boundary conditions required by the Fluent calculation are provided by the balance-of-system model created using RELAP5-3D. Fluent and RELAP5-3D have strengths that complement one another. CFD codes, such as Fluent, are commonly used to analyze the flow behavior in regions of a system where complex flow patterns are expected or present. On the other hand, RELAP5-3D was developed to analyze the behavior of two-phase systems that could be modeled in one-dimension. Empirical relationships were used where first-principle physics were not well developed. Both Fluent and RELAP5-3D are exemplary in their areas of specialization. The differences between Fluent and RELAP5 fundamentally stem from their field equations. This study focuses on the differences between the momentum equation representations in the two codes (the continuity equation formulations are equivalent for single phase flow). First the differences between the momentum equations are summarized. Next the effect of the differences in the momentum equations are examined by comparing the results obtained using both codes to study the same problem, i.e., fully-developed turbulent pipe flow. Finally, conclusions regarding the significance of the differences are given. (author)
International Nuclear Information System (INIS)
Three computer codes MATHEW, ADPIC of LLNL and GAMPUL of JAERI for prediction of wind field, concentration and external exposure rate of airborne radioactive materials are vectorized and the results are presented. Using the continuous equation of incompressible flow as a constraint, the MATHEW calculates the three dimensional wind field by a variational method. Using the particle-in -cell method, the ADPIC calculates the advection and diffusion of radioactive materials in three dimensional wind field and terrain, and gives the concentration of the materials in each cell of the domain. The GAMPUL calculates the external exposure rate assuming Gaussian plume type distribution of concentration. The vectorized code MATHEW attained 7.8 times speedup by a vector processor FACOM230-75 APU. The ADPIC and GAMPUL are estimated to attain 1.5 and 4 times speedup respectively on CRAY-1 type vector processor. (author)
Development of Three-dimensional Reactor Analysis Code System for Accelerator-Driven System, ADS3D
International Nuclear Information System (INIS)
To investigate an Accelerator-Driven System (ADS) with sub-criticality control mechanism such as control rods or burnable poison, the ADS3D code has been developed on MARBLE which is a next generation reactor analysis code system developed by JAEA. In the past neutronics calculation for the ADS, JAEA employed RZ calculation models to realize efficient investigations. However, it was very difficult to model sub-criticality control mechanisms in RZ calculation models. The ADS3D code system is able to calculate the transportation of protons and neutrons, the burn-up calculation and the fuel exchange in three-dimensional calculation models. It means this code system can treat ADS concepts with sub-criticality control mechanism and makes it possible to investigate a new concept of ADS. (author)
A Students Attendance System Using QR Code
Fadi Masalha; Nael Hirzallah
2014-01-01
Smartphones are becoming more preferred companions to users than desktops or notebooks. Knowing that smartphones are most popular with users at the age around 26, using smartphones to speed up the process of taking attendance by university instructors would save lecturing time and hence enhance the educational process. This paper proposes a system that is based on a QR code, which is being displayed for students during or at the beginning of each lecture. The students will need to scan the co...
Cooperative Regenerating Codes for Distributed Storage Systems
Shum, Kenneth W.
2011-01-01
When there are multiple node failures in a distributed storage system, regenerating the failed storage nodes individually in a one-by-one manner is suboptimal as far as repair-bandwidth minimization is concerned. If data exchange among the newcomers is enabled, we can get a better tradeoff between repair bandwidth and the storage per node. An explicit and optimal construction of cooperative regenerating code is illustrated.
Blind Calculation of RD-14M Small Break LOCA Tests by CATHENA Code
International Nuclear Information System (INIS)
KAERI participated with the computer code CATHENA, which is used to analyze Pressurized Heavy Water Reactors (PHWRs), in an IAEA International Collaborative Standard Problem (ICSP) with the objective to benchmark and validate thermal-hydraulic computer code against qualified data for Small Break Loss of Coolant Accident (SBLOCA) scenario generated on RD-14M Test Facility. Two specific SBLOCA tests selected for this ICSP titled 'Comparison of HWR Code Predictions with SBLOCA Experimental Data', are B9006 and B9802. Test B9006 is a 7-mm inlet header break experiment with pressurized accumulator emergency coolant injection and represents most complete SBLOCA test conducted in RD-14M. Test B9802 is a 3-mm inlet header break experiment with full channel power to study boiling in channels and condensation in steam generators in a slowly depressurizing loop rather than a blow down. This report presents the blind calculation results for these tests conducted by CATHENA code before the test data are distributed to participants. For B9006 test, CATHENA code simulated all the phases of the transient such as blowdown, high-pressure ECI, secondary pressure ramp, refill, switch from high pressure ECI to low pressure ECI, exponential pump ramp, and natural circulation. For B9802 test, CATHENA calculation was intended to predict temperature rise of the FES sheath due to channel boiling, and power supply trip on high FES sheath temperature (600 .deg. C) process protection trip
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.)
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.
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.
Development of an effective delayed neutron fraction calculation code, BETA-K
Energy Technology Data Exchange (ETDEWEB)
Kim, Taek Kyum; Song, Hoon; Kim, Young Il; Kim, Young In; Kim, Young Jin [Korea Atomic Energy Research Institute, Taejon (Korea)
1998-08-01
BETA-K, an effective delayed neutron fraction calculation code consistent with Nodal Expansion Method (NEM), has been developed. By using relevant output files of DIF3D code, it can calculate the effective delayed neutron fraction({beta}{sub eff}), neutron lifetime(l{sub eff}), fission spectrum ({chi}-bar) and fission yield data({nu}) for each fissionable isotope, composition of fuels and over the whole core. BETA-K code has been validated by comparing the calculated values to the measured ones of effective delayed neutron fraction in two critical experiments, BFS73-1 and BFS55-1. BFS73-1 is a metal uranium core and BFS55-1 is a metal plutonium core. The C/E values, 1.007 and 0.992 for BFS73-1 and BFS55-1 respectively, agreed well with the experimental values within the experiment errors. BETA-K code predicts 0.00709 and 0.356 {mu}sec as the effective delayed neutron fraction and neutron life time for the uranium metallic fueled equilibrium core of 150MWe KALIMER. (author). 9 refs., 6 figs., 12 tabs.
Linear calculations of edge current driven kink modes with BOUT++ code
International Nuclear Information System (INIS)
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
The spectral code Apollo2: from lattice to 2D core calculations
Energy Technology Data Exchange (ETDEWEB)
Coste-Delclaux, M.; Santandrea, S.; Damian, F.; Blanc-Tranchant, P.; Zmijarevic, I. [CEA Saclay (DEN/DANS/SERMA), 91 - Gif-sur-Yvette (France); Santamarina, A. [CEA Cadarache (CEA/DEN/DER/SPRC), 13 - Saint Paul lez Durance (France)
2005-07-01
Apollo2 is a powerful code dedicated to neutron transport, it is a highly qualified tool for a wide range of applications from research and development studies to industrial applications. Today Apollo2 is part of several advanced 3-dimensional nuclear code packages dedicated to reactor physics, fuel cycle, criticality and safety analysis. The presentations have been organized into 7 topics: -) an introduction to Apollo2, -) cross-sections, -) flux calculation, -) advanced applications, -) Apollo2 users, specialized packages, -) qualification program, and -) the future of Apollo2. This document gathers only the slides of the presentations.
International Nuclear Information System (INIS)
The paper presents the computer code Mitra (Multicomponent isotope transport) which has been constructed to calculate the release of radioactive fission products from nuclear fuels under non-stationary conditions. The code is based on a new integration method fo the mass transport equation in the presence of precipitation, re-solution and radioactive decay. The starting equations and the assumed physical models are briefly described in the main part of the report. A very detailed description of the formulae used and of the Mitra subprograms are presented in extended appendices
A computer code for calculations in the algebraic collective model of the atomic nucleus
Welsh, T. A.; Rowe, D. J.
2016-03-01
A Maple code is presented for algebraic collective model (ACM) calculations. The ACM is an algebraic version of the Bohr model of the atomic nucleus, in which all required matrix elements are derived by exploiting the model's SU(1 , 1) × SO(5) dynamical group. This paper reviews the mathematical formulation of the ACM, and serves as a manual for the code. The code enables a wide range of model Hamiltonians to be analysed. This range includes essentially all Hamiltonians that are rational functions of the model's quadrupole moments qˆM and are at most quadratic in the corresponding conjugate momenta πˆN (- 2 ≤ M , N ≤ 2). The code makes use of expressions for matrix elements derived elsewhere and newly derived matrix elements of the operators [ π ˆ ⊗ q ˆ ⊗ π ˆ ] 0 and [ π ˆ ⊗ π ˆ ] LM. The code is made efficient by use of an analytical expression for the needed SO(5)-reduced matrix elements, and use of SO(5) ⊃ SO(3) Clebsch-Gordan coefficients obtained from precomputed data files provided with the 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.
Calculation of Gamma-ray Responses for HPGe Detectors with TRIPOLI-4 Monte Carlo Code
Lee, Yi-Kang; Garg, Ruchi
2014-06-01
The gamma-ray response calculation of HPGe (High Purity Germanium) detector is one of the most important topics of the Monte Carlo transport codes for nuclear instrumentation applications. In this study the new options of TRIPOLI-4 Monte Carlo transport code for gamma-ray spectrometry were investigated. Recent improvements include the gamma-rays modeling of the electron-position annihilation, the low energy electron transport modeling, and the low energy characteristic X-ray production. The impact of these improvements on the detector efficiency of the gamma-ray spectrometry calculations was verified. Four models of HPGe detectors and sample sources were studied. The germanium crystal, the dead layer of the crystal, the central hole, the beryllium window, and the metal housing are the essential parts in detector modeling. A point source, a disc source, and a cylindrical extended source containing a liquid radioactive solution were used to study the TRIPOLI-4 calculations for the gamma-ray energy deposition and the gamma-ray self-shielding. The calculations of full-energy-peak and total detector efficiencies for different sample-detector geometries were performed. Using TRIPOLI-4 code, different gamma-ray energies were applied in order to establish the efficiency curves of the HPGe gamma-ray detectors.
Emergency Doses (ED) - Revision 3: A calculator code for environmental dose computations
International Nuclear Information System (INIS)
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
Reference calculations on critical assemblies with Apollo2 code working with a fine multigroup mesh
International Nuclear Information System (INIS)
The objective of this thesis is to add to the multigroup transport code APOLLO2 the capability to perform deterministic reference calculations, for any type of reactor, using a very fine energy mesh of several thousand groups. This new reference tool allows us to validate the self-shielding model used in industrial applications, to perform depletion calculations, differential effects calculations, critical buckling calculations or to evaluate precisely data required by the self shielding model. At its origin, APOLLO2 was designed to perform routine calculations with energy meshes around one hundred groups. That is why, in the current format of cross sections libraries, almost each value of the multigroup energy transfer matrix is stored. As this format is not convenient for a high number of groups (concerning memory size), we had to search out a new format for removal matrices and consequently to modify the code. In the new format we found, only some values of removal matrices are kept (these values depend on a reconstruction precision choice), the other ones being reconstructed by a linear interpolation, what reduces the size of these matrices. Then we had to show that APOLLO2 working with a fine multigroup mesh had the capability to perform reference calculations on any assembly geometry. For that, we successfully carried out the validation with several calculations for which we compared APOLLO2 results (obtained with the universal mesh of 11276 groups) to results obtained with Monte Carlo codes (MCNP, TRIPOLI4). Physical analysis led with this new tool have been very fruitful and show a great potential for such an R and D tool. (author)
Wall-touching kink mode calculations with the M3D code
International Nuclear Information System (INIS)
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
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.
Post test calculations of a severe accident experiment for VVER-440 reactors by the ATHLET code
Energy Technology Data Exchange (ETDEWEB)
Gyoergy, Hunor [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques (BME NTI); Trosztel, Istvan [Hungarian Academy of Sciences, Budapest (Hungary). Centre for Energy Research (MTA EK)
2013-09-15
Severe accident - if no mitigation action is taken - leads to core melt. An effective severe accident management strategy can be the external reactor pressure vessel cooling for corium localization and stabilization. For some time discussion was going on, whether the in-vessel retention can be applied for the VVER-440 type reactors. It had to be demonstrated that the available space between the reactor vessel and biological protection allows sufficient cooling to keep the melted core in the vessel, without the reactor pressure vessel losing its integrity. In order to demonstrate the feasibility of the concept an experimental facility was realized in Hungary. The facility called Cooling Effectiveness on the Reactor External Surface (CERES) is modeling the vessel external surface and the biological protection of Paks NPP. A model of the CERES facility for the ATHLET TH system code was developed. The results of the ATHLET calculation agree well with the measurements showing that the vessel cooling can be insured for a long time in a VVER-440 reactor. (orig.)
SPLOSH III. A code for calculating reactivity and flow transients in CSGHWR
International Nuclear Information System (INIS)
SPLOSH is a time dependent, one dimensional, finite difference (in time and space) coupled neutron kinetics and thermal hydraulics code for studying pressurised faults and control transients in water reactor systems. An axial single channel model with equally spaced mesh intervals is used to represent the neutronics of the reactor core. A radial finite difference model is used for heat conduction through the fuel pin, gas gap and can. Appropriate convective, boiling or post-dryout heat transfer correlations are used at the can-coolant interface. The hydraulics model includes the important features of the SGHWR primary loop including 'slave' channels in parallel with the 'mean' channel. Standard mass, energy and momentum equations are solved explicitly. Circuit features modelled include pumps, spray cooling and the SGHWR steam drum. Perturbations to almost any feature of the circuit model may be specified by the user although blowdown calculations resulting in critical or reversed flows are not permitted. Automatic reactor trips may be defined and the ensuing actions of moderator dumping and rod firing can be specified. (UK)
Design of a transport calculation system for logging sondes simulation
International Nuclear Information System (INIS)
Analysis of available resources in earth crust is performed by different techniques, one of them is neutron logging. Design of sondes that are used to make such logging is supported by laboratory experiments as well as by numerical calculations.This work presents several calculation schemes, designed to simplify the task of whom has to planify such experiments or optimize parameters of this kind of sondes.These schemes use transport calculation codes, especially DaRT, TORT and MCNP, and cross section processing modules from SCALE system.Additionally a system for DaRT and TORT data postprocessing using OpenDX is presented.It allows scalar flux spatial distribution analysis, as wells as cross section condensation and reaction rates calculation
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.
Comprehensive nuclear model calculations: Introduction to the theory and use of the GNASH code
International Nuclear Information System (INIS)
A user's manual describing the theory and operation of the GNASH nuclear reaction computer code is presented. This work is based on a series of lectures describing the statistical Hauser-Feshbach plus preequilibrium version of the code with full angular momentum conservation. This version is expected to be most applicable for incident particle energies between 1 key and 50 MeV. General features of the code, the nuclear models that are utilized, input parameters needed to perform calculations, and the output quantities from typical problems are described in detail. The computational structure of the code and the subroutines and functions that are called are summarized as well. Two detailed examples are considered: 14-MeV neutrons incident on 93Nb and 12-MeV neutrons incident on 238U. The former example illustrates a typical calculation aimed at determining neutron, proton, and alpha emission spectra from 14-MeV reactions, and the latter example demonstrates use of the fission model in GNASH
Problems of optimal data coding in hodoscopic systems
International Nuclear Information System (INIS)
An analogy system of algebraic coding theory and of hodoscopic system coding theory is considered. The connection between main parameters of coding devices and parameters of parallel coders applied in hodoscopic systems is established. The efficiency of using a proposed analogy system is illustrated on some examples of designing parallel coders with given properties
Calculation of effective delayed neutron fraction with modified library of Monte Carlo code
International Nuclear Information System (INIS)
Highlights: ► We propose a new Monte Carlo method to calculate the effective delayed neutron fraction by changing the library. ► We study the stability of our method. When the particles and cycles are sufficiently great, the stability is very good. ► The final result is determined to make the deviation least. ► We verify our method on several benchmarks, and the results are very good. - Abstract: A new Monte Carlo method is proposed to calculate the effective delayed neutron fraction βeff. Based on perturbation theory, βeff is calculated with modified library of Monte Carlo code. To verify the proposed method, calculations are performed on several benchmarks. The error of the method is analyzed and the way to reduce error is proposed. The results are in good agreement with the reference data
BETHSY 6.2TC test calculation with TRACE and RELAP5 computer code
International Nuclear Information System (INIS)
The TRACE code is still under development and it will have all capabilities of RELAP5. The purpose of the present study was therefore to assess the accuracy of the TRACE calculation of BETHSY 6.2TC test, which is 15.24 cm equivalent diameter horizontal cold leg break. For calculations the TRACE V5.0 Patch 1 and RELAP5/MOD3.3 Patch 4 were used. The overall results obtained with TRACE were similar to the results obtained by RELAP5/MOD3.3. The results show that the discrepancies were reasonable. (author)
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)
International Nuclear Information System (INIS)
For safety evaluation of nuclear fuel facilities, a nuclear decay data library named JDDL and a computer code COMRAD have been developed to calculate isotopic composition of each nuclide, radiation source intensity, energy spectrum of γ-ray and neutron, and decay heat of spent fuel. JDDL has been produced mainly from the evaluated nuclear data file ENSDF to use new nuclear data. To supplement the data file for short life nuclides, the JNDC data set were also used which had been evaluated by Japan Nuclear Data Committee. Using these data, calculations became possible from short period to long period after irradiation. (author)
Use of generalized curvilinear coordinate systems in electromagnetic and hybrid codes
Energy Technology Data Exchange (ETDEWEB)
Swift, D.W. [Univ. of Alaska, Fairbanks, AK (United States)
1995-07-01
The author develops a code to simulate the dynamics in the magnetosphere system. The calculation involves a single level, structured, curvilinear 2D mesh. The mesh density is varied to support regions which demand higher resolution.
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.
System analysis of bar code laser scanner
Wang, Jianpu; Chen, Zhaofeng; Lu, Zukang
1996-10-01
This paper focuses on realizing the three important aspects of bar code scanner: generating a high quality scanning light beam, acquiring a fairly even distribution characteristic of light collection, achieving a low signal dynamic range over a large depth of field. To do this, we analyze the spatial distribution and propagation characteristics of scanning laser beam, the vignetting characteristic of optical collection system and their respective optimal design; propose a novel optical automatic gain control method to attain a constant collection over a large working depth.
Deep-penetration calculation for the ISIS target station shielding using the MARS Monte Carlo code
International Nuclear Information System (INIS)
A calculation of neutron penetration through a thick shield was performed with a three-dimensional multi-layer technique using the MARS14(02) Monte Carlo code to compare with the experimental shielding data in 1998 at the ISIS spallation neutron source facility. In this calculation, secondary particles from a tantalum target bombarded by 800-MeV protons were transmitted through a bulk shield of approximately 3-m-thick iron and 1-m-thick concrete. To accomplish this deep-penetration calculation with good statistics, the following three techniques were used in this study. First, the geometry of the bulk shield was three-dimensionally divided into several layers of about 50-cm thickness, and a step-by-step calculation was carried out to multiply the number of penetrated particles at the boundaries between the layers. Second, the source particles in the layers were divided into two parts to maintain the statistical balance on the spatial-flux distribution. Third, only high-energy particles above 20 MeV were transported up to approximately 1 m before the region for benchmark calculation. Finally, the energy spectra of neutrons behind the very thick shield were calculated down to the thermal energy with good statistics, and typically agree well within a factor of two with the experimental data over a broad energy range. The 12C(n,2n)11C reaction rates behind the bulk shield were also calculated, which agree with the experimental data typically within 60%. These results are quite impressive in calculation accuracy for deep-penetration problem. In this report, the calculation conditions, geometry and the variance reduction techniques used in the deep-penetration calculation with the MARS14 code are clarified, and several subroutines of MARS14 which were used in our calculation are also given in the appendix. The numerical data of the calculated neutron energy spectra, reaction rates, dose rates and their C/E (Calculation/Experiment) values are also summarized. The
Dubois, Vincent; Desbiens, Nicolas; Auroux, Eric
2010-07-01
We present the improvements of the CARTE thermochemical code which provides thermodynamic properties and chemical compositions of CHON systems over a large range of temperature and pressure with a very small computational cost. The detonation products are split in one or two fluid phase (s), treated with the MCRSR equation of state (EOS), and one condensed phase of carbon, modeled with a multiphase EOS which evolves with the chemical composition of the explosives. We have developed a new optimization procedure to obtain an accurate multicomponents EOS. We show here that the results of CARTE code are in good agreement with the specific data of molecular systems and measured detonation properties for several explosives.
MTR coded PRML systems for perpendicular magnetic recording
Energy Technology Data Exchange (ETDEWEB)
Okamoto, Yoshihiro E-mail: okamoto@rec.ee.ehime-u.ac.jp; Sato, Mitsuteru; Osawa, Hisashi; Saito, Hidetoshi; Muraoka, Hiroaki; Nakamura, Yoshihisa
2001-10-01
We evaluate the BER performance of various MTR coded PRML systems characterized by the polynomials with only positive coefficients in a perpendicular magnetic recording channel using a double-layered medium with jitter-like noise by computer simulation. The results show that ((3)/(4)) MTR coded PRML systems exhibit good performances compared with ((16)/(17)) MTR coded PRML systems.
DESIGN OF EXACT REGENERATING HIERARCHICAL CODE FOR DISTRIBUTED STORAGE SYSTEM
Institute of Scientific and Technical Information of China (English)
Hao Jie; Lu Yanbo; Liu Xinji; Xia Shutao
2013-01-01
Erasure code is widely used as the redundancy scheme in distributed storage system.When a storage node fails,the repair process often requires to transfer a large amount of data.Regenerating code and hierarchical code are two classes of codes proposed to reduce the repair bandwidth cost.Regenerating codes reduce the amount of data transferred by each helping node,while hierarchical codes reduce the number of nodes participating in the repair process.In this paper,we propose a "sub-code nesting framework" to combine them together.The resulting regenerating hierarchical code has low repair degree as hierarchical code and lower repair cost than hierarchical code.Our code can achieve exact regeneration of the failed node,and has the additional property of low updating complexity.
Program POD. A computer code to calculate cross sections for neutron-induced nuclear reactions
International Nuclear Information System (INIS)
A computer code, POD, was developed for neutron-induced nuclear data evaluations. This program is based on four theoretical models, (1) the optical model to calculate shape-elastic scattering and reaction cross sections, (2) the distorted wave Born approximation to calculate neutron inelastic scattering cross sections, (3) the preequilibrium model, and (4) the multi-step statistical model. With this program, cross sections can be calculated for reactions (n, γ), (n, n'), (n, p), (n, α), (n, d), (n, t), (n, 3He), (n, 2n), (n, np), (n, nα), (n, nd), and (n, 3n) in the neutron energy range above the resonance region to 20 MeV. The computational methods and input parameters are explained in this report, with sample inputs and outputs. (author)
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.)
Recent R and D around the Monte-Carlo code Tripoli-4 for criticality calculation
Energy Technology Data Exchange (ETDEWEB)
Hugot, F.X.; Lee, Y.K.; Malvagi, F. [CEA - DEN/DANS/DM2S/SERMA/LTSD, Saclay (France)
2008-07-01
TRIPOLI-4 [1] is the fourth generation of the TRIPOLI family of Monte Carlo codes developed from the 60's by CEA. It simulates the 3D transport of neutrons, photons, electrons and positrons as well as coupled neutron-photon propagation and electron-photons cascade showers. The code addresses radiation protection and shielding problems, as well as criticality and reactor physics problems through both critical and subcritical neutronics calculations. It uses full pointwise as well as multigroup cross-sections. The code has been validated through several hundred benchmarks as well as measurement campaigns. It is used as a reference tool by CEA as well as its industrial and institutional partners, and in the NURESIM [2] European project. Section 2 reviews its main features, with emphasis on the latest developments. Section 3 presents some recent R and D for criticality calculations. Fission matrix, Eigen-values and eigenvectors computations will be exposed. Corrections on the standard deviation estimator in the case of correlations between generation steps will be detailed. Section 4 presents some preliminary results obtained by the new mesh tally feature. The last section presents the interest of using XML format output files. (authors)
THYDE-NEU: Nuclear reactor system analysis code
International Nuclear Information System (INIS)
THYDE-NEU is applicable not only to transient analyses, but also to steady state analyses of nuclear reactor systems (NRSs). In a steady state analysis, the code generates a solution satisfying the transient equations without external disturbances. In a transient analysis, the code calculates temporal NRS behaviors in response to various external disturbances in such a way that mass and energy of the coolant as well as the number of neutrons conserve. The first half of the report is the description of the methods and models for use in the THYDE-NEU code, i.e., (1) the thermal-hydraulic network model, (2) the spatial kinetics model, (3) the heat sources in fuel, (4) the heat transfer correlations, (5) the mechanical behavior of clad and fuel, and (6) the steady state adjustment. The second half of the report is the users' mannual containing the items; (1) the program control, (2) the input requirements, (3) the execution of THYDE-NEU jobs, (4) the output specifications and (5) the sample calculation. (author)
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
Calculation of the Phenix end-of-life test “control rod withdrawal” with the ERANOS code
International Nuclear Information System (INIS)
The Institute of Radioprotection and Nuclear Safety (IRSN) being established as technical support organization for French public authorities is in charge of safety assessment of both operating and under construction reactors and nuclear facilities. It provides safety studies of advanced and innovative projects like fast sodium cooled reactors as well. In this context, one of the IRSN objectives is to evaluate comprehensively the accuracy of numerical tools and their performance on studies of safety relay items. Reactor physics studies step in the safety assessment support from different points of view, among which the design of core and its protection system. They are essential in the cores behavior analysis in normal, perturbed and accidental conditions in order to assess the integrity of the first barrier and the exclusion of prompt criticality and re-criticality risks. The codes capability to compute in an accurate manner the fission power distribution in the core during the whole reactor lifetime could indicate the codes' accuracy for many so-called spatial dependent values calculations. The IAEA Coordinated Research Project on the Phenix end-of-life test “Control Rod Withdrawal” has been a good opportunity to check the capability of calculation tools by comparison with the measured radial power distributions on fast reactor. IRSN participated to this benchmark with the ERANOS code package developed by CEA for fast reactors studies. The challenge for this code package was that in the considered core configurations the neutron fields were notably deformed. This paper presents the results obtained in the framework of the benchmark activity. A relatively good agreement has been found with available measures considering the approximations done in the modeling. The work underlines the importance of precise knowledge of the details of burn-up distribution as it could impact the calculations of the power distribution. (author)
Calculation system analysis for radiation shielding
International Nuclear Information System (INIS)
This work consists of the computational system implementation for nuclear reactor shielding analysis. The system has as objectives to facilitate the installation of the calculation framework, problem set-up, and results analysis. Several computational programmes commonly used for cross-section preparation and radiation transport were chosen for the system. This work represents the capacity necessary for nuclear reactor and particle accelerator shielding design, to aid in nuclear experiments and in the utilization of nuclear techniques that require the radiation field calculation. The system was implemented in PC-DOS environment and consists of the necessary and sufficient programs and data for generation of the cross sections, groups constants, self-shielding factors, activation sources, for the calculation of neutron and gamma-ray fluence, dose rates, and other types of response functions. (author). 11 refs., 8 figs
Development of burnup calculation function in reactor Monte Carlo code RMC
International Nuclear Information System (INIS)
This paper presents the burnup calculation capability of RMC, which is a new Monte Carlo (MC) neutron transport code developed by Reactor Engineering Analysis Laboratory (REAL) in Tsinghua University of China. Unlike most of existing MC depletion codes which explicitly couple the depletion module, RMC incorporates ORIGEN 2.1 in an implicit way. Different burn step strategies, including the middle-of-step approximation and the predictor-corrector method, are adopted by RMC to assure the accuracy under large burnup step size. RMC employs a spectrum-based method of tallying one-group cross section, which can considerably saves computational time with negligible accuracy loss. According to the validation results of benchmarks and examples, it is proved that the burnup function of RMC performs quite well in accuracy and efficiency. (authors)
Stellarator-specific developments for the systems code PROCESS
Energy Technology Data Exchange (ETDEWEB)
Warmer, Felix; Beidler, Craig; Dinklage, Andreas; Feng, Yuehe; Geiger, Joachim; Schauer, Felix; Turkin, Yuriy; Wolf, Robert; Xanthopoulos, Pavlos [Max-Planck-Institut fuer Plasmaphysik, Wendelsteinstrasse 1, D-17491 Greifswald (Germany); Knight, Peter; Ward, David [Culham Centre for Fusion Energy, Abingdon, Oxfordshire, OX14 3DB (United Kingdom)
2014-07-01
The ultimate goal of fusion research is to demonstrate the feasibility of economic production of electricity. The most promising concepts to achieve this by magnetic confinement are the Tokamak and the Stellarator. System codes are used to study the general properties of a fusion power plant. Built in a modular way systems codes describe the physical and technical properties of the power plant components. For the Helical Advanced Stellarator (HELIAS) concept modules have been developed in the frame of the existing Tokamak systems code PROCESS. These include: A geometry model based on Fourier coefficients which represent the complex 3-D plasma shape, a divertor model which assumes diffusive cross-field transport and high radiation at the X-point, a coil model which uses a scaling based on the HELIAS design and a transport model which either employs empirical confinement time scalings or sophisticated 1-D collisional and turbulent transport calculations. This approach aims at a direct comparison between Tokamak and Stellarator power plant designs.
Risk calculation method for complex engineering system
Directory of Open Access Journals (Sweden)
Li-ping WANG
2011-09-01
Full Text Available This paper presents a rapid and simple risk calculation method for large and complex engineering systems, the simulated maximum entropy method (SMEM, which is based on integration of the advantages of the Monte Carlo and maximum entropy methods, thus avoiding the shortcoming of the slow convergence rate of the Monte Carlo method in risk calculation. Application of SMEM in the calculation of reservoir flood discharge risk shows that this method can make full use of the known information under the same conditions and obtain the corresponding probability distribution and the risk value. It not only greatly improves the speed, compared with the Monte Carlo method, but also provides a new approach for the risk calculation in large and complex engineering systems.
DCHAIN-SP 2001: High energy particle induced radioactivity calculation code
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)
2001-03-01
For the purpose of contribution to safety design calculations for induced radioactivities in the JAERI/KEK high-intensity proton accelerator project facilities, the DCHAIN-SP which calculates the high energy particle induced radioactivity has been updated to DCHAIN-SP 2001. The following three items were improved: (1) Fission yield data are included to apply the code to experimental facility design for nuclear transmutation of long-lived radioactive waste where fissionable materials are treated. (2) Activation cross section data below 20 MeV are revised. In particular, attentions are paid to cross section data of materials which have close relation to the facilities, i.e., mercury, lead and bismuth, and to tritium production cross sections which are important in terms of safety of the facilities. (3) User-interface for input/output data is sophisticated to perform calculations more efficiently than that in the previous version. Information needed for use of the code is attached in Appendices; the DCHAIN-SP 2001 manual, the procedures of installation and execution of DCHAIN-SP, and sample problems. (author)
Energy Technology Data Exchange (ETDEWEB)
Kalin, J., E-mail: jan.kalin@zag.s [Slovenian National Building and Civil Engineering Institute, Dimiceva 12, SI-1000 Ljubljana (Slovenia); Petkovsek, B., E-mail: borut.petkovsek@zag.s [Slovenian National Building and Civil Engineering Institute, Dimiceva 12, SI-1000 Ljubljana (Slovenia); Montarnal, Ph., E-mail: philippe.montarnal@cea.f [CEA/Saclay, DM2S/SFME/LSET, Gif-sur-Yvette, 91191 cedex (France); Genty, A., E-mail: alain.genty@cea.f [CEA/Saclay, DM2S/SFME/LSET, Gif-sur-Yvette, 91191 cedex (France); Deville, E., E-mail: estelle.deville@cea.f [CEA/Saclay, DM2S/SFME/LSET, Gif-sur-Yvette, 91191 cedex (France); Krivic, J., E-mail: jure.krivic@geo-zs.s [Geological Survey of Slovenia, Dimiceva 14, SI-1000 Ljubljana (Slovenia); Ratej, J., E-mail: joze.ratej@geo-zs.s [Geological Survey of Slovenia, Dimiceva 14, SI-1000 Ljubljana (Slovenia)
2011-04-15
In the past years the Slovenian Performance Analysis/Safety Assessment team has performed many generic studies for the future Slovenian low and intermediate level waste repository, most recently a Special Safety Analysis for the Krsko site. The modelling approach was to split the problem into three parts: near-field (detailed model of the repository), far-field (i.e., geosphere) and biosphere. In the Special Safety Analysis the code used to perform the near-field calculations was Hydrus2D. Recently the team has begun a cooperation with the French Commisariat al'Energie Atomique/Saclay (CEA/Saclay) and, as a part of this cooperation, began investigations into using the Alliances numerical platform for near-field calculations in order to compare the overall approach and calculated results. The article presents the comparison between these two codes for a silo-type repository that was considered in the Special Safety Analysis. The physical layout and characteristics of the repository are presented and a hydraulic and transport model of the repository is developed and implemented in Alliances. Some analysis of sensitivity to mesh fineness and to simulation timestep has been preformed and is also presented. The compared quantity is the output flux of radionuclides on the boundary of the model. Finally the results from Hydrus2D and Alliances are compared and the differences and similarities are commented.
ORPHEE research reactor: 3D core depletion calculation using Monte-Carlo code TRIPOLI-4®
Damian, F.; Brun, E.
2014-06-01
ORPHEE is a research reactor located at CEA Saclay. It aims at producing neutron beams for experiments. This is a pool-type reactor (heavy water), and the core is cooled by light water. Its thermal power is 14 MW. ORPHEE core is 90 cm height and has a cross section of 27x27 cm2. It is loaded with eight fuel assemblies characterized by a various number of fuel plates. The fuel plate is composed of aluminium and High Enriched Uranium (HEU). It is a once through core with a fuel cycle length of approximately 100 Equivalent Full Power Days (EFPD) and with a maximum burnup of 40%. Various analyses under progress at CEA concern the determination of the core neutronic parameters during irradiation. Taking into consideration the geometrical complexity of the core and the quasi absence of thermal feedback for nominal operation, the 3D core depletion calculations are performed using the Monte-Carlo code TRIPOLI-4® [1,2,3]. A preliminary validation of the depletion calculation was performed on a 2D core configuration by comparison with the deterministic transport code APOLLO2 [4]. The analysis showed the reliability of TRIPOLI-4® to calculate a complex core configuration using a large number of depleting regions with a high level of confidence.
A computer code for beam optics calculation--third order approximation
Institute of Scientific and Technical Information of China (English)
L(U) Jianqin; LI Jinhai
2006-01-01
To calculate the beam transport in the ion optical systems accurately, a beam dynamics computer program of third order approximation is developed. Many conventional optical elements are incorporated in the program. Particle distributions of uniform type or Gaussian type in the ( x, y, z ) 3D ellipses can be selected by the users. The optimization procedures are provided to make the calculations reasonable and fast. The calculated results can be graphically displayed on the computer monitor.
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.
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.
Improved FEC Code Based on Concatenated Code for Optical Transmission Systems
Institute of Scientific and Technical Information of China (English)
YUAN Jian-guo; JIANG Ze; MAO You-ju
2006-01-01
The improved three novel schemes of the super forward error correction(super-FEC) concatenated codes are proposed after the development trend of long-haul optical transmission systems and the defects of the existing FEC codes have been analyzed. The performance simulation of the Reed-Solomon(RS)+Bose-Chaudhuri-Hocguenghem(BCH) inner-outer serial concatenated code is implemented and the conceptions of encoding/decoding the parallel-concatenated code are presented. Furthermore,the simulation results for the RS(255,239)+RS(255,239) code and the RS(255,239)+RS(255,223) code show that the two consecutive concatenated codes are a superior coding scheme with such advantages as the better error correction,moderate redundancy and easy realization compared to the classic RS(255,239) code and other codes,and their signal to noise ratio gains are respectively 2～3 dB more than that of the RS(255,239)code at the bit error rate of 1×10-13. Finally,the frame structure of the novel consecutive concatenated code is arranged to lay a firm foundation in designing its hardware.
Cattania, C.; Khalid, F.
2016-09-01
The estimation of space and time-dependent earthquake probabilities, including aftershock sequences, has received increased attention in recent years, and Operational Earthquake Forecasting systems are currently being implemented in various countries. Physics based earthquake forecasting models compute time dependent earthquake rates based on Coulomb stress changes, coupled with seismicity evolution laws derived from rate-state friction. While early implementations of such models typically performed poorly compared to statistical models, recent studies indicate that significant performance improvements can be achieved by considering the spatial heterogeneity of the stress field and secondary sources of stress. However, the major drawback of these methods is a rapid increase in computational costs. Here we present a code to calculate seismicity induced by time dependent stress changes. An important feature of the code is the possibility to include aleatoric uncertainties due to the existence of multiple receiver faults and to the finite grid size, as well as epistemic uncertainties due to the choice of input slip model. To compensate for the growth in computational requirements, we have parallelized the code for shared memory systems (using OpenMP) and distributed memory systems (using MPI). Performance tests indicate that these parallelization strategies lead to a significant speedup for problems with different degrees of complexity, ranging from those which can be solved on standard multicore desktop computers, to those requiring a small cluster, to a large simulation that can be run using up to 1500 cores.
OPT13B and OPTIM4 - computer codes for optical model calculations
International Nuclear Information System (INIS)
OPT13B is a computer code in FORTRAN for optical model calculations with automatic search. A summary of different formulae used for computation is given. Numerical methods are discussed. The 'search' technique followed to obtain the set of optical model parameters which produce best fit to experimental data in a least-square sense is also discussed. Different subroutines of the program are briefly described. Input-output specifications are given in detail. A modified version of OPT13B specifications are given in detail. A modified version of OPT13B is OPTIM4. It can be used for optical model calculations where the form factors of different parts of the optical potential are known point by point. A brief description of the modifications is given. (author)
Determination of Solution Accuracy of Numerical Schemes as Part of Code and Calculation Verification
Energy Technology Data Exchange (ETDEWEB)
Blottner, F.G.; Lopez, A.R.
1998-10-01
This investigation is concerned with the accuracy of numerical schemes for solving partial differential equations used in science and engineering simulation codes. Richardson extrapolation methods for steady and unsteady problems with structured meshes are presented as part of the verification procedure to determine code and calculation accuracy. The local truncation error de- termination of a numerical difference scheme is shown to be a significant component of the veri- fication procedure as it determines the consistency of the numerical scheme, the order of the numerical scheme, and the restrictions on the mesh variation with a non-uniform mesh. Genera- tion of a series of co-located, refined meshes with the appropriate variation of mesh cell size is in- vestigated and is another important component of the verification procedure. The importance of mesh refinement studies is shown to be more significant than just a procedure to determine solu- tion accuracy. It is suggested that mesh refinement techniques can be developed to determine con- sistency of numerical schemes and to determine if governing equations are well posed. The present investigation provides further insight into the conditions and procedures required to effec- tively use Richardson extrapolation with mesh refinement studies to achieve confidence that sim- ulation codes are producing accurate numerical solutions.
Opacity calculation for target physics using the ABAKO/RAPCAL code
Mínguez, E.; Florido, R.; Rodríguez, R.; Gil, J. M.; Rubiano, J. G.; Mendoza, M. A.; Suárez, D.; Martel, P.
2010-01-01
Radiative properties of hot dense plasmas remain a subject of current interest since they play an important role in inertial confinement fusion (ICF) research, as well as in studies on stellar physics. In particular, the understanding of ICF plasmas requires emissivities and opacities for both hydro-simulations and diagnostics. Nevertheless, the accurate calculation of these properties is still an open question and continuous efforts are being made to develop new models and numerical codes that can facilitate the evaluation of such properties. In this work the set of atomic models ABAKO/RAPCAL is presented, as well as a series of results for carbon and aluminum to show its capability for modeling the population kinetics of plasmas in both LTE and NLTE regimes. Also, the spectroscopic diagnostics of a laser-produced aluminum plasma using ABAKO/RAPCAL is discussed. Additionally, as an interesting application of these codes, fitting analytical formulas for Rosseland and Planck mean opacities for carbon plasmas are reported. These formulas are useful as input data in hydrodynamic simulation of targets where the computation task is so hard that in line computation with sophisticated opacity codes is prohibitive.
SMARTIES: User-friendly codes for fast and accurate calculations of light scattering by spheroids
Somerville, W. R. C.; Auguié, B.; Le Ru, E. C.
2016-05-01
We provide a detailed user guide for SMARTIES, a suite of MATLAB codes for the calculation of the optical properties of oblate and prolate spheroidal particles, with comparable capabilities and ease-of-use as Mie theory for spheres. SMARTIES is a MATLAB implementation of an improved T-matrix algorithm for the theoretical modelling of electromagnetic scattering by particles of spheroidal shape. The theory behind the improvements in numerical accuracy and convergence is briefly summarized, with reference to the original publications. Instructions of use, and a detailed description of the code structure, its range of applicability, as well as guidelines for further developments by advanced users are discussed in separate sections of this user guide. The code may be useful to researchers seeking a fast, accurate and reliable tool to simulate the near-field and far-field optical properties of elongated particles, but will also appeal to other developers of light-scattering software seeking a reliable benchmark for non-spherical particles with a challenging aspect ratio and/or refractive index contrast.
Study on the entire system of maintenance codes and standards
International Nuclear Information System (INIS)
In this study, a structure of code and standard system for plant maintenance is discussed along a process of maintenance activities. As a result of consideration, it was concluded as follows. (1) It is assumed that the entire system of maintenance codes and standards consists of four standards, that is, standards regarding maintenance planning, maintenance implementation, evaluation of inspection/maintenance results and corrective measures. (2) The maintenance guidelines and fitness-for-service codes discussed already so far occupies a position in the entire system of maintenance codes and standards. (3) Maintenance codes and standards, which have higher priority, should be developed. (author)
A GPU-based Monte Carlo dose calculation code for photon transport in a voxel phantom
International Nuclear Information System (INIS)
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)
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)
Deep-penetration calculation for the ISIS target station shielding using the MARS Monte Carlo code
Nunomiya, T; Nakamura, T; Nakao, N
2002-01-01
A calculation of neutron penetration through a thick shield was performed with a three-dimensional multi-layer technique using the MARS14(02) Monte Carlo code to compare with the experimental shielding data in 1998 at the ISIS spallation neutron source facility. In this calculation, secondary particles from a tantalum target bombarded by 800-MeV protons were transmitted through a bulk shield of approximately 3-m-thick iron and 1-m-thick concrete. To accomplish this deep-penetration calculation with good statistics, the following three techniques were used in this study. First, the geometry of the bulk shield was three-dimensionally divided into several layers of about 50-cm thickness, and a step-by-step calculation was carried out to multiply the number of penetrated particles at the boundaries between the layers. Second, the source particles in the layers were divided into two parts to maintain the statistical balance on the spatial-flux distribution. Third, only high-energy particles above 20 MeV were trans...
Three-dimensional whole core transport calculation method and performance of the DeCART code
International Nuclear Information System (INIS)
The three-dimensional (3D) transport calculation method implemented in a whole core neutron transport code DeCART is presented and its performance is examined in terms of solution accuracy and execution speed. The 3D flux calculation in DeCART is based on a transverse-integration method in which the radial and axial dependencies are handled separately. The radial dependence is resolved by the elaborated two-dimensional method of characteristics (MOC) whereas the axial dependence is dealt with the simple one-dimensional diffusion model. The global balance of the 3D flux distribution is incorporated by the coarse mesh finite difference (CMFD) formulation. It is shown that the CMFD formulation enables the approximate three-dimensional transport calculation through the transverse-integration, and furthermore it is very effective in achieving rapid convergence. The accuracy of the approximate 3D whole-core transport calculation method is proved by analyzing rodded variations of the C5G7 MOX heterogeneous core benchmark problem for which Monte Carlo solutions are generated as the reference
International Nuclear Information System (INIS)
The present report describes a computer code DEEP which calculates the organ dose equivalents and the effective dose equivalent for external photon exposure by the Monte Carlo method. MORSE-CG, Monte Carlo radiation transport code, is incorporated into the DEEP code to simulate photon transport phenomena in and around a human body. The code treats an anthropomorphic phantom represented by mathematical formulae and user has a choice for the phantom sex: male, female and unisex. The phantom can wear personal dosimeters on it and user can specify their location and dimension. This document includes instruction and sample problem for the code as well as the general description of dose calculation, human phantom and computer code. (author)
A New Arithmetic Coding System Combining Source Channel Coding and MAP Decoding
Institute of Scientific and Technical Information of China (English)
PANG Yu-ye; SUN Jun; WANG Jia
2007-01-01
A new arithmetic coding system combining source channel coding and maximum a posteriori decoding were proposed.It combines source coding and error correction tasks into one unified process by introducing an adaptive forbidden symbol.The proposed system achieves fixed length code words by adaptively adjusting the probability of the forbidden symbol and adding tail digits of variable length.The corresponding improved MAP decoding metric was derived.The proposed system can improve the performance.Simulations were performed on AWGN channels with various noise levels by using both hard and soft decision with BPSK modulation.The results show its performance is slightly better than that of our adaptive arithmetic error correcting coding system using a forbidden symbol.
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…
Deductive Glue Code Synthesis for Embedded Software Systems Based on Code Patterns
Liu, Jian; Fu, Jicheng; Zhang, Yansheng; Bastani, Farokh; Yen, I-Ling; Tai, Ann; Chau, Savio N.
2006-01-01
Automated code synthesis is a constructive process that can be used to generate programs from specifications. It can, thus, greatly reduce the software development cost and time. The use of formal code synthesis approach for software generation further increases the dependability of the system. Though code synthesis has many potential benefits, the synthesis techniques are still limited. Meanwhile, components are widely used in embedded system development. Applying code synthesis to component based software development (CBSD) process can greatly enhance the capability of code synthesis while reducing the component composition efforts. In this paper, we discuss the issues and techniques for applying deductive code synthesis techniques to CBSD. For deductive synthesis in CBSD, a rule base is the key for inferring appropriate component composition. We use the code patterns to guide the development of rules. Code patterns have been proposed to capture the typical usages of the components. Several general composition operations have been identified to facilitate systematic composition. We present the technique for rule development and automated generation of new patterns from existing code patterns. A case study of using this method in building a real-time control system is also presented.
Code Based Analysis for Object-Oriented Systems
Institute of Scientific and Technical Information of China (English)
Swapan Bhattacharya; Ananya Kanjilal
2006-01-01
The basic features of object-oriented software makes it difficult to apply traditional testing methods in objectoriented systems. Control Flow Graph (CFG) is a well-known model used for identification of independent paths in procedural software. This paper highlights the problem of constructing CFG in object-oriented systems and proposes a new model named Extended Control Flow Graph (ECFG) for code based analysis of Object-Oriented (OO) software. ECFG is a layered CFG where nodes refer to methods rather than statements. A new metrics - Extended Cyclomatic Complexity (E-CC) is developed which is analogous to McCabe's Cyclomatic Complexity (CC) and refers to the number of independent execution paths within the OO software. The different ways in which CFG's of individual methods are connected in an ECFG are presented and formulas for E-CC for these different cases are proposed. Finally we have considered an example in Java and based on its ECFG, applied these cases to arrive at the E-CC of the total system as well as proposed a methodology for calculating the basis set, i.e., the set of independent paths for the OO system that will help in creation of test cases for code testing.
International Nuclear Information System (INIS)
The computer code GORGON, which calculates the energy deposition and slowing down of ions in cold materials and hot plasmas is described, and analyzed in this report. This code is in a state of continuous development but an intermediate stage has been reached where it is considered useful to document the 'state of the art' at the present time. The GORGON code is an improved version of a code developed by Zinamon et al. as part of a more complex program system for studying the hydrodynamic motion of plane metal targets irradiated by intense beams of protons. The improvements made in the code were necessary to improve its usefulness for problems related to the design and burn of heavy ion beam driven inertial confinement fusion targets. (orig./GG)
MELCOR Accident Consequence Code System (MACCS)
Energy Technology Data Exchange (ETDEWEB)
Jow, H.N.; Sprung, J.L.; Ritchie, L.T. (Sandia National Labs., Albuquerque, NM (USA)); Rollstin, J.A. (GRAM, Inc., Albuquerque, NM (USA)); Chanin, D.I. (Technadyne Engineering Consultants, Inc., Albuquerque, NM (USA))
1990-02-01
This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management. 59 refs., 14 figs., 15 tabs.
MELCOR Accident Consequence Code System (MACCS)
International Nuclear Information System (INIS)
This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management. 59 refs., 14 figs., 15 tabs
Monte Carlo Code System Development for Liquid Metal Reactor
Energy Technology Data Exchange (ETDEWEB)
Kim, Chang Hyo; Shim, Hyung Jin; Han, Beom Seok; Park, Ho Jin; Park, Dong Gyu [Seoul National University, Seoul (Korea, Republic of)
2007-03-15
We have implemented the composition cell class and the use cell to MCCARD for hierarchy input processing. For the inputs of KALlMER-600 core consisted of 336 assemblies, we require the geometric data of 91,056 pin cells. Using hierarchy input processing, it was observed that the system geometries are correctly handled with the geometric data of total 611 cells; 2 cells for fuel rods, 2 cells for guide holes, 271 translation cells for rods, and 336 translation cells for assemblies. We have developed monte carlo decay-chain models based on decay chain model of REBUS code for liquid metal reactor analysis. Using developed decay-chain models, the depletion analysis calculations have performed for the homogeneous and heterogeneous model of KALlMER-600. The k-effective for the depletion analysis agrees well with that of REBUS code. and the developed decay chain models shows more efficient performance for time and memories, as compared with the existing decay chain model The chi-square criterion has been developed to diagnose the temperature convergence for the MC TjH feedback calculations. From the application results to the KALlMER pin and fuel assembly problem, it is observed that the new criterion works well Wc have applied the high efficiency variance reduction technique by splitting Russian roulette to estimate the PPPF of the KALIMER core at BOC. The PPPF of KALlMER core at BOC is 1.235({+-}0.008). The developed technique shows four time faster calculation, as compared with the existin2 calculation Subject Keywords Monte Carlo
Energy Technology Data Exchange (ETDEWEB)
Dekens, O.; Lefebvre, J.C.; Rohart, M. [Electricite de France (EDF), 69 -Villeurbanne (France); Chiron, M. [CEA Centre d`Etudes de Saclay, 91 -Gif-sur-Yvette (France). Direction des Reacteurs Nucleaires; Wouters, R. de [TRACTEBEL, Brussels (Belgium)
1997-10-01
During the accident of TM12, the signal delivered by source detectors was exceptionally high. This phenomenon was found out to be due to the water inventory in the primary system. Thus, in their research activity, Electricite de France (EdF) and Commissariat a l`Energie Atomique (CEA) have jointly launched a programme, whose aim was to determine to what extent the response of ex-vessel neutron detectors are representative of reactor water level (or sources positions) in a French 900 MWe PWR. In this framework, both partners developed the methods needed for each step of the calculation chain. Finally, a simulation of a LOCA indicates that the loss of coolant can be detected by existing monitoring system, and could be more efficiently found by changing the position of the source range detectors. (authors). 11 refs.
Implantation of Magint-Maggraf code in the Microvax-3600 system
International Nuclear Information System (INIS)
An auxiliary code, named MAGGRAF, was developed and implemented on the MicroVAX-3600 system of the Plasma Laboratory at the Institute for Space Research, in order to perform the graphical output of the numerical results provided by the magnetic field calculation code MAGINT. In this report we present a brief description of the graphical code, of the parameters which specify the different output options, and of the structure of the data file containing these parameters. Some examples are shown to illustrate the versatility of the code, as well as the quality of the graphs. (author)
Morse Monte Carlo Radiation Transport Code System
Energy Technology Data Exchange (ETDEWEB)
Emmett, M.B.
1975-02-01
The report contains sections containing descriptions of the MORSE and PICTURE codes, input descriptions, sample problems, deviations of the physical equations and explanations of the various error messages. The MORSE code is a multipurpose neutron and gamma-ray transport Monte Carlo code. Time dependence for both shielding and criticality problems is provided. General three-dimensional geometry may be used with an albedo option available at any material surface. The PICTURE code provide aid in preparing correct input data for the combinatorial geometry package CG. It provides a printed view of arbitrary two-dimensional slices through the geometry. By inspecting these pictures one may determine if the geometry specified by the input cards is indeed the desired geometry. 23 refs. (WRF)
MARS code manual volume I: code structure, system models, and solution methods
International Nuclear Information System (INIS)
Korea Advanced Energy Research Institute (KAERI) conceived and started the development of MARS code with the main objective of producing a state-of-the-art realistic thermal hydraulic systems analysis code with multi-dimensional analysis capability. MARS achieves this objective by very tightly integrating the one dimensional RELAP5/MOD3 with the multi-dimensional COBRA-TF codes. The method of integration of the two codes is based on the dynamic link library techniques, and the system pressure equation matrices of both codes are implicitly integrated and solved simultaneously. In addition, the Equation-Of-State (EOS) for the light water was unified by replacing the EOS of COBRA-TF by that of the RELAP5. This theory manual provides a complete list of overall information of code structure and major function of MARS including code architecture, hydrodynamic model, heat structure, trip / control system and point reactor kinetics model. Therefore, this report would be very useful for the code users. The overall structure of the manual is modeled on the structure of the RELAP5 and as such the layout of the manual is very similar to that of the RELAP. This similitude to RELAP5 input is intentional as this input scheme will allow minimum modification between the inputs of RELAP5 and MARS3.1. MARS3.1 development team would like to express its appreciation to the RELAP5 Development Team and the USNRC for making this manual possible
International Nuclear Information System (INIS)
A working party was organized for two years (2001-2002) on common reactor physics code systems under the Research Committee on Reactor Physics of JAERI. This final report is compilation of activity of the working party on common reactor physics code systems during two years. Objectives of the working party is to clarify basic concept of common reactor physics code systems to improve convenience of reactor physics code systems for reactor physics researchers in Japan on their various field of research and development activities. We have held four meetings during 2 years, investigated status of reactor physics code systems and innovative software technologies, and discussed basic concept of common reactor physics code systems. (author)
AVS 3D Video Coding Technology and System
Institute of Scientific and Technical Information of China (English)
Siwei Ma; Shiqi Wang; Wen Gao
2012-01-01
Following the success of the audio video standard （AVS） for 2D video coding, in 2008, the China AVS workgroup started developing 3D video （3DV） coding techniques. In this paper, we discuss the background, technical features, and applications of AVS 3DV coding technology. We introduce two core techniques used in AVS 3DV coding： inter-view prediction and enhanced stereo packing coding. We elaborate on these techniques, which are used in the AVS real-time 3DV encoder. An application of the AVS 3DV coding system is presented to show the great practical value of this system. Simulation results show that the advanced techniques used in AVS 3DV coding provide remarkable coding gain compared with techniques used in a simulcast scheme.
Energy Technology Data Exchange (ETDEWEB)
Hugtenburg, Richard P., E-mail: r.p.hugtenburg@swansea.ac.u [School of Medicine, Swansea University, Swansea SA2 8PP (United Kingdom); Department of Medical Physics and Clinical Engineering, Abertawe Bro Morgannwg University, LHB, Swansea SA2 8QA (United Kingdom); Adegunloye, A.S.; Bradley, David A. [Department of Physics, Surrey University, Guildford (United Kingdom)
2010-07-21
Microbeam radiation therapy (MRT) is currently being considered for the treatment of glioblastoma multiforme. A high degree of dosimetric accuracy (around 5%) is known to be required for a successful outcome in conventional radiation therapy, Modelling of MRT beams, measurements and treatments have been performed with Monte Carlo methods using the code EGS5, which features improved physics models for low energy scattering processes including linear polarisation. Polarisation of the X-ray source leads to distortions in beam profiles that exceed the usual clinical tolerances. Changes in the energy spectrum also effect the response of many dosimetry systems. Anatomical (CT) data has been used in the dose calculations and the manipulation of dose data with the open-source software treatment planning system, PlanUNC, is demonstrated, in order that the therapeutic effects of the different components, e.g. the microbeam and scattered photons, can examined separately in relation to relevant anatomy.
MELCOR Accident Consequence Code System (MACCS)
International Nuclear Information System (INIS)
This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previous CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. This document, Volume 1, the Users's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems
MELCOR Accident Consequence Code System (MACCS)
Energy Technology Data Exchange (ETDEWEB)
Chanin, D.I. (Technadyne Engineering Consultants, Inc., Albuquerque, NM (USA)); Sprung, J.L.; Ritchie, L.T.; Jow, Hong-Nian (Sandia National Labs., Albuquerque, NM (USA))
1990-02-01
This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previous CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. This document, Volume 1, the Users's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems.
Treatment Registration and Nuclide Decay Calculation System
Institute of Scientific and Technical Information of China (English)
WU Jian-guo; XU Bo; CHEN Zhi-jun; ZHOU Ai-qing; WANG Xue-qin; ZHANG Bin; MA Tao; SHEN Jun-jin; LIU Jie; JIN Hai-xia
2008-01-01
Objective:To design a software to do the complicated and multiple calcula-tions automatically in routine internal radionuclide irradiation therapy to avoid mistakes and shorten patients waiting times. Methods:The software is designed on the Microsoft Windows XP operating system. Visual Basic 5.0 and Microsoft Access 2000 are used re-spectively as the programming language and database system here. The data and DBGrid controls and VB data window guide of Visual Basic were used to control access to and Ac-cess database. Results: Not only can the radioactivity of any radionuclide be calculated, but also the administered total iodine dose of therapy for hyperthyroidism or thyroid cancer and the total administered 153 Sm-EDTMP solutions for remedy of bone metastasis of malig-nant tumor can be ciphered out. Conclusion: The work becomes easier, faster, more cor-rect and interesting when the software can make the complicated and multiple calculations automatically. Patients' information, diagnosis and treatment can be recorded for further study.
Dose and shielding calculation of galactic cosmic ray using FLUKA Mont Carlo code
Energy Technology Data Exchange (ETDEWEB)
Jalali, Hamide B. [Physics Department, University of Qom, Qom (Iran); Raisali, Golamreza; Babazade, Alireza [Radiation Applications Research School, Nuclear Science and Technology Research Institute, Atomic Energy Organization of Iran, Tehran (Iran); Feghhi, Amirhosein [Physics and Nuclear Engineering Department, Amirkabir University, Tehran (Iran)
2009-07-01
Astronauts' exposure to space radiation is a limiting factor for long-term missions. Therefore shielding is a critical issue in space mission success. In this work the FLUKA Monte Carlo code has been coupled with simple models of the spacecraft and equivalent phantom to calculate skin averaged doses due to exposure to Galactic Cosmic Rays (GCR) beyond various thicknesses of aluminium and polyethylene shields. Simulations have been performed for the most abundant elements including H, He, C and Fe ions. The spectra of these ions have been taken from Badhwar-O'Neill's model, and LET distribution of the ions and electrons calculated using SRIM and ESTAR computer programs, respectively. It has been observed that GCR absorbed dose behind the shields remained approximately constant with increasing shield thicknesses, but dose equivalent shows a slight decrease. It is also found that although polyethylene is a more effective GCR shield than aluminum as indicated in the results of similar investigations, but the practical thicknesses of polyethylene are still insufficient to shield high energy GCR ions encountered in long-term space missions.
Study of adaptive modulation and LDPC coding in multicarrier systems
Institute of Scientific and Technical Information of China (English)
无
2008-01-01
An adaptive modulation (AM) algorithm is proposed and the application of the adapting algorithm together with low-density parity-check (LDPC) codes in multicarrier systems is investigated.The AM algorithm is based on minimizing the average bit error rate (BER) of systems,the combination of AM algorithm and LDPC codes with different code rates (half and three-fourths) are studied.The proposed AM algorithm with that of Fischer et al is compared.Simulation results show that the performance of the proposed AM algorithm is better than that of the Fischer's algorithm.The results also show that application of the proposed AM algorithm together with LDPC codes can greatly improve the performance of multicarrier systems.Results also show that the performance of the proposed algorithm is degraded with an increase in code rate when code length is the same.
Lifeline system network reliability calculation based on GIS and FTA
Institute of Scientific and Technical Information of China (English)
TANG Ai-ping; OU Jin-ping; LU Qin-nian; ZHANG Ke-xu
2006-01-01
Lifelines, such as pipeline, transportation, communication, electric transmission and medical rescue systems, are complicated networks that always distribute spatially over large geological and geographic units.The quantification of their reliability under an earthquake occurrence should be highly regarded, because the performance of these systems during a destructive earthquake is vital in order to estimate direct and indirect economic losses from lifeline failures, and is also related to laying out a rescue plan. The research in this paper aims to develop a new earthquake reliability calculation methodology for lifeline systems. The methodology of the network reliability for lifeline systems is based on fault tree analysis (FTA) and geological information system(GIS). The interactions existing in a lifeline system are considered herein. The lifeline systems are idealized as equivalent networks, consisting of nodes and links, and are described by network analysis in GIS. Firstly, the node is divided into two types: simple node and complicated node, where the reliability of the complicated node is calculated by FTA and interaction is regarded as one factor to affect performance of the nodes. The reliability of simple node and link is evaluated by code. Then, the reliability of the entire network is assessed based on GIS and FTA. Lastly, an illustration is given to show the methodology.
Digital system detects binary code patterns containing errors
Muller, R. M.; Tharpe, H. M., Jr.
1966-01-01
System of square loop magnetic cores associated with code input registers to react to input code patterns by reference to a group of control cores in such a manner that errors are canceled and patterns containing errors are accepted for amplification and processing. This technique improves reception capabilities in PCM telemetry systems.
RELAP5/MOD3 code manual: Code structure, system models, and solution methods. Volume 1
International Nuclear Information System (INIS)
The RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents, and operational transients, such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling, approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. RELAP5/MOD3 code documentation is divided into seven volumes: Volume I provides modeling theory and associated numerical schemes
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.
Neutronic calculation to the TRIGA Ipr-R1 reactor using the WIMSD4 and CITATION codes
International Nuclear Information System (INIS)
The WIMSD4 and CITATION codes are used to calculate neutronic parameters of a TRIGA reactor. The results are compared with experimental values. Five configurations are analysed and the excess reactivity worth, the fuel temperature reactivity coefficient, the control reactivity worth, safety and regulation rod of the TRIGA IPR-R1 reactor are calculated. The idea is to obtain the systematic error for k∞ for this methodology comparing the calculated and the experimental results
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.
Institute of Scientific and Technical Information of China (English)
CHENG Yuxin; ZHANG Lei; YI Na; XIANG Haige
2007-01-01
Bit-interleaved coded modulation (BICM) is suitable to bandwidth-efficient communication systems. Hybrid automatic repeat request (HARQ) can provide more reliability to high-speed wireless data transmission. A new path weight complementary convolutional (PWCC) code used in the type-ll BICM-HARQ system is proposed. The PWCC code is composed of the original code and the complimentary code. The path in trellis with large hamming weight of the complimentary code is designed to compensate for the path in trellis with small hamming weight of the original code. Hence, both of the original code and the complimentary code can achieve the performance of the good code criterion of corresponding code rate. The throughput efficiency of the BICM-HARQ system wit PWCC code is higher than repeat code system, a little higher than puncture code system in low signal-to-noise ratio (SNR) values and much higher than puncture code system, the same as repeat code system in high SNR values. These results are confirmed by the simulation.
Partial iterated function system-based fractal image coding
Wang, Zhou; Yu, Ying Lin
1996-06-01
A recent trend in computer graphics and image processing has been to use iterated function system (IFS) to generate and describe images. Barnsley et al. presented the conception of fractal image compression and Jacquin was the first to propose a fully automatic gray scale still image coding algorithm. This paper introduces a generalization of basic IFS, leading to a conception of partial iterated function system (PIFS). A PIFS operator is contractive under certain conditions and when it is applied to generate an image, only part of it is actually iteratedly applied. PIFS provides us a flexible way to combine fractal coding with other image coding techniques and many specific algorithms can be derived from it. On the basis of PIFS, we implement a partial fractal block coding (PFBC) algorithm and compare it with basic IFS based fractal block coding algorithm. Experimental results show that coding efficiency is improved and computation time is reduced while image fidelity does not degrade very much.
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.
Programme Code for Projecting of WDM Fiber Optic Sensor Systems
Probstner, R.; J. Turan
1993-01-01
Wavelength division multiplex (WDM) offers a potentially powerful technique for use within optical fibre sensor systems. The paper deals with short description of methodology and a programme code for WDM fiber optic sensor system projecting with use of CAD.
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.
Energy Technology Data Exchange (ETDEWEB)
Di Maio, Francesco, E-mail: francesco.dimaio@polimi.it [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Nicola, Giancarlo [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Zio, Enrico [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Chair on System Science and Energetic Challenge Fondation EDF, Ecole Centrale Paris and Supelec, Paris (France); Yu, Yu [School of Nuclear Science and Engineering, North China Electric Power University, 102206 Beijing (China)
2015-08-15
Highlights: • Uncertainties of TH codes affect the system failure probability quantification. • We present Finite Mixture Models (FMMs) for sensitivity analysis of TH codes. • FMMs approximate the pdf of the output of a TH code with a limited number of simulations. • The approach is tested on a Passive Containment Cooling System of an AP1000 reactor. • The novel approach overcomes the results of a standard variance decomposition method. - Abstract: For safety analysis of Nuclear Power Plants (NPPs), Best Estimate (BE) Thermal Hydraulic (TH) codes are used to predict system response in normal and accidental conditions. The assessment of the uncertainties of TH codes is a critical issue for system failure probability quantification. In this paper, we consider passive safety systems of advanced NPPs and present a novel approach of Sensitivity Analysis (SA). The approach is based on Finite Mixture Models (FMMs) to approximate the probability density function (i.e., the uncertainty) of the output of the passive safety system TH code with a limited number of simulations. We propose a novel Sensitivity Analysis (SA) method for keeping the computational cost low: an Expectation Maximization (EM) algorithm is used to calculate the saliency of the TH code input variables for identifying those that most affect the system functional failure. The novel approach is compared with a standard variance decomposition method on a case study considering a Passive Containment Cooling System (PCCS) of an Advanced Pressurized reactor AP1000.
Experimental transport analysis code system in JT-60
International Nuclear Information System (INIS)
Transport analysis codes have been developed in order to study confinement properties related to particle and energy balance in ohmically and neutral beam heated plasmas of JT-60. The analysis procedure is divided into three steps as follows: 1) LOOK ; The shape of the plasma boundary is identified with a fast boundary identification code of FBI by using magnetic data, and flux surfaces are calculated with a MHD equilibrium code of SELENE. The diagnostic data are mapped to flux surfaces for neutral beam heating calculation and/or for radial transport analysis. 2) OFMC ; On the basis of transformed data, an orbit following Monte Carlo code of OFMC calculates both profiles of power deposition and particle source of neutral beam injected into a plasma. 3) SCOOP ; In the last stage, a one dimensional transport code of SCOOP solves particle and energy balance for electron and ion, in order to evaluate transport coefficients as well as global parameters such as energy confinement time and the stored energy. The analysis results are provided to a data bank of DARTS that is used to find an overview of important consideration on confinement with a regression analysis code of RAC. (author)
System Measures Errors Between Time-Code Signals
Cree, David; Venkatesh, C. N.
1993-01-01
System measures timing errors between signals produced by three asynchronous time-code generators. Errors between 1-second clock pulses resolved to 2 microseconds. Basic principle of computation of timing errors as follows: central processing unit in microcontroller constantly monitors time data received from time-code generators for changes in 1-second time-code intervals. In response to any such change, microprocessor buffers count of 16-bit internal timer.
A computer code to calculate the fast induced signals by electron swarms in gases
Energy Technology Data Exchange (ETDEWEB)
Tobias, Carmen C.B. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Mangiarotti, Alessio [Universidade de Coimbra (Portugal). Dept. de Fisica. Lab. de Instrumentacao e Fisica Experimental de Particulas
2010-07-01
Full text: The study of electron transport parameters (i.e. drift velocity, diffusion coefficients and first Townsend coefficient) in gases is very important in several areas of applied nuclear science. For example, they are a relevant input to the design of particle detector employing micro-structures (MSGC's, micromegas, GEM's) and RPC's (resistive plate chambers). Moreover, if the data are accurate and complete enough, they can be used to derive a set of electron impact cross-sections with their energy dependence, that are a key ingredient in micro-dosimetry calculations. Despite the fundamental need of such data and the long age of the field, the gases of possible interest are so many and the effort of obtaining good quality data so time demanding, that an important contribution can still be made. As an example, electrons drift velocity at moderate field strengths (up to 50 Td) in pure Isobutane (a tissue equivalent gas) has been measured only recently by the IPEN-LIP collaboration using a dedicated setup. The transport parameters are derived from the recorded electric pulse induced by a swarm started with a pulsed laser shining on the cathode. To aid the data analysis, a special code has been developed to calculate the induced pulse by solving the electrons continuity equation including growth, drift and diffusion. A realistic profile of the initial laser beam is taken into account as well as the boundary conditions at the cathode and anode. The approach is either semi-analytic, based on the expression derived by P. H. Purdie and J. Fletcher, or fully numerical, using a finite difference scheme improved over the one introduced by J. de Urquijo et al. The agreement between the two will be demonstrated under typical conditions for the mentioned experimental setup. A brief discussion on the stability of the finite difference scheme will be given. The new finite difference scheme allows a detailed investigation of the importance of back diffusion to
Calculation of Sodium Fire Test-I (Run-E6) using sodium combustion analysis code ASSCOPS version 2.0
Energy Technology Data Exchange (ETDEWEB)
Nakagiri, Toshio; Ohno, Shuji; Miyake, Osamu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center
1997-11-01
The calculation of Sodium Fire Test-I (Run-E6) was performed using the ASSCOPS (Analysis of Simultaneous Sodium Combustions in Pool and Spray) code version 2.0 in order to determine the parameters used in the code for the calculations of sodium combustion behavior of small or medium scale sodium leak, and to validate the applicability of the code. The parameters used in the code were determined and the validation of the code was confirmed because calculated temperatures, calculated oxygen concentration and other calculated values almost agreed with the test results. (author)
Dose calculation accuracy of lung planning with a commercial IMRT treatment planning system.
McDermott, Patrick N; He, Tongming; DeYoung, A
2003-01-01
The dose calculation accuracy of a commercial pencil beam IMRT planning system is evaluated by comparison with Monte Carlo calculations and measurements in an anthropomorphic phantom. The target volume is in the right lung and mediastinum and thus significant tissue inhomogeneities are present. The Monte Carlo code is an adaptation of the MCNP code and the measurements were made with TLD and film. Both the Monte Carlo code and the measurements show very good agreement with the treatment planning system except in regions where the dose is high and the electron density is low. In these regions the commercial system shows doses up to 10% higher than Monte Carlo and film. The average calculated dose for the CTV is 5% higher with the commercial system as compared to Monte Carlo. PMID:14604424
Study of nuclear computer code maintenance and management system
International Nuclear Information System (INIS)
Software maintenance is one of the most important problems since late 1970's.We wish to develop a nuclear computer code system to maintenance and manage KAERI's nuclear software. As a part of this system, we have developed three code management programs for use on CYBER and PC systems. They are used in systematic management of computer code in KAERI. The first program is embodied on the CYBER system to rapidly provide information on nuclear codes to the users. The second and the third programs were embodied on the PC system for the code manager and for the management of data in korean language, respectively. In the requirement analysis, we defined each code, magnetic tape, manual and abstract information data. In the conceptual design, we designed retrieval, update, and output functions. In the implementation design, we described the technical considerations of database programs, utilities, and directions for the use of databases. As a result of this research, we compiled the status of nuclear computer codes which belonged KAERI until September, 1988. Thus, by using these three database programs, we could provide the nuclear computer code information to the users more rapidly. (Author)
International Nuclear Information System (INIS)
A short description of the TOPRA-s computer code is presented. The code is developed to calculate the thermophysical cross-section characteristics of the WWER fuel rods: fuel temperature distributions and fuel-to-cladding gap conductance. The TOPRA-s input does not require the fuel rod irradiation pre-history (time dependent distributions of linear power, fast neutron flux and coolant temperature along the rod). The required input consists of the considered cross-section data (coolant temperature, burnup, linear power) and the overall fuel rod data (burnup and linear power). TOPRA-s is included into the KASKAD code package. Some results of the TOPRA-s code validation using the SOFIT-1 and IFA-503.1 experimental data, are shown. A short description of the TRANSURANUS code for thermal and mechanical predictions of the LWR fuel rod behavior at various irradiation conditions and its version for WWER reactors, are presented. (Authors)
International Nuclear Information System (INIS)
A computer code OSCAR, operated on a main frame computer was developed for the calculation of the yield of radioisotopes produced by charged-particle induced nuclear reactions. The excitation functions required for calculating the yield were evaluated by means of an empirical rule which we developed on the basis of a systematics derived from a number of experimental data reported in the literature. The rule is valid for light ion (Z ≤ 2)-induced reactions followed by neutron emission processes. Other excitation functions are also obtainable from the data file in OSCAR. In addition, the code possesses functions useful for the calculation of the stopping power and range. The energy loss and the distribution of recoil products in stacked targets are also provided as options. The formalism, structure, and direction for the usage of the code are described together with the explanation of the functions of some routines. (author)
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.
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.
International Nuclear Information System (INIS)
A flexible and interactive code, NEPTUN, has been written in FORTRAN IV for the PDP-10 computer to assess the impact on man of radionuclides in aquatic food chains. NEPTUN is based on an equilibrium model of the linear-chain type, and calculates aquatic food concentrations and doses to man. A decay term is included for the holdup time of the various food types. A total of seven food types can be selected, which include drinking water, freshwater and salt-water plants, inverebrates and fish. Thirty different diets can be implemented and five different dose factor files can be chosen. These include dose conversion factors for infants and adults based on ICRP 2 and ICRP 26 methodologies. All dose factors involve a dose commitment of 50 years, or equivalently, 50 years of chronic exposure. To date, only stochastic ICRP 26 dose caluclations have been implemented. The basic concentration factor file contains data for 211 different radionuclides; the dose factor files are less comprehensive. However, all files can be readily expanded. The output includes tables of concentrations and doses for individual radionuclides, as well as summaries for groups of radionuclides. Existing aquatic food chain models and the sources of currently-used generic concentration factors are briefly reviewed, and dose factors based on ICRP 2 and ICRP 26 methodologies are contrasted. (auth)
Wall touching kink mode calculations with the M3D code
Breslau, J. A.
2014-10-01
In recent years there have been a number of results published concerning the transient vessel currents and forces occurring during a tokamak VDE, as predicted by simulations with the nonlinear MHD code M3D. The nature of the simulations is such that these currents and forces occur at the boundary of the computational domain, making the proper choice of boundary conditions critical to the reliability of the results. The M3D boundary condition includes the prescription that the normal component of the velocity vanish at the wall. It has been argued that this prescription invalidates the calculations because it would seem to rule out the possibility of advection of plasma surface currents into the wall. This claim has been tested by applying M3D to an idealized case - a kink-unstable plasma column - in order to abstract the essential physics from the complications involved in the attempt to model real devices. While comparison of the results is complicated by effects arising from the higher dimensionality and complexity of M3D, we have verified 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.
2D Resistive Magnetohydrodynamics Calculations with an Arbitrary Lagrange Eulerian Code
Rousculp, C. L.; Gianakon, T. A.; Lipnikov, K. N.; Nelson, E. M.
2015-11-01
Single fluid resistive MHD is useful for modeling Z-pinch configurations in cylindrical geometry. One such example is thin walled liners for shock physics or HEDP experiments driven by capacitor banks such as the LANL's PHELIX or Sandia-Z. MHD is also useful for modeling high-explosive-driven flux compression generators (FCGs) and their high-current switches. The resistive MHD in our arbitrary Lagrange Eulerian (ALE) code operates in one and two dimensions in both Cartesian and cylindrical geometry. It is implemented as a time-step split operator, which consists of, ideal MHD connected to the explicit hydro momentum and energy equations and a second order mimetic discretization solver for implicit solution of the magnetic diffusion equation. In a staggered grid scheme, a single-component of cell-centered magnetic flux is conserved in the Lagrangian frame exactly, while magnetic forces are accumulated at the nodes. Total energy is conserved to round off. Total flux is conserved under the ALE relaxation and remap. The diffusion solver consistently computes Ohmic heating. Both Neumann and Dirichlet boundary conditions are available with coupling to external circuit models. Example calculations will be shown.
Criticality calculation in TRIGA MARK II PUSPATI Reactor using Monte Carlo code
International Nuclear Information System (INIS)
A Monte Carlo simulation of the Malaysian nuclear reactor has been performed using MCNP Version 5 code. The purpose of the work is the determination of the multiplication factor (keff) for the TRIGA Mark II research reactor in Malaysia based on Monte Carlo method. This work has been performed to calculate the value of keff for two cases, which are the control rod either fully withdrawn or fully inserted to construct a complete model of the TRIGA Mark II PUSPATI Reactor (RTP). The RTP core was modeled as close as possible to the real core and the results of keff from MCNP5 were obtained when the control fuel rods were fully inserted, the keff value indicates the RTP reactor was in the subcritical condition with a value of 0.98370±0.00054. When the control fuel rods were fully withdrawn the value of keff value indicates the RTP reactor is in the supercritical condition, that is 1.10773±0.00083. (Author)
Arithmetic coding as a non-linear dynamical system
Nagaraj, Nithin; Vaidya, Prabhakar G.; Bhat, Kishor G.
2009-04-01
In order to perform source coding (data compression), we treat messages emitted by independent and identically distributed sources as imprecise measurements (symbolic sequence) of a chaotic, ergodic, Lebesgue measure preserving, non-linear dynamical system known as Generalized Luröth Series (GLS). GLS achieves Shannon's entropy bound and turns out to be a generalization of arithmetic coding, a popular source coding algorithm, used in international compression standards such as JPEG2000 and H.264. We further generalize GLS to piecewise non-linear maps (Skewed-nGLS). We motivate the use of Skewed-nGLS as a framework for joint source coding and encryption.
Optical Code Processing System, Device, and its Application
Directory of Open Access Journals (Sweden)
Naoya Wada
2010-02-01
Full Text Available Recent progress of optical code processing technology_ is explained. Ultra-high speed time domain, spectral domain, hybrid_ domain, and multiple optical code processing deices and systems are shown. As application of these technologies, OCDMA-PON, OPS network, and ultra high-speed optical clock generation will be demonstrated.
Lee, L.-N.
1977-01-01
Concatenated coding systems utilizing a convolutional code as the inner code and a Reed-Solomon code as the outer code are considered. In order to obtain very reliable communications over a very noisy channel with relatively modest coding complexity, it is proposed to concatenate a byte-oriented unit-memory convolutional code with an RS outer code whose symbol size is one byte. It is further proposed to utilize a real-time minimal-byte-error probability decoding algorithm, together with feedback from the outer decoder, in the decoder for the inner convolutional code. The performance of the proposed concatenated coding system is studied, and the improvement over conventional concatenated systems due to each additional feature is isolated.
MORSE Monte Carlo radiation transport code system
International Nuclear Information System (INIS)
For a number of years the MORSE user community has requested additional help in setting up problems using various options. The sample problems distributed with MORSE did not fully demonstrate the capability of the code. At Oak Ridge National Laboratory the code originators had a complete set of sample problems, but funds for documenting and distributing them were never available. Recently the number of requests for listings of input data and results for running some particular option the user was trying to implement has increased to the point where it is not feasible to handle them on an individual basis. Consequently it was decided to package a set of sample problems which illustrates more adequately how to run MORSE. This write-up may be added to Part III of the MORSE report. These sample problems include a combined neutron-gamma case, a neutron only case, a gamma only case, an adjoint case, a fission case, a time-dependent fission case, the collision density case, an XCHEKR run and a PICTUR run
International Nuclear Information System (INIS)
INDRA is a program system for calculating the neutronics and photonics characteristics of fusion reactor blankets. It incorporates a total of 19 different codes and 5 large data libraries. 10 of the codes are available from the code distribution organizations. Some of them, however, have been slightly modified in order to permit a convenient transfer of information from one program module to the next. The remaining 9 programs have been prepared by the authors to complete the system with respect to flexibility and to facilitate the handling of the results. (orig./WBU)
TRANS4: a computer code calculation of solid fuel penetration of a concrete barrier
International Nuclear Information System (INIS)
The computer code, TRANS4, models the melting and penetration of a solid barrier by a solid disc of fuel following a core disruptive accident. This computer code has been used to model fuel debris penetration of basalt, limestone concrete, basaltic concrete, and magnetite concrete. Sensitivity studies were performed to assess the importance of various properties on the rate of penetration. Comparisons were made with results from the GROWS II code
Energy Technology Data Exchange (ETDEWEB)
Gorodkov, S.S.; Kalugin, M.A. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)
2015-09-15
Up to now core calculations with Monte Carlo provided only average cross-sections of mesh cells for further use either in finite difference calculations or as benchmark ones for approximate spectral algorithms. Now MCU code is capable to handle functions, which may be interpreted as average diffusion coefficients. Subsequently the results of finite difference calculations with cells characteristic sets obtained in such a way can be compared with Monte Carlo results as benchmarks, giving reliable information on quality of production code under consideration. As an example of such analysis, the results of mesh calculations with 1-, 2-, 4-, 8- and 12 neutron groups of some model VVER fuel assembly are presented in comparison with the exact Monte Carlo solution. As a second example, an analysis is presented of water gap approximate enlargement between fuel assemblies, allowing VVER core region be covered by regular mesh.
CODING IN THE MAMMALIAN GUSTATORY SYSTEM
Carleton, Alan; Accolla, Riccardo; Simon, Sidney A.
2010-01-01
To understand gustatory physiology and associated dysfunctions it is important to know how stimuli placed in the mouth are encoded both in the periphery and in taste-related brain centres. The identification of distinct taste receptors, together with electrophysiological recordings and behavioural 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. In 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. PMID:20493563
MULTIPLE TRELLIS CODED ORTHOGONAL TRANSMIT SCHEME FOR MULTIPLE ANTENNA SYSTEMS
Institute of Scientific and Technical Information of China (English)
无
2006-01-01
In this paper, a novel multiple trellis coded orthogonal transmit scheme is proposed to exploit transmit diversity in fading channels. In this scheme, a unique vector from a set of orthogonal vectors is assigned to each transmit antenna. Each of the output symbols from the multiple trellis encoder is multiplied with one of these orthogonal vectors and transmitted from corresponding transmit antennas. By correlating with corresponding orthogonal vectors, the receiver separates symbols transmitted from different transmit antennas.This scheme can be adopted in coherent/differential systems with any number of transmit antennas. It is shown that the proposed scheme encompasses the conventional trellis coded unitary space-time modulation based on the optimal cyclic group codes as a special case. We also propose two better designs over the conventional trellis coded unitary space-time modulation. The first design uses 8 Phase Shift Keying (8-PSK) constellations instead of 16 Phase Shift Keying (16-PSK) constellations in the conventional trellis coded unitary space-time modulation. As a result, the product distance of this new design is much larger than that of the conventional trellis coded unitary space-time modulation. The second design introduces constellations with multiple levels of amplitudes into the design of the multiple trellis coded orthogonal transmit scheme. For both designs, simulations show that multiple trellis coded orthogonal transmit schemes can achieve better performance than the conventional trellis coded unitary space-time schemes.
International Nuclear Information System (INIS)
This study made a comparison between some of the major transport codes that employ the Monte Carlo stochastic approach in dosimetric calculations in nuclear medicine. We analyzed in detail the various physical and numerical models used by MCNP5 code in relation with codes like EGS and Penelope. The identification of its potential and limitations for solving microdosimetry problems were highlighted. The condensed history methodology used by MCNP resulted in lower values for energy deposition calculation. This showed a known feature of the condensed stories: its underestimates both the number of collisions along the trajectory of the electron and the number of secondary particles created. The use of transport codes like MCNP and Penelope for micrometer scales received special attention in this work. Class I and class II codes were studied and their main resources were exploited in order to transport electrons, which have particular importance in dosimetry. It is expected that the evaluation of available methodologies mentioned here contribute to a better understanding of the behavior of these codes, especially for this class of problems, common in microdosimetry. (author)
Codes, standards, and PV power systems. A 1996 status report
Energy Technology Data Exchange (ETDEWEB)
Wiles, J
1996-06-01
As photovoltaic (PV) electrical power systems gain increasing acceptance for both off-grid and utility-interactive applications, the safety, durability, and performance of these systems gains in importance. Local and state jurisdictions in many areas of the country require that all electrical power systems be installed in compliance with the requirements of the National Electrical Code{reg_sign} (NEC{reg_sign}). Utilities and governmental agencies are now requiring that PV installations and components also meet a number of Institute of Electrical and Electronic Engineers (IEEE) standards. PV installers are working more closely with licensed electricians and electrical contractors who are familiar with existing local codes and installation practices. PV manufacturers, utilities, balance of systems manufacturers, and standards representatives have come together to address safety and code related issues for future PV installations. This paper addresses why compliance with the accepted codes and standards is needed and how it is being achieved.
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...
Study on New Concatenated Code in WDM Optical Transmission Systems
Institute of Scientific and Technical Information of China (English)
YUAN Jian-guo; JIANG Ze; MAO You-ju; YE Wen-wei
2007-01-01
A new concatenated code of RS(255,239)+BCH(2 040,1 930) code to be suitable for WDM optical transmission systems is proposed.The simulation results show that this new concatenated code,compared with the RS(255,239)+CSOC(k0/n0=6/7,J=8) code in ITU-T G.75.1,has a lower redundancy and better error-correction performance,furthermore,its net coding gain(NCG) is respectively 0.46 dB,0.43 dB more than that of RS(255,239)+CSOC(k0/n0 =6/7,J=8) code and BCH(3 860,3 824)+BCH(2 040,1 930) code in ITU-T G.75.1 at the third iteration for the bit error rate(BER) of 10-12.Therefore,the new super forward error correction(Super-FEC) concatenated code can be better used in ultra long-haul,ultra large-capacity and ultra high-speed WDM optical communication systems.
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.
International Nuclear Information System (INIS)
These scientific days were organised by the 'technical protection' Section of the French Society of Radiation Protection (SFRP) in cooperation with the French society of medical physicists (SFPM), the Swiss Romandie association of radioprotection (ARRAD) and the associated laboratories of radio-physics and dosimetry (LARD). The objective of these days was to review the existing calculation codes used in radiation transport, source estimation and dose management, and to identify some future prospects. This document brings together the available presentations (slides) together with their corresponding abstracts (in French) and dealing with: 1 - Presentation of the conference days (L. De Carlan); 2 - Simulating radionuclide transfers in the environment: what calculation codes and for what? (C. Mourlon); 3 - Contribution of Monte-Carlo calculation to the theoretical foundation analysis of calibration procedures and dosemeters design for radioprotection photon dosimetry (J.M. Bordy); 4 - Use of calculation codes in R and D for the development of a new passive dosemeter for photons and beta radiations (B. Moreno); 5 - Development of a new virtual sources model for the Monte-Carlo prediction of EPID (Electronic Portal Imaging Device) images and implementation in PENELOPE (I. Chabert); 6 - Prediction of high-resolution EPID images for in-vivo dosimetry (D. Patin); 7 - 4D thorax modeling by artificial neural networks (P.E. Leni); 8 - Presentation of the calculation utilities of the book 'Calculation of ionizing radiations generated doses' (Vivier, Lopez, EDP Sciences 2012) (A. Vivier); 9 - RayXpertC: a 3D modeling and Monte-Carlo dose rate calculation software (C. Dossat); 10 - TRIPOLI-4R Version 9 S Monte-Carlo code for radioprotection (F. Damian); 11 - Realistic radioprotection training with the digital school workshop (E. Courageot); 12 - Use of BEAMNRC code for dental prostheses influence evaluation in ENT cancers treatment by external radiotherapy (C. De Conto); 13
DCHAIN 2: a computer code for calculation of transmutation of nuclides
International Nuclear Information System (INIS)
DCHAIN2 is a one-point depletion code which solves the coupled equation of radioactive growth and decay for a large number of nuclides by the Bateman method. A library of nuclear data for 1170 fission products has been prepared for providing input data to this code. The Bateman method surpasses the matrix exponential method in computational accuracies and in saving computer storage for the code. However, most existing computer codes based on the Bateman method have shown serious drawbacks in treating cyclic chains and more than a few specific types of decay chains. The present code has surmounted the above drawbacks by improving the code FP-S, and has the following characteristics: (1) The code can treat any type of transmutation through decays or neutron induced reactions. Multiple decays and reactions are allowed for a nuclide. (2) Unknown decay energy in the nuclear data library can be estimated. (3) The code constructs the decay scheme of each nuclide in the code and breaks it up into linear chains. Nuclide names, decay types and branching ratios of mother nuclides are necessary as the input data for each nuclide. Order of nuclides in the library is arbitrary because each nuclide is destinguished by its nuclide name. (4) The code can treat cyclic chains by an approximation. A library of the nuclear data has been prepared for 1170 fission products, including the data for half-lives, decay schemes, neutron absorption cross sections, fission yields, and disintegration energies. While DCHAIN2 is used to compute the compositions, radioactivity and decay heat of fission products, the gamma-ray spectrum of fission products can be computed also by a separate code FPGAM using the composition obtained from DCHAIN2. (J.P.N.)
Recording and Replaying System Specific, Source Code Transformations
Santos, Gustavo; Etien, Anne; Anquetil, Nicolas; Ducasse, Stéphane; Tulio Valente, Marco
2015-01-01
International audience During its lifetime, a software system is under continuous maintenance to remain useful. Maintenance can be achieved in activities such as adding new features, fixing bugs, improving the system's structure, or adapting to new APIs. In such cases, developers sometimes perform sequences of code changes in a systematic way. These sequences consist of small code changes (e.g., create a class, then extract a method to this class), which are applied to groups of related co...
Optimal Coding Predicts Attentional Modulation of Activity in Neural Systems
Jaramillo, Santiago; Pearlmutter, Barak A.
2007-01-01
Neuronal activity in response to a fixed stimulus has been shown to change as a function of attentional state, implying that the neural code also changes with attention. We propose an information-theoretic account of such modulation: that the nervous system adapts to optimally encode sensory stimuli while taking into account the changing relevance of different features. We show using computer simulation that such modulation emerges in a coding system informed about the uneven relevance of ...
Recent development for the ITS code system: Parallel processing and visualization
Energy Technology Data Exchange (ETDEWEB)
Fan, W.C.; Turner, C.D.; Halbleib, J.A. Sr.; Kensek, R.P.
1996-03-01
A brief overview is given for two software developments related to the ITS code system. These developments provide parallel processing and visualization capabilities and thus allow users to perform ITS calculations more efficiently. Timing results and a graphical example are presented to demonstrate these capabilities.
LWR-WIMS, a computer code for light water reactor lattice calculations
International Nuclear Information System (INIS)
LMR-WIMS is a comprehensive scheme of computation for studying the reactor physics aspects and burnup behaviour of typical lattices of light water reactors. This report describes the physics methods that have been incorporated in the code, and the modifications that have been made since the code was issued in 1972. (U.K.)
Electron and ion cyclotron heating calculations in the tandem-mirror modeling code MERTH
International Nuclear Information System (INIS)
To better understand and predict tandem-mirror experiments, we are building a comprehensive Mirror Equilibrium Radial Transport and Heating (MERTH) code. In this paper we first describe our method for developing the code. Then we report our plans for the installation of physics packages for electron- and ion-cyclotron heating of the plasma
New Burnup Calculation System for Fusion-Fission Hybrid System
International Nuclear Information System (INIS)
Investigation of nuclear waste incineration has positively been carried out worldwide from the standpoint of environmental issues. Some candidates such as ADS, FBR are under discussion for possible incineration technology. Fusion reactor is one of such technologies, because it supplies a neutron-rich and volumetric irradiation field, and in addition the energy is higher than nuclear reactor. However, it is still hard to realize fusion reactor right now, as well known. An idea of combination of fusion and fission concepts, so-called fusion-fission hybrid system, was thus proposed for the nuclear waste incineration. Even for a relatively lower plasma condition, neutrons can be well multiplied by fission in the nuclear fuel, tritium is thus bred so as to attain its self-sufficiency, enough energy multiplication is then expected and moreover nuclear waste incineration is possible. In the present study, to realize it as soon as possible with the presently proven technology, i.e., using ITER model with the achieved plasma condition of JT60 in JAEA, Japan, a new calculation system for fusion-fission hybrid reactor including transport by MCNP and burnup by ORIGEN has been developed for the precise prediction of the neutronics performance. The author's group already has such a calculation system developed by them. But it had a problem that the cross section libraries in ORIGEN did not have a cross section library, which is suitable specifically for fusion-fission hybrid reactors. So far, those for FBR were approximately used instead in the analysis. In the present study, exact derivation of the collapsed cross section for ORIGEN has been investigated, which means it is directly evaluated from calculated track length by MCNP and point-wise nuclear data in the evaluated nuclear data file like JENDL-3.3. The system realizes several-cycle calculation one time, each of which consists of MCNP criticality calculation, MCNP fixed source calculation with a 3-dimensional precise
Performance Evaluation of Hybrid ARQ with Code Combining in Packet-Oriented CDMA System
Institute of Scientific and Technical Information of China (English)
CHENQingchun; FANPingzhi
2004-01-01
In this paper, an extended SNR (signal to noise ratio) concept is proposed to explicate the contribution of code combining to the performance improvement of hybrid ARQ (Automatic repeat request) over the additive white Gaussian noise channel. By extending the Pursley's SNR analysis to hybrid ARQ with code combining in packet-oriented CDMA (Code division multiple access)system, the extended SNR formula is derived, which describes explicitly the SNR variation of the code symbol involved in code combining. It is revealed that the extended SNR formula includes Pursley's SNR formula as a specialcase. Moreover, it is shown that the effective SNR of the combined symbol is increased by a coefficient, which is proportional to the number of repeated replicas involved in the code combining. Based on the extended SNR formula and the resultant SNR variation, a quasi-analytical approximation method is proposed for the performance evaluation of hybrid ARQ with code combining. The residual error rates, average transmission number together with throughput performance are presented by means of numerical analysis and through simulations. It is validated that the extended SNR formula and the resultant quasi-analytical approximations offer a simplified routine to estimate the performance of hybrid ARQ with code combining, particularly for the applications whose reliability performance with respect to the FEC counterpart system could be numerically calculated or evaluated through simulations.
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.)
International Nuclear Information System (INIS)
In 1986 the CEC sponsored a benchmark exercise on aerosol calculation based on the Demona B3 experiment. The results of this exercise were very sensitive to the calculation of energy and mass transfer between the phases. In view of the results of the study mentioned above, it had been decided to carry out a benchmark exercise for severe accident containment thermal-hydraulics codes. This exercise is based on experiment B3 in the Demona programme. The experiment B3 was a simulation of a late overpressure failure scenario in a PWR. The main objective of the benchmark exercise has been to assess the ability of the participating codes to predict atmosphere saturation levels and bulk condensation rates under conditions similar to those predicted to follow a severe accident in a PWR. Several research organizations of the Community member countries have participated at this benchmark exercise. The paper presents the comparison of the experimental results with the following calculated quantities: total system pressure, steam pressure, containment temperature, saturation ratio, steam condensation rate on walls and in the bulk volume, water mass in sump, sump temperature, heat transfer coefficient, steam mass flow rate and structural temperature for a real time period of up to 90 hours. The paper also presents the major conclusions from the exercise in order to identify the status of present codes versus the requirements needed as input and for coupling with aerosol analysis codes
Gholamzadeh Zohreh; Hossein Feghhi Seyed Amir; Soltani Leila; Rezazadeh Marzieh; Tenreiro Claudio; Joharifard Mahdi
2014-01-01
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. N...
Sequence Coding and Search System for licensee event reports: code listings. Volume 2
International Nuclear Information System (INIS)
Operating experience data from nuclear power plants are essential for safety and reliability analyses, especially analyses of trends and patterns. The licensee event reports (LERs) that are submitted to the Nuclear Regulatory Commission (NRC) by the nuclear power plant utilities contain much of this data. The NRC's Office for Analysis and Evaluation of Operational Data (AEOD) has developed, under contract with NSIC, a system for codifying the events reported in the LERs. The primary objective of the Sequence Coding and Search System (SCSS) is to reduce the descriptive text of the LERs to coded sequences that are both computer-readable and computer-searchable. This system provides a structured format for detailed coding of component, system, and unit effects as well as personnel errors. The database contains all current LERs submitted by nuclear power plant utilities for events occurring since 1981 and is updated on a continual basis. Volume 2 contains all valid and acceptable codes used for searching and encoding the LER data. This volume contains updated material through amendment 1 to revision 1 of the working version of ORNL/NSIC-223, Vol. 2
Macroscopic multigroup constants for accelerator driven system core calculation
International Nuclear Information System (INIS)
The high-level wastes stored in facilities above ground or shallow repositories, in close connection with its nuclear power plant, can take almost 106 years before the radiotoxicity became of the order of the background. While the disposal issue is not urgent from a technical viewpoint, it is recognized that extended storage in the facilities is not acceptable since these ones cannot provide sufficient isolation in the long term and neither is it ethical to leave the waste problem to future generations. A technique to diminish this time is to transmute these long-lived elements into short-lived elements. The approach is to use an Accelerator Driven System (ADS), a sub-critical arrangement which uses a Spallation Neutron Source (SNS), after separation the minor actinides and the long-lived fission products (LLFP), to convert them to short-lived isotopes. As an advanced reactor fuel, still today, there is a few data around these type of core systems. In this paper we generate macroscopic multigroup constants for use in calculations of a typical ADS fuel, take into consideration, the ENDF/BVI data file. Four energy groups are chosen to collapse the data from ENDF/B-VI data file by PREPRO code. A typical MOX fuel cell is used to validate the methodology. The results are used to calculate one typical subcritical ADS core. (author)
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 translatio
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
Unidirectional Error Correcting Codes for Memory Systems: A Comparative Study
Al-Ani, Muzhir
2010-01-01
In order to achieve fault tolerance, highly reliable system often require the ability to detect errors as soon as they occur and prevent the speared of erroneous information throughout the system. Thus, the need for codes capable of detecting and correcting byte errors are extremely important since many memory systems use b-bit-per-chip organization. Redundancy on the chip must be put to make fault-tolerant design available. This paper examined several methods of computer memory systems, and then a proposed technique is designed to choose a suitable method depending on the organization of memory systems. The constructed codes require a minimum number of check bits with respect to codes used previously, then it is optimized to fit the organization of memory systems according to the requirements for data and byte lengths.
EBT reactor systems analysis and cost code: description and users guide (Version 1)
International Nuclear Information System (INIS)
An ELMO Bumpy Torus (EBT) reactor systems analysis and cost code that incorporates the most recent advances in EBT physics has been written. The code determines a set of reactors that fall within an allowed operating window determined from the coupling of ring and core plasma properties and the self-consistent treatment of the coupled ring-core stability and power balance requirements. The essential elements of the systems analysis and cost code are described, along with the calculational sequences leading to the specification of the reactor options and their associated costs. The input parameters, the constraints imposed upon them, and the operating range over which the code provides valid results are discussed. A sample problem and the interpretation of the results are also presented
BCG: a computer code for calculating neutron spectra and criticality in cells of fast reactors
International Nuclear Information System (INIS)
The BCG code for determining the space and energy neutron flux distribution and criticality of fast reactor cylindrical cells is discussed. The code solves the unidimensional neutron transport equation together with interface current relations at each energy point in an unionized energy grid prepared for the cell and at an arbitrary number of spatial zones. While the spatial resolution is user specified, the energy dependence of the flux distribution is resolved according to the degree of variation in the reconstruced total microscopic cross sections of the atomic species in the cell. Results for a simplified fuel cell illustrate the high resolution and accuracy that can be obtained with the code. (author)
Directory of Open Access Journals (Sweden)
Skrzypek Maciej
2015-09-01
Full Text Available The main object of interest was a typical fuel assembly, which constitutes a core of the nuclear reactor. The aim of the paper is to describe the phenomena and calculate thermal-hydraulic characteristic parameters in the fuel assembly for a European Pressurized Reactor (EPR. To perform thermal-hydraulic calculations, the RELAP5 code was used. This code allows to simulate steady and transient states for reactor applications. It is also an appropriate calculation tool in the event of a loss-of-coolant accident in light water reactors. The fuel assembly model with nodalization in the RELAP5 (Reactor Excursion and Leak Analysis Program code was presented. The calculations of two steady states for the fuel assembly were performed: the nominal steady-state conditions and the coolant flow rate decreased to 60% of the nominal EPR flow rate. The calculation for one transient state for a linearly decreasing flow rate of coolant was simulated until a new level was stabilized and SCRAM occurred. To check the correctness of the obtained results, the authors compared them against the reactor technical documentation available in the bibliography. The obtained results concerning steady states nearly match the design data. The hypothetical transient showed the importance of the need for correct cooling in the reactor during occurrences exceeding normal operation. The performed analysis indicated consequences of the coolant flow rate limitations during the reactor operation.
Construction Zero Cross Correlation Code using Permutation Matrix for SAC-OCDMA Systems
Nisar, K. S.
2016-01-01
This paper present a new method for constructing zero cross correlation code with the help of permutation matrices. The benefits of this newly proposed code are easy way code construction, the code weight exist for every natural number and the code length is acceptable. The numerical comparison shows that the proposed code has better or compatible code length compared with other existing zero cross correlation code in Optical Spectrum Code Division Multiple Access (OSCDMA) systems.
Development of a GPU-based Monte Carlo dose calculation code for coupled electron-photon transport
Jia, Xun; Sempau, Josep; Choi, Dongju; Majumdar, Amitava; Jiang, Steve B
2009-01-01
Monte Carlo simulation is the most accurate method for absorbed dose calculations in radiotherapy. Its efficiency still requires improvement for routine clinical applications, especially for online adaptive radiotherapy. In this paper, we report our recent development on a GPU-based Monte Carlo dose calculation code for coupled electron-photon transport. We have implemented the Dose Planning Method (DPM) Monte Carlo dose calculation package (Sempau et al, Phys. Med. Biol., 45(2000)2263-2291) on GPU architecture under CUDA platform. The implementation has been tested with respect to the original sequential DPM code on CPU in two cases. Our results demonstrate the adequate accuracy of the GPU implementation for both electron and photon beams in radiotherapy energy range. A speed up factor of 4.5 and 5.5 times have been observed for electron and photon testing cases, respectively, using an NVIDIA Tesla C1060 GPU card against a 2.27GHz Intel Xeon CPU processor .
International Nuclear Information System (INIS)
. The neutronic calculation has been carried out for TRIGA 2 MW reactor. These included criticality flux and power distributions. Computer code Citation which solves 7-groups, 3-dimensional hexagonal geometry has been used. The multi groups-cross-section is generated by the WIMS-D/4 code.This 7-group-39x39x38-mesh-points problem takes about 90 minutes on the Pentium-133 MHz PC. The calculation of the initial core of TRIGA 2 MW reactor shows that the excess reactivity of the core is 7,8% and the thermal fluxes in the irradiation positions are between 1.0-2.9*1013n cm-2s-1. The results are about 10% deviate from those calculated by General Atomics. In the initial core, the highest power is produced in the C-9 position. The fuel element in this position produces 30.7 k W thermal power
Automatic code generation for distributed robotic systems
International Nuclear Information System (INIS)
Hetero Helix is a software environment which supports relatively large robotic system development projects. The environment supports a heterogeneous set of message-passing LAN-connected common-bus multiprocessors, but the programming model seen by software developers is a simple shared memory. The conceptual simplicity of shared memory makes it an extremely attractive programming model, especially in large projects where coordinating a large number of people can itself become a significant source of complexity. We present results from three system development efforts conducted at Oak Ridge National Laboratory over the past several years. Each of these efforts used automatic software generation to create 10 to 20 percent of the system
Hydrogen detection systems leak response codes
International Nuclear Information System (INIS)
A loss in tightness of a water tube inside a Steam Generator Unit of a Fast Reactor is usually monitored by hydrogen detection systems. Such systems have demonstrated in the past their ability to detect a leak in a SGU. However, the increase in size of the SGU or the choice of ferritic material entails improvement of these systems in order to avoid secondary leak or to limit damages to the tube bundle. The R and D undertaken in France on this subject is presented. (author). 11 refs, 10 figs
Building Secure Networked Systems with Code Attestation
Perrig, Adrian
Attestation is a promising approach for building secure systems. The recent development of a Trusted Platform Module (TPM) by the Trusted Computing Group (TCG) that is starting to be deployed in common laptop and desktop platforms is fueling research in attestation mechanisms. In this talk, we will present approaches on how to build secure systems with advanced TPM architectures. In particular, we have designed an approach for fine-grained attestation that enables the design of efficient secure distributed systems, and other network protocols.We demonstrate this approach by designing a secure routing protocol.
Total System Performance Assessment Code (TOSPAC)
International Nuclear Information System (INIS)
TOSPAC is a computer program that calculates partially saturated groundwater flow with the transport of water-soluble contaminants. TOSPAC Version 1 is restricted to calculations involving one-dimensional, vertical columns of one or more media. TOSPAC was developed to help answer questions surrounding the burial of toxic wastes in arid regions. Burial of wastes in arid regions is attractive because of generally low population densities and little groundwater flow, in the unsaturated zone, to disturb the waste. TOSPAC helps to quantify groundwater flow and the spread of contamination, offering an idea of what could happen in the distant future. Figure 1.1 illustrates the problem TOSPAC was designed to investigate. For groundwater flow, TOSPAC can provide saturations, velocities, and and travel tunes for water in the rock matrix or the fractures in the unsaturated zone. TOSPAC can determine how hydrologic conditions vary when the rate of infiltration changes. For contaminant transport, TOSPAC can compute how much of a contaminant is dissolved in the water and how it is distributed. TOSPAC can determine how fast the solute is moving and the shape of the concentration front. And TOSPAC can be used to investigate how much of the contaminant remains in the inventory of a repository, how much is adsorbed onto the soil or rock matrix, and how much reaches the water table. Effective use of TOSPAC requires knowledge in a number of diverse disciplines, including real groundwater flow and transport, the mathematical models of groundwater flow and transport, real-world data required for the models, and the numerical solution of differential equations. Equally important is a realization of the limitations intrinsic to a computer model of complex physical phenomena. This User's Guide not only describes the mechanics of executing TOSPAC on a computer, but also examines these other topics
SRAC: JAERI thermal reactor standard code system for reactor design and analysis
International Nuclear Information System (INIS)
The SRAC (Standard Reactor Analysis Code) is a code system for nuclear reactor analysis and design. It is composed of neutron cross section libraries and auxiliary processing codes, neutron spectrum routines, a variety of transport, 1-, 2- and 3-D diffusion routines, dynamic parameters and cell burn-up routines. By making the best use of the individual code function in the SRAC system, the user can select either the exact method for an accurate estimate of reactor characteristics or the economical method aiming at a shorter computer time, depending on the purpose of study. The user can select cell or core calculation; fixed source or eigenvalue problem; transport (collision probability or Sn) theory or diffusion theory. Moreover, smearing and collapsing of macroscopic cross sections are separately done by the user's selection. And a special attention is paid for double heterogeneity. Various techniques are employed to access the data storage and to optimize the internal data transfer. Benchmark calculations using the SRAC system have been made extensively for the Keff values of various types of critical assemblies (light water, heavy water and graphite moderated systems, and fast reactor systems). The calculated results show good prediction for the experimental Keff values. (author)
International Nuclear Information System (INIS)
All of the three exercises of the Organization for Economic Cooperation and Development/Nuclear Regulatory Commission pressurized water reactor main steam line break (PWR MSLB) benchmark were calculated at VTT, the Technical Research Centre of Finland. For the first exercise, the plant simulation with point-kinetic neutronics, the thermal-hydraulics code SMABRE was used. The second exercise was calculated with the three-dimensional reactor dynamics code TRAB-3D, and the third exercise with the combination TRAB-3D/SMABRE. VTT has over ten years' experience of coupling neutronic and thermal-hydraulic codes, but this benchmark was the first time these two codes, both developed at VTT, were coupled together. The coupled code system is fast and efficient; the total computation time of the 100-s transient in the third exercise was 16 min on a modern UNIX workstation. The results of all the exercises are similar to those of the other participants. In order to demonstrate the effect of secondary circuit modeling on the results, three different cases were calculated. In case 1 there is no phase separation in the steam lines and no flow reversal in the aspirator. In case 2 the flow reversal in the aspirator is allowed, but there is no phase separation in the steam lines. Finally, in case 3 the drift-flux model is used for the phase separation in the steam lines, but the aspirator flow reversal is not allowed. With these two modeling variations, it is possible to cover a remarkably broad range of results. The maximum power level reached after the reactor trip varies from 534 to 904 MW, the range of the time of the power maximum being close to 30 s. Compared to the total calculated transient time of 100 s, the effect of the secondary side modeling is extremely important
Energy Technology Data Exchange (ETDEWEB)
Eisenbach, Markus [ORNL; Larkin, Jeff [NVIDIA, Santa Clara, CA; Lutjens, Justin [NVIDIA, Santa Clara, CA; Rennich, Steven [NVIDIA, Santa Clara, CA; Rogers, James H [ORNL
2016-01-01
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. 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.
JPEG2000 COMPRESSION CODING USING HUMAN VISUAL SYSTEM MODEL
Institute of Scientific and Technical Information of China (English)
Xiao Jiang; Wu Chengke
2005-01-01
In order to apply the Human Visual System (HVS) model to JPEG2000 standard,several implementation alternatives are discussed and a new scheme of visual optimization isintroduced with modifying the slope of rate-distortion. The novelty is that the method of visual weighting is not lifting the coefficients in wavelet domain, but is complemented by code stream organization. It remains all the features of Embedded Block Coding with Optimized Truncation (EBCOT) such as resolution progressive, good robust for error bit spread and compatibility of lossless compression. Well performed than other methods, it keeps the shortest standard codestream and decompression time and owns the ability of VIsual Progressive (VIP) coding.
Effects of bar coding on a pharmacy stock replenishment system.
Chester, M I; Zilz, D A
1989-07-01
A bar-code stock ordering system installed in the ambulatory-care pharmacy and sterile products area of a hospital pharmacy was compared with a manual paper system to quantify overall time demands and determine the error rate associated with each system. The bar-code system was implemented in the ambulatory-care pharmacy in November 1987 and in the sterile products area in January 1988. It consists of a Trakker 9440 transaction manager with a digital scanner; labels are printed with a dot matrix printer. Electronic scanning of bar-code labels and entry of the amount required using the key-pad on the transaction manager replaced use of a preprinted form for ordering items. With the bar-code system, ordering information is transferred electronically via cable to the pharmacy inventory computer; with the manual system, this information was input by a stockroom technician. To compare the systems, the work of technicians in the ambulatory-care pharmacy and sterile products area was evaluated before and after implementation of the bar-code system. The time requirements for information gathering and data transfer were recorded by direct observation; the prevalence of errors under each system was determined by comparing unprocessed ordering information with the corresponding computer-generated "pick lists" (itemized lists including the amount of each product ordered). Time consumed in extra trips to the stockroom to replace out-of-stock items was self-reported. Significantly less time was required to order stock and transfer data to the pharmacy inventory computer with the bar-code system than with the manual system.(ABSTRACT TRUNCATED AT 250 WORDS) PMID:2757044
Energy Technology Data Exchange (ETDEWEB)
Takeda, N. [Electrotechnical Laboratory, 1-1-4 Umezono, Tsukuba-shi, Ibaraki 305-8568 (Japan); Kudo, K. [Electrotechnical Laboratory, 1-1-4 Umezono, Tsukuba-shi, Ibaraki 305-8568 (Japan); Toyokawa, H. [Electrotechnical Laboratory, 1-1-4 Umezono, Tsukuba-shi, Ibaraki 305-8568 (Japan); Torii, T. [Japan Power Reactor and Nuclear Fuel Development Corporation, Tsuruga Office, Fukui 919-12 (Japan); Hashimoto, M. [Japan Power Reactor and Nuclear Fuel Development Corporation, O-arai Engineering Center, Ibaraki 311-13 (Japan); Sugita, T. [Science System Laboratory, Ibaraki 309-17 (Japan); Dietze, G. [Physikalisch-Technische Bundesanstalt, 38023 Braunschweig (Germany); Yang, X. [China Institute of Atomic Energy (China)
1999-02-11
A Monte Carlo code Neutron RESPonse function for Gas counters (NRESPG) has been developed for the calculation of neutron response functions and efficiencies for neutron energies up to 20 MeV, which can be applied for {sup 3}He, H{sub 2}, or BF{sub 3} gas proportional counters with or without moderator. This code can simulate the neutron behavior in a two-dimensional detector configuration and treat the thermal motion of a moderator atom which becomes important as the neutron energy becomes sufficiently low. Further, a more precise measured data was taken to simulate the position-dependent gas multiplication in the sensitive and insensitive gas region of a proportional counter. The NRESPG code has been applied for the calculation of response functions of {sup 3}He cylindrical proportional counters to determine neutron energy and neutron fluence in a monoenergetic calibration field. Thus, a remarkable discrepancy in the lower portion of the full-energy peak produced by the {sup 3}He(n,p)T reaction can be removed which results in a good agreement between simulations and experiments. The code has been also used for the simulation of the response of a McTaggart-type long counter consisting of a central cylindrical BF{sub 3} counter surrounded by a polyethylene moderator. The results of the NRESPG simulations were compared with those obtained from MCNP calculations.
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)
International Nuclear Information System (INIS)
A Monte Carlo code Neutron RESPonse function for Gas counters (NRESPG) has been developed for the calculation of neutron response functions and efficiencies for neutron energies up to 20 MeV, which can be applied for 3He, H2, or BF3 gas proportional counters with or without moderator. This code can simulate the neutron behavior in a two-dimensional detector configuration and treat the thermal motion of a moderator atom which becomes important as the neutron energy becomes sufficiently low. Further, a more precise measured data was taken to simulate the position-dependent gas multiplication in the sensitive and insensitive gas region of a proportional counter. The NRESPG code has been applied for the calculation of response functions of 3He cylindrical proportional counters to determine neutron energy and neutron fluence in a monoenergetic calibration field. Thus, a remarkable discrepancy in the lower portion of the full-energy peak produced by the 3He(n,p)T reaction can be removed which results in a good agreement between simulations and experiments. The code has been also used for the simulation of the response of a McTaggart-type long counter consisting of a central cylindrical BF3 counter surrounded by a polyethylene moderator. The results of the NRESPG simulations were compared with those obtained from MCNP calculations
SCALE: A modular code system for performing standardized computer analyses for licensing evaluation
International Nuclear Information System (INIS)
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
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.
The FORTRAN static source code analyzer program (SAP) system description
Decker, W.; Taylor, W.; Merwarth, P.; Oneill, M.; Goorevich, C.; Waligora, S.
1982-01-01
A source code analyzer program (SAP) designed to assist personnel in conducting studies of FORTRAN programs is described. The SAP scans FORTRAN source code and produces reports that present statistics and measures of statements and structures that make up a module. The processing performed by SAP and of the routines, COMMON blocks, and files used by SAP are described. The system generation procedure for SAP is also presented.
Energy Technology Data Exchange (ETDEWEB)
Paratte, J.M.; Grimm, P.; Hollard, J.M. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)
1996-02-01
ELCOS is a flexible code system for the stationary simulation of light water reactor cores. It consists of the four computer codes ETOBOX, BOXER, CORCOD and SILWER. The user`s manual of the second one is presented here. BOXER calculates the neutronics in cartesian geometry. The code can roughly be divided into four stages: - organisation: choice of the modules, file manipulations, reading and checking of input data, - fine group fluxes and condensation: one-dimensional calculation of fluxes and computation of the group constants of homogeneous materials and cells, - two-dimensional calculations: geometrically detailed simulation of the configuration in few energy groups, - burnup: evolution of the nuclide densities as a function of time. This manual shows all input commands which can be used while running the different modules of BOXER. (author) figs., tabs., refs.
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.
Electrical utility generating system reliability analysis code, SYSREL. Social cost studies program
Energy Technology Data Exchange (ETDEWEB)
Hub, K.; Conley, L.; Buehring, W.; Rowland, B.; Stephenson, M.
1975-09-01
The system reliability code, SYSREL, is a system planning tool that can be used to assess the reliability and economic performance of alternative expansion patterns of electric utility generation systems. Given input information such as capacity, forced-outage rate, number of weeks of annual scheduled maintenance, and economic data for individual units along with the expected load characteristics, the code produces estimates of the mean time between system failures, required reserve capacity to meet a specified system-failure-frequency criterion, expected energy generation from each unit, and system energy cost. The categories of calculations performed by the code are maintenance scheduling, reliability, capacity requirement, energy production allocation, and energy cost. The code is designed to examine alternative generating units and system expansion patterns based on the constraints and general economic conditions imposed by the investigator. The computer running time to execute a study is short and many system alternatives can be examined at a relatively low cost. The report contains a technical description of the code, list of input data requirements, program listing, sample execution, and parameter studies. (auth)
Validation of system codes for plant application on selected experiments
Energy Technology Data Exchange (ETDEWEB)
Koch, Marco K.; Risken, Tobias; Agethen, Kathrin; Bratfisch, Christoph [Bochum Univ. (Germany). Reactor Simulation and Safety Group
2016-05-15
For decades, the Reactor Simulation and Safety Group at Ruhr-Universitaet Bochum (RUB) contributes to nuclear safety by computer code validation and model development for nuclear safety analysis. Severe accident analysis codes are relevant tools for the understanding and the development of accident management measures. The accidents in the plants Three Mile Island (USA) in 1979 and Fukushima Daiichi (Japan) in 2011 influenced these research activities significantly due to the observed phenomena, such as molten core concrete interaction and hydrogen combustion. This paper gives a brief outline of recent research activities at RUB in the named fields, contributing to code preparation for plant applications. Simulations of the molten core concrete interaction tests CCI-2 and CCI-3 with ASTEC and the hydrogen combustion test Ix9 with COCOSYS are presented exemplarily. Additionally, the application on plants is demonstrated on chosen results of preliminary Fukushima calculations.
Validation of system codes for plant application on selected experiments
International Nuclear Information System (INIS)
For decades, the Reactor Simulation and Safety Group at Ruhr-Universitaet Bochum (RUB) contributes to nuclear safety by computer code validation and model development for nuclear safety analysis. Severe accident analysis codes are relevant tools for the understanding and the development of accident management measures. The accidents in the plants Three Mile Island (USA) in 1979 and Fukushima Daiichi (Japan) in 2011 influenced these research activities significantly due to the observed phenomena, such as molten core concrete interaction and hydrogen combustion. This paper gives a brief outline of recent research activities at RUB in the named fields, contributing to code preparation for plant applications. Simulations of the molten core concrete interaction tests CCI-2 and CCI-3 with ASTEC and the hydrogen combustion test Ix9 with COCOSYS are presented exemplarily. Additionally, the application on plants is demonstrated on chosen results of preliminary Fukushima calculations.
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)
Overview of Particle and Heavy Ion Transport Code System PHITS
International Nuclear Information System (INIS)
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. 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. This paper briefly summarizes the physics models implemented in PHITS, and introduces some important functions useful for specific applications. The important functions of PHITS are an event generator mode for low-energy neutron interaction, beam transport functions, a function for calculating the displacement per atom (DPA), and a microdosimetric tally function. PHITS has been used by more than 1,000 users in various research and development fields, such as nuclear technology, accelerator design, medical physics, and cosmic-ray research
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.
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.
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)
Application bar-code system for solid radioactive waste management
International Nuclear Information System (INIS)
Solid radioactive wastes are generated from the post-irradiated fuel examination facility, the irradiated material examination facility, the research reactor, and the laboratories at KAERI. A bar-code system for a solid radioactive waste management of a research organization became necessary while developing the RAWMIS(Radioactive Waste Management Integration System) which it can generate personal history management for efficient management of a waste, documents, all kinds of statistics. This paper introduces an input and output application program design to do to database with data in the results and a stream process of a treatment that analyzed the waste occurrence present situation and data by bar-code system
Input modelling of PHT system stability analysis for CANFLEX-RU bundle by SOPHT code
Energy Technology Data Exchange (ETDEWEB)
Park, Joo Hwan; Jun, Ji Su; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea)
1999-01-01
The overall objective of this report is to undertake the stability analysis of primary heat transport (PHT) system for a CANDU reactor to be loaded with CANFLEX-RU bundle which is to provide a vehicle for the economic use of recycled uranium and for the economic provision of additional operating margins in aging CANDU reactors. The modelling report for the input data of SOPHT code is required in order to give the specific and accurate information for flow stability calculation. This report is consisted of the several sections which are described the usage of control cards, calculation methods for input data generation, the comparison of specific input data set for 37-element, CANFLEX-NU and CANFLEX-RU bundles and calculation results for the steady state condition. Those input data set prepared will be used for the flow stability analysis of PHT system of CANDU reactor by SOPHT code. (author). 1 ref., 2 figs., 16 tabs.
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.
Improved Iono PHMI Calculation for SBAS Systems
Mayer, Christoph; Blanch, Juan
2009-01-01
Besides the wide-area augmentation system (WAAS) in the US, an increasing number of space-based augmentation systems (SBAS) are being developed or planned, such as the and the European Geostationary Navigation Overlay System (EGNOS) in Europe, the Multi-functional Satellite Augmentation System (MSAS) system in Japan, and the future Ground-based Regional Augmentation System (GRAS) in Australia and the GPS and Geo Augmented Navigation (GAGAN) system covering the Indian subcontinent. The int...
International Nuclear Information System (INIS)
This report describes the calculation procedure of the TRANCS code, which deals with fission product transport in fuel rod of high temperature gas-cooled reactor (HTGR). The fundamental equation modeled in the code is a cylindrical one-dimensional diffusion equation with generation and decay terms, and the non-stationary solution of the equation is obtained numerically by a finite difference method. The generation terms consist of the diffusional release from coated fuel particles, recoil release from outer-most coating layer of the fuel particle and generation due to contaminating uranium in the graphite matrix of the fuel compact. The decay term deals with neutron capture as well as beta decay. Factors affecting the computation error has been examined, and further extention of the code has been discussed in the fields of radial transport of fission products from graphite sleeve into coolant helium gas and axial transport in the fuel rod. (author)
Optical System Design For High Speed Bar Code Scanning
Hellekson, Ronald; Reddersen, Brad; Campbell, Scott
1987-04-01
Spectra-Physics recently introduced the Model 750 SL scanner for use in the European point-of-sale market, to meet the European requirement for a scanner of less than 13 cm height. The model 750 SL uses a higher density computer designed scan pattern with a retrodirective collection system to scan and detect UPC, EAN, and JAN bar codes. The scanner "reads" these bar codes in such a way that the user need not precisely align the bar code symbol with respect to the window in the scanner even at package speeds up to 100 inches per second. By using a unique geometrical arrangement of mirrors, a polygonal mirror assembly, and a custom-designed plastic bifocal lens, a design was developed to meet these requirements. This paper describes the design of this new low cost scanner, the use of computer-aided design in the development of this scanner, and some observations on the future of bar code scanning.
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.
Energy Technology Data Exchange (ETDEWEB)
Glueckstern, P.; Reed, S.A.; Wilson, J.V.
1976-11-01
The reverse osmosis process has been used extensively for the conversion of brackish waters to potable water. The process is now nearing commercialization as a means for the conversion of seawater. The computer program (RO-75) is a Fortran code for the optimizatin of the design and economics of seawater reverse osmosis plants. The examples described are based on currently available, commercial membrane modules and prevailing prices. However, the code is very flexible and can be used to optimize plants utilizing future technological improvements and different economic parameters.
Energy Technology Data Exchange (ETDEWEB)
Forestier, Benoit; Miss, Joachim; Bernard, Franck; Dorval, Aurelien [Institut de Radioprotection et Surete Nucleaire, Fontenay aux Roses (France); Jacquet, Olivier [Independent consultant (France); Verboomen, Bernard [Belgian Nuclear Research Center - SCK-CEN (Belgium)
2008-07-01
The MORET code is a three dimensional Monte Carlo criticality code. It is designed to calculate the effective multiplication factor (k{sub eff}) of any geometrical configuration as well as the reaction rates in the various volumes and the neutron leakage out of the system. A recent development for the MORET code consists of the implementation of an alternate neutron tracking method, known as the pseudo-scattering tracking method. This method has been successfully implemented in the MORET code and its performances have been tested by mean of an extensive parametric study on very simple geometrical configurations. In this context, the goal of the present work is to validate the pseudo-scattering method against realistic configurations. In this perspective, pebble-bed cores are particularly well-adapted cases to model, as they exhibit large amount of volumes stochastically arranged on two different levels (the pebbles in the core and the TRISO particles inside each pebble). This paper will introduce the techniques and methods used to model pebble-bed cores in a realistic way. The results of the criticality calculations, as well as the pseudo-scattering tracking method performance in terms of computation time, will also be presented. (authors)
System-Level Genetic Codes Using a Transposable Element-Like Mechanism with Applications to Cancer
McGowan, John F.
2000-01-01
A system-level genetic code is a hypothetical genetic code that exclusively or preferentially codes systems of interacting coadapted parts. System-level genetic codes differ from part-level genetic codes in which each discrete part is coded independently. In general, a system-level genetic code requires coding discrete interacting parts such as organs or proteins in an interdependent way. Changing a single symbol or "gene" in a system-level genetic code affects two or more parts in a coordina...
International Nuclear Information System (INIS)
The need of the experimental support for validation of the computational tools to be applied to analyze the mixing of diluted slugs has been recognized in various countries. The test series for the International Standard Problem ISP-43 provides a platform for experiences to be applied to the simulation of a well-defined test series. Test A and B of the UM2x4 loop test facility were calculated with the CFD Code CFX-4.3. The results show qualitatively good agreement with the experimental data for both tests. The structure of the flow field and the form of the propagating temperature perturbation front are well modeled by the CFD code. However, deviations occur at local positions. Comparative calculations with and without taking into account buoyancy have shown, that buoyancy effects are noticeable, but the mixing is mainly momentum controlled. (orig.)
International Nuclear Information System (INIS)
In this paper, multi-group microscopic cross-section uncertainty is propagated through the DRAGON (Version 4) lattice code, in order to perform uncertainty analysis on k∞ and 2-group homogenized macroscopic cross-sections predictions. A statistical methodology is employed for such purposes, where cross-sections of certain isotopes of various elements belonging to the 172 groups DRAGLIB library format, are considered as normal random variables. This library is based on JENDL-4 data, because JENDL-4 contains the largest amount of isotopic covariance matrixes among the different major nuclear data libraries. The aim is to propagate multi-group nuclide uncertainty by running the DRAGONv4 code 500 times, and to assess the output uncertainty of a test case corresponding to a 17 x 17 PWR fuel assembly segment without poison. The chosen sampling strategy for the current study is Latin Hypercube Sampling (LHS). The quasi-random LHS allows a much better coverage of the input uncertainties than simple random sampling (SRS) because it densely stratifies across the range of each input probability distribution. Output uncertainty assessment is based on the tolerance limits concept, where the sample formed by the code calculations infers to cover 95% of the output population with at least a 95% of confidence. This analysis is the first attempt to propagate parameter uncertainties of modern multi-group libraries, which are used to feed advanced lattice codes that perform state of the art resonant self-shielding calculations such as DRAGONv4. (authors)
WETAIR: A computer code for calculating thermodynamic and transport properties of air-water mixtures
Fessler, T. E.
1979-01-01
A computer program subroutine, WETAIR, was developed to calculate the thermodynamic and transport properties of air water mixtures. It determines the thermodynamic state from assigned values of temperature and density, pressure and density, temperature and pressure, pressure and entropy, or pressure and enthalpy. The WETAIR calculates the properties of dry air and water (steam) by interpolating to obtain values from property tables. Then it uses simple mixing laws to calculate the properties of air water mixtures. Properties of mixtures with water contents below 40 percent (by mass) can be calculated at temperatures from 273.2 to 1497 K and pressures to 450 MN/sq m. Dry air properties can be calculated at temperatures as low as 150 K. Water properties can be calculated at temperatures to 1747 K and pressures to 100 MN/sq m. The WETAIR is available in both SFTRAN and FORTRAN.
The Wims-Traca code for the calculation of fuel elements. User's manual and input data
International Nuclear Information System (INIS)
The set of modifications and new options developped for the Wims-D code is explained. The input data of the new version Wims-Traca are described. The printed output of results is also explained. The contents and the source of the nuclear data in the basic library is exposed. (author)
V.S.O.P.('94) computer code system for reactor physics and fuel cycle simulation
International Nuclear Information System (INIS)
V.S.O.P. (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories and temporary in-depth research. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation with depletion and shut-down features, in-core and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to HTR's). Various techniques have been employed to accelerate the iterative processes and to optimize the internal data transfer. The storage requirement is confined to 17 M-Bytes. The code system has extensively been used for comparison studies of reactors, their fuel cycles, simulation of safety features, developmental research, and reactor assessments. Beside its use in research and development work for the gas cooled High Temperature Reactor the code has succesfully been applied to Light Water Reactors, Heavy Water Reactors, and hybride systems with different moderators. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Zagar, T.; Ravnik, M.; Persic, A. (J.Stefan Institute, Ljubljana (Slovenia))
1999-12-15
Results of fuel element burn-up determination by measurement and calculation are given. Fuel element burn-up was calculated with two different programs TRIGLAV and TRIGAC using different models. New TRIGLAV code is based on cylindrical, two-dimensional geometry with four group diffusion approximation. TRIGAC program uses one-dimensional cylindrical geometry with twogroup diffusion approximation. Fuel element burn-up was measured with reactivity method. In this paper comparison and analysis of these three methods is presented. Results calculated with TRIGLAV show considerably better alignment with measured values than results calculated with TRIGAC. Some two-dimensional effects in fuel element burn-up can be observed, for instance smaller standard fuel element burn-up in mixed core rings and control rod influence on nearby fuel elements. (orig.)
Schmidt, James F.
1995-01-01
An off-design axial-flow compressor code is presented and is available from COSMIC for predicting the aerodynamic performance maps of fans and compressors. Steady axisymmetric flow is assumed and the aerodynamic solution reduces to solving the two-dimensional flow field in the meridional plane. A streamline curvature method is used for calculating this flow-field outside the blade rows. This code allows for bleed flows and the first five stators can be reset for each rotational speed, capabilities which are necessary for large multistage compressors. The accuracy of the off-design performance predictions depend upon the validity of the flow loss and deviation correlation models. These empirical correlations for the flow loss and deviation are used to model the real flow effects and the off-design code will compute through small reverse flow regions. The input to this off-design code is fully described and a user's example case for a two-stage fan is included with complete input and output data sets. Also, a comparison of the off-design code predictions with experimental data is included which generally shows good agreement.
Energy Technology Data Exchange (ETDEWEB)
Liang, Jingang; Wang, Kan; Qiu, Yishu [Dept. of Engineering Physics, LiuQing Building, Tsinghua University, Beijing (China); Chai, Xiao Ming; Qiang, Sheng Long [Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu (China)
2016-06-15
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.
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.
Energy Technology Data Exchange (ETDEWEB)
Ohta, Masayuki, E-mail: ohta.masayuki@jaea.go.jp [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan); Takakura, Kosuke; Ochiai, Kentaro; Sato, Satoshi; Konno, Chikara [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan)
2013-10-15
In order to examine a basic performance of the TRIPOLI code, two types of analyses were carried out with TRIPOLI-4.4 and MCNP5-1.40; one is a simple model calculation and the other is an analysis of iron fusion neutronics experiments with DT neutrons at the Fusion Neutronics Source (FNS) facility in Japan Atomic Energy Agency (JAEA). In the simple model calculation, we adopted a sphere of 0.5 m in radius with a 20 MeV neutron source in the center and calculated leakage neutron spectra from the sphere. We also analyzed in situ and Time-of-Flight (TOF) experiments for iron at JAEA/FNS. For the in situ experiment, neutron spectra and reaction rates for dosimetry reactions were calculated for several points inside the assembly. For the TOF experiment, angular neutron leakage spectra from the assembly were calculated. Results with TRIPOLI were comparable to those with MCNP in most calculations, but a difference between TRIPOLI and MCNP calculation results, probably caused by inadequate treatment of inelastic scattering data in TRIPOLI, appears in some calculations.
A systems neurophysiology approach to voluntary event coding.
Petruo, Vanessa A; Stock, Ann-Kathrin; Münchau, Alexander; Beste, Christian
2016-07-15
Mechanisms responsible for the integration of perceptual events and appropriate actions (sensorimotor processes) have been subject to intense research. Different theoretical frameworks have been put forward with the "Theory of Event Coding (TEC)" being one of the most influential. In the current study, we focus on the concept of 'event files' within TEC and examine what sub-processes being dissociable by means of cognitive-neurophysiological methods are involved in voluntary event coding. This was combined with EEG source localization. We also introduce reward manipulations to delineate the neurophysiological sub-processes most relevant for performance variations during event coding. The results show that processes involved in voluntary event coding included predominantly stimulus categorization, feature unbinding and response selection, which were reflected by distinct neurophysiological processes (the P1, N2 and P3 ERPs). On a system's neurophysiological level, voluntary event-file coding is thus related to widely distributed parietal-medial frontal networks. Attentional selection processes (N1 ERP) turned out to be less important. Reward modulated stimulus categorization in parietal regions likely reflecting aspects of perceptual decision making but not in other processes. The perceptual categorization stage appears central for voluntary event-file coding. PMID:27153981
Application of CASMO-3/MASTER Code System to the OPR1000
Energy Technology Data Exchange (ETDEWEB)
You, Guk Jong; Sim, Jung Hoon; Kim, Han Gon [KHN, Daejeon (Korea, Republic of)
2007-10-15
MASTER(Multi-purpose Analyzer for Static and Transient Effects of Reactors), which was developed by KAERI, is the nuclear design code having the capability of static core design, transient core analysis and operational support. And CASMO-3, which is a fuel assembly burnup program, is the lattice calculation code to generate cross sections for core design code. To validate the core design of APR1400 CASMO- 3/MASTER codes have been selected as independent code system. The core design of APR1400, however, is in progress and the final design data and analysis results are not produced. Therefore, OPR1000, which has sufficient information, is selected as a reference plant to demonstrate the performance of CASMO-3/MASTER code package. This demonstration has been performed using design data of UCN no.4 Cycle1 and the results are compared to Nuclear Design Report(NDR) of the UCN no.4 Cycle 1. The performance of the code package is verified through uncertainty quantification according to the uncertainty evaluation report written by KAERI.
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.
Directory of Open Access Journals (Sweden)
Mohammadnia Meysam
2013-01-01
Full Text Available The flux expansion nodal method is a suitable method for considering nodalization effects in node corners. In this paper we used this method to solve the intra-nodal flux analytically. Then, a computer code, named MA.CODE, was developed using the C# programming language. The code is capable of reactor core calculations for hexagonal geometries in two energy groups and three dimensions. The MA.CODE imports two group constants from the WIMS code and calculates the effective multiplication factor, thermal and fast neutron flux in three dimensions, power density, reactivity, and the power peaking factor of each fuel assembly. Some of the code's merits are low calculation time and a user friendly interface. MA.CODE results showed good agreement with IAEA benchmarks, i. e. AER-FCM-101 and AER-FCM-001.
Rateless Space Time Block Code for Massive MIMO Systems
Directory of Open Access Journals (Sweden)
Ali H. Alqahtani
2014-01-01
Full Text Available This paper presents a rateless space time block code (RSTBC for massive MIMO systems. The paper illustrates the basis of rateless space time codes deployments in massive MIMO transmissions over wireless erasure channels. In such channels, data may be lost or is not decodable at the receiver due to a variety of factors such as channel fading, interference, or antenna element failure. We show that RSTBC guarantees the reliability of the system in such cases, even when the data loss rate is 25% or more. In such a highly lossy channel, the conventional fixed-rate codes fail to perform well, particularly when channel state information is not available at the transmitter. Simulation results are provided to demonstrate the BER performance and the spectral efficiency of the proposed scheme.
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)
Verification calculations as per CFD FLOWVISION code for sodium-cooled reactor plants
International Nuclear Information System (INIS)
The paper studies the experience in application of CFD FlowVision software for analytical validation of sodium-cooled fast reactor structure components and the results of performed verification, namely: – development and implementation of new model of turbulent heat transfer in liquid sodium (LMS) in FlowVision software and model verification based on thermohydraulic characteristics studied by experiment at TEFLU test facility; – simulation of flowing and mixing of coolant with different temperatures in the upper mixing chamber of fast neutron reactor through the example of BN-600 (comparison with the results obtained at the operating reactor). Based on the analysis of the results obtained, the efficiency of CFD codes application for the considered problems is shown, and the proposals for CFD codes verification development as applied to the advanced sodium-cooled fast reactor designs are stated. (author)
Calculation of anisotropic few-group constants in asymptotic cells: the code ANICELL
International Nuclear Information System (INIS)
The theoretical background of the ANICELL computer program together with a user's manual is presented. ANICELL is a nuclear reactor neutron transport code which solves the traditional asymptotic and the so-called tilted flux transport problems in one-dimensional cylindrical geometry using linearly anisotropic scattering. The method of solution used is the first flight collision probability technique. Few-group constants including radial and axial diffusion coefficients for the cell are also prepared by the program. (author)
Fast neutron reaction data calculations with the computer code STAPRE-H
International Nuclear Information System (INIS)
Description of the specific features of the version STAPRE-H are given. Illustration of the model options and parameter influence on the calculated results is done to trace the accurate reproducing of large body of correlated data. (authors)
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...
The Facial Expression Coding System (FACES): Development, Validation, and Utility
Kring, Ann M.; Sloan, Denise M.
2007-01-01
This article presents information on the development and validation of the Facial Expression Coding System (FACES; A. M. Kring & D. Sloan, 1991). Grounded in a dimensional model of emotion, FACES provides information on the valence (positive, negative) of facial expressive behavior. In 5 studies, reliability and validity data from 13 diverse…
International Nuclear Information System (INIS)
The last 12 years studies about the CABRI, SCARABEE and PHEBUS projects are summarized. It describes the object and the genesis of the cores, the evolution of the core concept and the associated neutronic problems. The calculational scheme used is presented, together with its qualification. The formalism, and the qualification of the different modules of GOLEM are presented. COXYS: module of physical analysis in order to determine the best energetic and spatial mesh for the case of interest. GOLU.B: input data management module. VAREC: calculation module of perturbations due to materials enables to compute perturbed flux and reactivity variation. VARYX: calculation module of geometric perturbations. TRACASYN: module of 3D power shape calculation. Finally TRACASTORE: module of management and graphic exploitation of results. Then, one gives utilization directions for these different modules. Qualification results show that GOLEM is able to analyse the fine physics of many various cases, to calculate by perturbation effects greater than 5000 pcm, to rebuild perturbed flux with margins near 3% for difficult situations, like reactor voiding or spectral or spectral variation in a PWR. Furthermore, 3D hot spots are calculated within margins of a magnitude comparable to experimental ones
International Nuclear Information System (INIS)
The System of Computerized Analysis for Licensing at Atomic industry (SCALA) is a Russian analogue of the well-known SCALE system. For criticality evaluations the ABBN-93 system is used with TWODANT and with joined American KENO and Russian MMK Monte-Carlo code MMKKENO. Using the same cross sections and input models, all these codes give results that coincide within the statistical uncertainties (for Monte-Carlo codes). Validation of criticality calculations using SCALA was performed using data presented in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. Another task of the work was to test the burnup capability of SCALA system in complex geometry in compare with other codes. Benchmark models of VVER type reactor assemblies with UO2 and MOX fuel including the cases with burnable gadolinium absorbers were calculated. KENO-VI and MMK codes were used for power distribution calculations, ORIGEN code was used for the isotopic kinetics calculations. (authors)
Analysis of a 12-Finger Rod Drop using RETRAN/MASTER Code System for APR1400
International Nuclear Information System (INIS)
The Optimized Power Reactor 1000 (OPR1000) has 4-finger and 12-finger Control Element Assemblies (CEAs). When the 12-finger CEA is dropped, Core Protection Calculator System (CPCS) shuts down the reactor to prevent fuel damage that could occur from the sudden reactor power peaking. By contrast, the improved CPCS of Advanced Power Reactor 1400 (APR1400), which has systems similar to those of the OPR1000, decreases reactor power rapidly using its Reactor Power Cutback System (RPCS) to avoid unwanted reactor trips caused by the CPCS during a 12- finger CEA drop event. RETRAN is a best-estimate code for transient analysis of Non-LOCA. The RETRAN control logic, which includes the function of reducing reactor power during a 12-Finger CEA drop, has been developed for the APR1400. A MATRAN program has also been developed. MATRAN is the interface program for realtime processing to connect RETRAN with MASTER code which is a nuclear analysis and design code. MATRAN supplies adequate feedback reactivities from the MASTER code to RETRAN code. The purpose of this study is to analyze the behavior of a nuclear reactor core and its primary system using conventional RETRAN analysis procedure and MATRAN program analysis procedure during a 12- finger CEA drop. In addition, the axial power distribution and Axial Shape Index (ASI) are produced by the MATRAN program and they are confirmed as within operation limits
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.
International Nuclear Information System (INIS)
Computer codes incorporating advanced nuclear models (optical, statistical and pre-equilibrium decay nuclear reaction models) were used to calculate neutron cross sections needed for fusion reactor technology. The elastic and inelastic scattering (n,2n), (n,p), (n,n'p), (n,d) and (n,γ) cross sections for stable molybdenum isotopes Mosup(92,94,95,96,97,98,100) and incident neutron energy from about 100 keV or a threshold to 20 MeV were calculated using the consistent set of input parameters. The hydrogen production cross section which determined the radiation damage in structural materials of fusion reactors can be simply deduced from the presented results. The more elaborated microscopic models of nuclear level density are required for high accuracy calculations
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)
International Nuclear Information System (INIS)
The computer program, TRANCS, has been developed for evaluating the fractional release of long-lived fission products from coated fuel particles. This code numerically gives the non-stationary solution of the diffusion equation with birth and decay terms. The birth term deals with the fissile material in the fuel kernel, the contamination in the coating layers and the fission-recoil transfer from the kernel into the buffer layer; and the decay term deals with effective decay not only due to beta decay but also due to neutron capture, if appropriate input data are given. The code calculates the concentration profile, the release to birth rates (R/B), and the release and residual fractions in the coated fuel particle. Results obtained numerically have been in good agreement with the corresponding analytical solutions after the Booth model. Thus, the validity of the present code was confirmed, and further undate of the code has been discussed for extention of its computation scopes and models. (author)
Optimized load calculation in piping systems based on new experimental data
Energy Technology Data Exchange (ETDEWEB)
Popp, Margit [AREVA NP GmbH, Erlangen (Germany)
2009-07-01
The aim of dynamic fluid calculation is to efficiently design piping systems for power plants. Sudden changes of the fluid velocity, caused by e.g. pump or valve trips produce high pressure surges which exceed many times the operational system pressure. The existing piping codes consider most of the standard load cases satisfactorily. Nevertheless, due to increasing requirements on piping systems and for costreduction, the actual unused potential shall be accessed by calculating the fluid loads more realistically and less conservatively. Especially for the computer simulation of dynamic incidents there is still space for innovation and improvement. Since 2006 the AREVA NP GmbH in cooperation with the University of Erlangen-Nuernberg is investigating pressure surge effects in water-filled piping systems (Figure 1, [1]). By means of the experimental data, the existing piping codes shall be validated to improve the models and to get more realistic loads. (orig.)
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)
WASP: A flexible FORTRAN 4 computer code for calculating water and steam properties
Hendricks, R. C.; Peller, I. C.; Baron, A. K.
1973-01-01
A FORTRAN 4 subprogram, WASP, was developed to calculate the thermodynamic and transport properties of water and steam. The temperature range is from the triple point to 1750 K, and the pressure range is from 0.1 to 100 MN/m2 (1 to 1000 bars) for the thermodynamic properties and to 50 MN/m2 (500 bars) for thermal conductivity and to 80 MN/m2 (800 bars) for viscosity. WASP accepts any two of pressure, temperature, and density as input conditions. In addition, pressure and either entropy or enthalpy are also allowable input variables. This flexibility is especially useful in cycle analysis. The properties available in any combination as output include temperature, density, pressure, entropy, enthalpy, specific heats, sonic velocity, viscosity, thermal conductivity, surface tension, and the Laplace constant. The subroutine structure is modular so that the user can choose only those subroutines necessary to his calculations. Metastable calculations can also be made by using WASP.
PINSPEC. A Monte Carlo code for pin cell spectral calculations for educational applications
International Nuclear Information System (INIS)
Students in many reactor physics courses are exposed to canonical reactor physics concepts through theoretical problems simplified to allow for tractable analytical solutions. Such problems typically require tedious mathematical derivation which is often not the most effective approach to teaching basic reactor physics concepts. A new complementary methodology to introduce these concepts is made possible with PINSPEC, a pin cell Monte Carlo code for educational use. PINSPEC enables students to simulate pin cell models for various reactor types with a simple-to-use Python interface. PINSPEC uses point-wise cross section data and includes a module for Single-Level Breit-Wigner cross-section generation and Doppler broadening. The PINSPEC code supports a variety of tallies which students may use to compute resonance integrals, multi-group cross sections, and more for various materials and pin configurations. PINSPEC is undergoing review for open source release in the near future such that it will be a free and accessible tool for instructors developing reactor physics curricula with an applied and interactive approach to learning. (author)
Reactor pressure vessel strength calculations - comparing the AD/TRD and the ASME code. Pt. 1
International Nuclear Information System (INIS)
The dimensioning criteria applied in the various technical rules are illustrated by the example of a reactor pressure vessel nozzle, especially with a view to the characteristic data of the materials used. Using a detailed finite element analysis of the main coolant nozzle permits an evaluation of the different calculation methods. The second part of the report discusses safety problems, e.g. fatigue analysis, the necessity of carrying out 3D-elastoplastic FE calculations, or the assessment of transient loads on the reactor pressure vessel by means of a fracture-mechanical analysis. (orig./HP)
The three-dimensional PWR transient code ANTI; rod ejection test calculation
International Nuclear Information System (INIS)
ANTI is a computer program being developed for three-dimensional coupled neutronics and thermal-hydraulics description of a PWR core under transient conditions. In this report a test example calculated by the program is described. The test example is a simulation of a control rod ejection from a very small reactor core (to save somputing time). In order to show the influence of cross flow between adjacent fuel elements the same calculation was performed both with the cross flow option and with closed hydraulic channels. (author)
International Nuclear Information System (INIS)
Highlights: • The assessment of RELAP5, TRACE and CATHARE system codes on integral experiments is presented. • Code benchmark of CATHARE, DYN2B, and ATHLET on PHENIX natural circulation experiment. • Grid-free pool modelling based on proper orthogonal decomposition for system codes is explained. • The code coupling methodologies are explained. • The coupling of several CFD/system codes is tested against integral experiments. - Abstract: The THINS project of the 7th Framework EU Program on nuclear fission safety is devoted to the investigation of crosscutting thermal–hydraulic issues for innovative nuclear systems. A significant effort in the project has been dedicated to the qualification and validation of system codes currently employed in thermal–hydraulic transient analysis for nuclear reactors. This assessment is based either on already available experimental data, or on the data provided by test campaigns carried out in the frame of THINS project activities. Data provided by TALL and CIRCE facilities were used in the assessment of system codes for HLM reactors, while the PHENIX ultimate natural circulation test was used as reference for a benchmark exercise among system codes for sodium-cooled reactor applications. In addition, a promising grid-free pool model based on proper orthogonal decomposition is proposed to overcome the limits shown by the thermal–hydraulic system codes in the simulation of pool-type systems. Furthermore, multi-scale system-CFD solutions have been developed and validated for innovative nuclear system applications. For this purpose, data from the PHENIX experiments have been used, and data are provided by the tests conducted with new configuration of the TALL-3D facility, which accommodates a 3D test section within the primary circuit. The TALL-3D measurements are currently used for the validation of the coupling between system and CFD codes
Challenges in calculating molecular systems with Coulomb interactions
Kirnosov, Nikita; Sharkey, Keeper; Adamowicz, Ludwik
2014-03-01
The highly accurate quantum mechanical calculations are not only crucial for high-resolution experimental data verification, but may also serve as a guide in the field of exotic systems exploration. Including all non-relativistic effects in a single-step variational approach and rigorously separating out the center of mass motion allows us to build a reliable model for calculating bound states of molecular systems with Coulomb interactions. In these calculations the wave function of the system is expanded in terms of explicitly correlated Gaussian (ECG) basis functions. Examples of calculations of energies and other properties of some molecular systems will be presented.
Zhang, X; Zhang, Xiao-he; Sutherland, Peter
1993-01-01
A new, fully dynamic and self-consistent radiation hydrodynamics code, suitable for the calculation of supernovae light curves and continuum spectra, is described. It is a multigroup (frequency-dependent) code and includes all important $O(v/c)$ effects. It is applied to the model W7 of Nomoto, Thielemann, \\& Yokoi (1984) for supernovae of type Ia. Radioactive energy deposition is incorporated through use of tables based upon Monte Carlo results. Effects of line opacity (both static or line blanketing and expansion or line blocking) are neglected, although these may prove to be important. At maximum light, models based upon different treatments of the opacity lead to values for $M_{B,max}$ in the range of -19.0 to -19.4. This range falls between the values for observed supernova claimed by Leibundgut \\& Tammann (1990) and by Pierce, Ressler, \\& Shure (1992).
International Nuclear Information System (INIS)
The ORION code is designed to determine very quickly the immediate consequences (such as plume passage time, instantaneous maximum hazards irradiation, inhalation, deposit) due to an accident spreading out radioactive or chemical pollution into the atmosphere, from a source point, a stack release, (with heightening calculation) outspread sources (transport accident such as, for instance, road fire or car crash) or from a cylindrical cloud defined by different vertical sources (for instance pyrotechnical accident, missile firing...). The diffusion code DOURY type (french official methods) is written in FORTRAN. Data are entered in a conversational mode with auto-checking. Results are output to tables an isorisks curves drawn at map scales. At the Bruyeres-le-Chatel Radiation Protection Unit, a team is on permanent duty, can carry out results in a few minutes and transmit the evaluation by TELEFAX anywhere on the National territory
An Algorithm for Constructing All Families of Codes of Arbitrary Requirement in an OCDMA System
Lu, Xiang; Chen, Jiajia; He, Sailing
2006-01-01
A novel code construction algorithm is presented to find all the possible code families for code reconfiguration in an OCDMA system. The algorithm is developed through searching all the complete subgraphs of a constructed graph. The proposed algorithm is flexible and practical for constructing optical orthogonal codes (OOCs) of arbitrary requirement. Simulation results show that one should choose an appropriate code length in order to obtain sufficient number of code families for code reconfi...
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...
Turkington, M. D.; Ballance, C. P.; Hibbert, A.; Ramsbottom, C. A.
2016-08-01
In this work we explore the validity of employing a modified version of the nonrelativistic structure code civ3 for heavy, highly charged systems, using Na-like tungsten as a simple benchmark. Consequently, we present radiative and subsequent collisional atomic data compared with corresponding results from a fully relativistic structure and collisional model. Our motivation for this line of study is to benchmark civ3 against the relativistic grasp0 structure code. This is an important study as civ3 wave functions in nonrelativistic R -matrix calculations are computationally less expensive than their Dirac counterparts. There are very few existing data for the W LXIV ion in the literature with which we can compare except for an incomplete set of energy levels available from the NIST database. The overall accuracy of the present results is thus determined by the comparison between the civ3 and grasp0 structure codes alongside collisional atomic data computed by the R -matrix Breit-Pauli and Dirac codes. It is found that the electron-impact collision strengths and effective collision strengths computed by these differing methods are in good general agreement for the majority of the transitions considered, across a broad range of electron temperatures.
International Nuclear Information System (INIS)
This thesis concerns the definition and the validation of the ERANOS neutronic calculation system for steel reflected fast reactors. The calculation system uses JEF2.2 evaluated nuclear data, the ECCO cell code and the BISTRO and VARIANT transport codes. After a description of the physical phenomena induced by the existence of the these sub-critical media, an inventory of the past studies related to steel reflectors is reported. A calculational scheme taking into account the important physical phenomena (strong neutronic slowing-down, presence of broad resonances of the structural materials and spatial variation of the spectrum in the reflector) is defined. This method is validated with the TRIPOLI4 reference Monte-Carlo code. The use of this upgraded calculation method for the analysis of the part of the CIRANO experimental program devoted to the study of steel reflected configurations leads to discrepancies between the calculated and measured values. These remaining discrepancies obtained for the reactivity and the fission rate traverses are due to inaccurate nuclear data for the structural materials. The adjustment of these nuclear data in order to reduce these discrepancies id demonstrated. The additional uncertainty associated to the integral parameters of interest for a nuclear reactor (reactivity and power distribution) induced by the replacement of a fertile blanket by a steel reflector is determined for the Superphenix reactor and is proved to be small. (author)
Upgrades to the NESS (Nuclear Engine System Simulation) Code
Fittje, James E.
2007-01-01
In support of the President's Vision for Space Exploration, the Nuclear Thermal Rocket (NTR) concept is being evaluated as a potential propulsion technology for human expeditions to the moon and Mars. The need for exceptional propulsion system performance in these missions has been documented in numerous studies, and was the primary focus of a considerable effort undertaken during the 1960's and 1970's. The NASA Glenn Research Center is leveraging this past NTR investment in their vehicle concepts and mission analysis studies with the aid of the Nuclear Engine System Simulation (NESS) code. This paper presents the additional capabilities and upgrades made to this code in order to perform higher fidelity NTR propulsion system analysis and design.
Phase-Space Analysis of Wavefront Coding Imaging Systems
Institute of Scientific and Technical Information of China (English)
YANG Qing-Guo; SUN Jian-Feng; LIU Li-Ren
2006-01-01
@@ We explore the use of the Radon-Wigner transform, which is associated with the fractional Fourier transform of the pupil function, for determining the point spread function (PSF) of an incoherent defocused optical system.Then we introduce these phase-space tools to analyse the wavefront coding imaging system. It is shown that the shape of the PSF for such a system is highly invariant to the defocus-related aberrations except for a lateral shift.The optical transfer function of this system is also investigated briefly from a new understanding of ambiguity function.
Applying Hamming Code to Memory System of Safety Grade PLC (POSAFE-Q) Processor Module
International Nuclear Information System (INIS)
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
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)
Development of a multi-group SN transport calculation code with unstructured tetrahedral meshes
International Nuclear Information System (INIS)
This paper reviews the computational methods used in the MUST (Multi-group Unstructured geometry SN Transport) code for solving the multi-group Sn transport equation in general geometries and describes the status of development of MUST. MUST solves the multi-group transport equation with unstructured tetrahedral meshes for modeling complicated geometrical problems. For tetrahedral mesh generation, input generation, and output visualization, we developed a management program where the mesh generation is based on Gmsh and TetGen that are open softwares. The geometrical modeling is done with the commercial CAD softwares such as CATIA. MUST uses the discontinuous finite element method (DFEM) and two-sub cell balance methods with linear discontinuous expansion (LDEM-SCB) to spatially discretize the transport equation. We applied MUST to three neutron and gamma coupled test problems for testing MUST. (author)
Shielding design calculations for ESS activated target system
International Nuclear Information System (INIS)
Full text of publication follows: The European Spallation Source (ESS) is a European common effort in designing and building a next generation large-scale user facility for studies of the structure and dynamics of materials. The ESS target, moderators and reflectors system through interactions with 5 MW proton beam (2.5 GeV, 20 Hz) will produce long pulse (2.8 ms width) neutrons in sub-thermal and thermal energy range. These neutrons are further transported to a variety of neutron scattering instruments. The aim of this work is to assess the strategy to be used for the safe handling and shipping of the ESS target and associated shaft. For safe maintenance, during operation as well as handling, transport and storage of the components of the ESS target station after their lifetime, detailed knowledge is required about the activation induced by the impinging protons and secondary radiation fields. The Monte Carlo transport code MCNPX2.6.0 was coupled with CINDER90 version 07.4 to calculate the residual nuclide production in the target wheel and associated shaft. Dose equivalent rates due to the residual radiation were further calculated with the MICROSHIELD and MCNPX codes using photon sources resulting from CINDER. Various decay times after ceasing operation of the target components were considered. The activation and decay heat density distributions of the target system together with the derived dose rates were analysed to assess the best strategy to be used for their safe removal and transport to a hot cell, eventual dismantling, storage on-site and shipping off-site as intermediate level waste packages. The derived photon sources were used afterwards to design the shielded exchange flasks that are needed to remove and transport the target after its lifespan to a hot cell. Design of a multipurpose cask able to accommodate the different highly activated components of the ESS target station and ship them to external conditioning facility is intended to be developed
International Nuclear Information System (INIS)
The radioactivity induced in PWR cluster control rods during reactor operation has been calculated using the computer programme ORIGEN. Neutron fluxes and spectrum conditions as well as the strongly shielded cross sections for the absorber materials Ag, In and Cd have been obtained by running the cell and assembly code CASMO for a couple of typical cases. The results show that Ag-110m, Fe-55 and Co-60 give the largest activity contributions in the interval 1-10 years after the end of irradiation, and Ni-63 and Cd-113m in a longer time perspective. (author)
CAPACITY CALCULATION OF TD-CDMA HIGH ALTITUDE PLATFORM SYSTEM
Institute of Scientific and Technical Information of China (English)
Wang Zhenyong; Liu Xiaowei; Li Zhuoshi
2011-01-01
A capacity calculation method of High Altitude Platform System (HAPS) is proposed in which TD-CDMA multiple access schemes are applied.With the influence of both power limit and bandwidth limit on capacity integrated,the paper derives the equations by which the capacity of TD-CDMA systems can be calculated,and performs calculation on a practical system.This calculation method is quite simple and effective with a comparatively small error,which is essential to the designing and research on HAPS.
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.
Impact of Different Spreading Codes Using FEC on DWT Based MC-CDMA System
Masum, Saleh; Kabir, M. Hasnat; Islam, Md. Matiqul; Shams, Rifat Ara; Ullah, Shaikh Enayet
2012-01-01
The effect of different spreading codes in DWT based MC-CDMA wireless communication system is investigated. In this paper, we present the Bit Error Rate (BER) performance of different spreading codes (Walsh-Hadamard code, Orthogonal gold code and Golay complementary sequences) using Forward Error Correction (FEC) of the proposed system. The data is analyzed and is compared among different spreading codes in both coded and uncoded cases. It is found via computer simulation that the performance...
Energy Technology Data Exchange (ETDEWEB)
Wulff, W; Cheng, H S; Diamond, D J; Khatib-Rahbar, M
1984-01-01
This report documents the physical models and the numerical methods employed in the BWR systems code RAMONA-3B. The RAMONA-3B code simulates three-dimensional neutron kinetics and multichannel core hydraulics of nonhomogeneous, nonequilibrium two-phase flows. RAMONA-3B is programmed to calculate the steady and transient conditions in the main steam supply system for normal and abnormal operational transients, including the performances of plant control and protection systems. Presented are code capabilities and limitations, models and solution techniques, the results of development code assessment and suggestions for improving the code in the future.
International Nuclear Information System (INIS)
This report documents the physical models and the numerical methods employed in the BWR systems code RAMONA-3B. The RAMONA-3B code simulates three-dimensional neutron kinetics and multichannel core hydraulics of nonhomogeneous, nonequilibrium two-phase flows. RAMONA-3B is programmed to calculate the steady and transient conditions in the main steam supply system for normal and abnormal operational transients, including the performances of plant control and protection systems. Presented are code capabilities and limitations, models and solution techniques, the results of development code assessment and suggestions for improving the code in the future
Diffusion Monte Carlo calculations of three-body systems
Institute of Scientific and Technical Information of China (English)
L(U) Meng-Jiao; REN Zhong-Zhou; LIN Qi-Hu
2012-01-01
The application of the diffusion Monte Carlo algorithm in three-body systems is studied.We develop a program and use it to calculate the property of various three-body systems.Regular Coulomb systems such as atoms,molecules,and ions are investigated.The calculation is then extended to exotic systems where electrons are replaced by muons.Some nuclei with neutron halos are also calculated as three-body systems consisting of a core and two external nucleons.Our results agree well with experiments and others' work.
Considerations on the calculation of volumes in two planning systems
International Nuclear Information System (INIS)
The discrepancies in the calculation of the same volume between different planning systems impact on dose-volume histograms and therefore clinical assessment of dosimetry for patients. The transfer, by a local network, tomographic study (CT) and contours of critical organs of patients, between our two planning systems allows us to evaluate the calculation of identical volumes.
TMRBAR: a code to calculate plasma parameters for tandem-mirror reactors operating in the MARS mode
Energy Technology Data Exchange (ETDEWEB)
Campbell, R.B.
1983-08-30
The purpose of this report is to document the plasma power balance model currently used by LLNL to calculate steady state operating points for tandem mirror reactors. The code developed from this model, TMRBAR, has been used to predict the performance and define supplementary heating requirements for drivers used in the Mirror Advanced Reactor Study (MARS) and for the Fusion Power Demonstration (FPD) study. The equations solved included particle and energy balance for central cell and end cell species, quasineutrality at several cardinal points in the end cell region, as well as calculations of volumes, densities and average energies based on given constraints of beta profiles and fusion power output. Alpha particle ash is treated self-consistently, but no other impurity species is treated.
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
PyVCI: A flexible open-source code for calculating accurate molecular infrared spectra
Sibaev, Marat; Crittenden, Deborah L.
2016-06-01
The PyVCI program package is a general purpose open-source code for simulating accurate molecular spectra, based upon force field expansions of the potential energy surface in normal mode coordinates. It includes harmonic normal coordinate analysis and vibrational configuration interaction (VCI) algorithms, implemented primarily in Python for accessibility but with time-consuming routines written in C. Coriolis coupling terms may be optionally included in the vibrational Hamiltonian. Non-negligible VCI matrix elements are stored in sparse matrix format to alleviate the diagonalization problem. CPU and memory requirements may be further controlled by algorithmic choices and/or numerical screening procedures, and recommended values are established by benchmarking using a test set of 44 molecules for which accurate analytical potential energy surfaces are available. Force fields in normal mode coordinates are obtained from the PyPES library of high quality analytical potential energy surfaces (to 6th order) or by numerical differentiation of analytic second derivatives generated using the GAMESS quantum chemical program package (to 4th order).
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.
Neural map formation and sensory coding in the vomeronasal system.
Brignall, Alexandra C; Cloutier, Jean-François
2015-12-01
Sensory systems enable us to encode a clear representation of our environment in the nervous system by spatially organizing sensory stimuli being received. The organization of neural circuitry to form a map of sensory activation is critical for the interpretation of these sensory stimuli. In rodents, social communication relies strongly on the detection of chemosignals by the vomeronasal system, which regulates a wide array of behaviours, including mate recognition, reproduction, and aggression. The binding of these chemosignals to receptors on vomeronasal sensory neurons leads to activation of second-order neurons within glomeruli of the accessory olfactory bulb. Here, vomeronasal receptor activation by a stimulus is organized into maps of glomerular activation that represent phenotypic qualities of the stimuli detected. Genetic, electrophysiological and imaging studies have shed light on the principles underlying cell connectivity and sensory map formation in the vomeronasal system, and have revealed important differences in sensory coding between the vomeronasal and main olfactory system. In this review, we summarize the key factors and mechanisms that dictate circuit formation and sensory coding logic in the vomeronasal system, emphasizing differences with the main olfactory system. Furthermore, we discuss how detection of chemosignals by the vomeronasal system regulates social behaviour in mice, specifically aggression. PMID:26329476
Pairwise codeword error probability for coded atmospheric optical communication systems
Institute of Scientific and Technical Information of China (English)
HAN Jia-jia; RONG Jian; ZHONG Xiao-chun
2006-01-01
To study the performance of various error-control coding schemes,exact expressions and upper bounds on the pairwise codeword error probability(PEP)for several modulation schemes(OOK,SC-BPSK,BPPM)used in atmospheric optical communication systems are derived.To simplify the computation,this research was under the assumption of weak turbulence.Moreover,by simulation of expressions,the performances of PEP in different modulation schemes are compared and the best one of them is given.
International Nuclear Information System (INIS)
To justify the use of a commercial Computational Fluid Dynamics (CFD) code for a CANDU fuel channel analysis, especially for the radiation heat transfer dominant conditions, the CFX-10 code is tested against three benchmark problems which were used for the validation of a radiation heat transfer in the CANDU analysis code, a CATHENA. These three benchmark problems are representative of the CANDU fuel channel configurations from a simple geometry to a whole fuel channel geometry. For the solutions of the benchmark problems, the temperature or the net radiation heat flux boundary conditions are prescribed for each radiating surface to determine the radiation heat transfer rate or the surface temperature, respectively by using the network method. The Discrete Transfer Model (DTM) is used for the CFX-10 radiation model and its calculation results are compared with the solutions of the benchmark problems. The CFX-10 results for the three benchmark problems are in close agreement with those solutions, so it is concluded that the CFX-10 with a DTM radiation model can be applied to the CANDU fuel channel analysis where a surface radiation heat transfer is a dominant mode of the heat transfer. (author)
Development of a computer code for shielding calculation in X-ray facilities
International Nuclear Information System (INIS)
The construction of an effective barrier against the interaction of ionizing radiation present in X-ray rooms requires consideration of many variables. The methodology used for specifying the thickness of primary and secondary shielding of an traditional X-ray room considers the following factors: factor of use, occupational factor, distance between the source and the wall, workload, Kerma in the air and distance between the patient and the receptor. With these data it was possible the development of a computer program in order to identify and use variables in functions obtained through graphics regressions offered by NCRP Report-147 (Structural Shielding Design for Medical X-Ray Imaging Facilities) for the calculation of shielding of the room walls as well as the wall of the darkroom and adjacent areas. With the built methodology, a program validation is done through comparing results with a base case provided by that report. The thickness of the obtained values comprise various materials such as steel, wood and concrete. After validation is made an application in a real case of radiographic room. His visual construction is done with the help of software used in modeling of indoor and outdoor. The construction of barriers for calculating program resulted in a user-friendly tool for planning radiographic rooms to comply with the limits established by CNEN-NN-3:01 published in September / 2011
Research of Wavelet Based Multicarrier Modulation System with Near Shannon Limited Codes
Institute of Scientific and Technical Information of China (English)
ZHANGHaixia; YUANDongfeng; ZHAOFeng
2005-01-01
In this paper, by using turbo codes and Low density parity codes (LDPC) as channel correcting code scheme, Wavelet based multicarrier modulation (WMCM) systems are proposed and investigated on different transmission scenarios. The Bit error rate (BER) performance of these two near Shannon limited codes is simulated and compared with various code parameters. Simulated results show that Turbo coded WMCM (TCWMCM) performs better than LDPC coded WMCM (LDPC-CWMCM) on both AWGN and Rayleigh fading channels when these two kinds of codes are of the same code parameters.
International Nuclear Information System (INIS)
The HARAD computer code, written in FORTRAN IV, calculates concentrations of radioactive daughters in air following the atmospheric release of a parent radionuclide under a variety of meteorological conditions. It can be applied most profitably to the assessment of doses to man from the noble gases such as 222Rn, 220Rn, and Xe and Kr isotopes. These gases can produce significant quantities of short-lived particulate daughters in an airborne plume, which are the major contributors to dose from these chains with gaseous parent radionuclides. The simultaneous processes of radioactive decay, buildup, and environmental losses through wet and dry deposition on ground surfaces are calculated for a daughter chain in an airborne plume as it is dispersed downwind from a point of release of a parent. The code employs exact solutions of the differential equations describing the above processes over successive discrete segments of downwind distance. Average values for the dry deposition coefficients of the chain members over each of these distance segments were treated as constants in the equations. The advantage of HARAD is its short computing time
International Nuclear Information System (INIS)
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)
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.
Evaluation of ANGLE(R), a code for calculating HPGe detector efficiencies
Energy Technology Data Exchange (ETDEWEB)
Homan, Victoria M [Los Alamos National Laboratory
2010-10-25
This paper evaluates the ANGLE(reg sign) software package, an advanced efficiency calibration software for high purity germanium detectors that is distributed by ORTEC(reg sign). ANGLE(reg sign) uses a semi-empirical approach, by way of the efficiency transfer method, based on the calculated effective solid angle. This approach would have an advantage over the traditional relative and stochastic methods by decreasing the chances for systematic errors and reducing sensitivity to uncertainties in detector parameters. For experimental confirmation, a closed-end coaxial HPGe detector was used with sample geometries frequently encountered at the Los Alamos National Laboratory. The results obtained were sufficient for detector-source configurations which included intercepting layers of plexiglass and carbon graphite, but somewhat insufficient for bare source configurations.
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)
International Nuclear Information System (INIS)
This work aims to calculate the fluence to effective dose conversion coefficients, (E/Φ), for monoenergetic neutrons from 10-9 to 20 MeV, based on the radiation (wR) and tissue (wT) weighting factors values recommended by ICRP publications numbers 60 and 103. The organs and tissues absorbed doses were calculated using the radiation transport code MCNPX and a female anthropomorphic voxel-based simulator, assuming whole-body irradiation by plane-parallel beams, on the geometries of the antero-posterior (AP) and postero-anterior (PA) irradiation. Dose calculations were performed for 21 selected organs of the body, for which the International Commission on Radiological Protection and the International Commission on Radiological Units and Measurements have set tissue weighting factors for the determination of the effective dose. From comparison between the dose results calculated and the data reported for the MIRD model, it can be concluded that, the fluence to effective dose conversion coefficients obtained using the voxel simulator are underestimated by a factor of up to 5 times when compared with the one obtained by ICRP 74, using mathematical simulators. (author)
The ICPC coding system in pharmacy : developing a subset, ICPC-Ph
van Mil, JWF; Brenninkmeijer, R; Tromp, TFJ
1998-01-01
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
International Nuclear Information System (INIS)
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
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.
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)
Advanced coding techniques for few mode transmission systems.
Okonkwo, Chigo; van Uden, Roy; Chen, Haoshuo; de Waardt, Huug; Koonen, Ton
2015-01-26
We experimentally verify the advantage of employing advanced coding schemes such as space-time coding and 4 dimensional modulation formats to enhance the transmission performance of a 3-mode transmission system. The performance gain of space-time block codes for extending the optical signal-to-noise ratio tolerance in multiple-input multiple-output optical coherent spatial division multiplexing transmission systems with respect to single-mode transmission performance are evaluated. By exploiting the spatial diversity that few-mode-fibers offer, with respect to single mode fiber back-to-back performance, significant OSNR gains of 3.2, 4.1, 4.9, and 6.8 dB at the hard-decision forward error correcting limit are demonstrated for DP-QPSK 8, 16 and 32 QAM, respectively. Furthermore, by employing 4D constellations, 6 × 28Gbaud 128 set partitioned quadrature amplitude modulation is shown to outperform conventional 8 QAM transmission performance, whilst carrying an additional 0.5 bit/symbol.
Research and implementation of flexible coding system oriented multi-view
Institute of Scientific and Technical Information of China (English)
ZHANG Xuhui; ZHANG Xu; NING Ruxin
2007-01-01
On the basis of the requirements of a product data management system (PDM) for the flexible coding system,the principle of the flexible coding system oriented multiview is analyzed. Generation and utilization of coding should be associated with the context of the object. The architecture of the flexible coding system oriented multi-view is studied and the implementation class diagram of the system is designed. The system can support the establishment of five types of code segments, provide the tools of flexible defining coding rules and drive the automatic generation of object coding in different views (contexts). On the foundation of the characteristics of the system, coding for parts is taken as a sample to validate and elaborate the flexible coding process of the system.
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....
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.
A study on the nuclear computer code maintenance and management system
International Nuclear Information System (INIS)
According to current software development and quality assurance trends. It is necessary to develop computer code management system for nuclear programs. For this reason, the project started in 1987. Main objectives of the project are to establish a nuclear computer code management system, to secure software reliability, and to develop nuclear computer code packages. Contents of performing the project in this year were to operate and maintain computer code information system of KAERI computer codes, to develop application tool, AUTO-i, for solving the 1st and 2nd moments of inertia on polygon or circle, and to research nuclear computer code conversion between different machines. For better supporting the nuclear code availability and reliability, assistance from users who are using codes is required. Lastly, for easy reference about the codes information, we presented list of code names and information on the codes which were introduced or developed during this year. (Author)
Towards Fully Converged GW Calculations for Large Systems
Gao, Weiwei; Gao, Xiang; Zhang, Peihong
2016-01-01
Although the GW approximation is recognized as one of the most accurate theories for predicting materials excited states properties, scaling up conventional GW calculations for large systems remains a major challenge. We present a powerful and simple-to-implement method that can drastically accelerate fully converged GW calculations for large systems. We demonstrate the performance of this new method by calculating the quasiparticle band gap of MgO supercells. A speed-up factor of nearly two orders of magnitude is achieved for a system contaning 256 atoms (1024 velence electrons) with a negligibly small numerical error of $\\pm 0.03$ eV.
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.
Study on a new meteorological sampling scheme developed for the OSCAAR code system
Energy Technology Data Exchange (ETDEWEB)
Liu Xinhe; Tomita, Kenichi; Homma, Toshimitsu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2002-03-01
One important step in Level-3 Probabilistic Safety Assessment is meteorological sequence sampling, on which the previous studies were mainly related to code systems using the straight-line plume model and more efforts are needed for those using the trajectory puff model such as the OSCAAR code system. This report describes the development of a new meteorological sampling scheme for the OSCAAR code system that explicitly considers population distribution. A group of principles set for the development of this new sampling scheme includes completeness, appropriate stratification, optimum allocation, practicability and so on. In this report, discussions are made about the procedures of the new sampling scheme and its application. The calculation results illustrate that although it is quite difficult to optimize stratification of meteorological sequences based on a few environmental parameters the new scheme do gather the most inverse conditions in a single subset of meteorological sequences. The size of this subset may be as small as a few dozens, so that the tail of a complementary cumulative distribution function is possible to remain relatively static in different trials of the probabilistic consequence assessment code. (author)
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.
Energy Technology Data Exchange (ETDEWEB)
Joo, Hyung Kook; Noh, Jae Man; Lee, Hyung Chul; Yoo, Jae Woon
2006-01-15
In this report, we verified the NUREC code transient calculation capability using OECD NEA/US NRC PWR MOX/UO2 Core Transient Benchmark Problem. The benchmark problem consists of Part 1, a 2-D problem with given T/H conditions, Part 2, a 3-D problem at HFP condition, Part 3, a 3-D problem at HZP condition, and Part 4, a transient state initiated by a control rod ejection at HZP condition in Part 3. In Part 1, the results of NUREC code agreed well with the reference solution obtained from DeCART calculation except for the pin power distributions at the rodded assemblies. In Part 2, the results of NUREC code agreed well with the reference DeCART solutions. In Part 3, some results of NUREC code such as critical boron concentration and core averaged delayed neutron fraction agreed well with the reference PARCS 2G solutions. But the error of the assembly power at the core center was quite large. The pin power errors of NUREC code at the rodded assemblies was much smaller the those of PARCS code. The axial power distribution also agreed well with the reference solution. In Part 4, the results of NUREC code agreed well with those of PARCS 2G code which was taken as the reference solution. From the above results we can conclude that the results of NUREC code for steady states and transient states of the MOX loaded LWR core agree well with those of the other codes.
Energy Technology Data Exchange (ETDEWEB)
Easter, M.E.
1985-07-01
The SCALE code system, utilizing the Monte Carlo computer code KENO V.a, was employed to calculate 37 critical experiments. The critical assemblies had /sup 235/U enrichments of 5% or less and cover a variety of geometries and materials. Values of k/sub eff/ were calculated using two different results using either of the cross-section libraries. The 16-energy-group Hansen-Roach and the 27-energy-group ENDF/B-IV cross-section libraries, available in SCALE, were used in this validation study, and both give good results for the experiments considered. It is concluded that the code and cross sections are adequate for low-enriched uranium systems and that reliable criticality safety calculations can be made for such systems provided the limits of validated applicability are not exceeded.
Applicability of the SCALE code system to MOX fuel transport systems for criticality safety analysis
Energy Technology Data Exchange (ETDEWEB)
Yamamoto, Toshihiro; Naito, Yoshitaka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Hayashi, Toshiaki; Takasugi, Masahiro; Natsume, Toshihiro; Tsuda, Kazuaki
1996-11-01
In order to ascertain feasibilities of the SCALE code system for MOX fuel transport systems, criticality analyses were performed for MOX fuel (Pu enrichment; 3.0 wt.%) criticality experiments at JAERI`s TCA and for infinite fuel rod arrays as parameters of Pu enrichment and lattice pitch. The comparison with a combination of the continuous energy Monte Carlo code MCNP and JENDL-3.2 indicated that the SCALE code system with GAM-THERMOS 123-group library can produce feasible results. Though HANSEN-ROACH 16-group library gives poorer results for MOS fuel transport systems, the errors are conservative except for high enriched fuels. (author)
International Nuclear Information System (INIS)
Implementation of the standardised cost structure, as defined in 'A Proposed Standardised List of Costs Items for Decommissioning Purposes' (OECD/NEA, IAEA, EC, 1999), into the decommissioning costing, supports the harmonisation of decommissioning costs. The decision making processes in decommissioning planning can be more effective if there is the possibility to compare the calculated data with the data of other projects, structured in standardised cost structure. The results of the decision making process should be based on evaluation of such a set of decommissioning options which covers the methods of decommissioning, the selected strategy and existing or planned decommissioning infrastructure. Aspects such as impact of time, waste management scenarios, uncertainties of input data and other aspects should be also evaluated. These issues of decision making process were implemented into the decommissioning costing code OMEGA. All activities of a decommissioning project are involved within single compact standardised calculation structure including waste management. The resulting costs have standardised format and no additional data conversion is needed. The calculation process is nuclide resolved and internally linked in such a way that it models the material and radioactivity flow in the decommissioning process. The effect of decay of radioactivity is considered. The options are optimised in the standard MS-Project software as Gantt charts. The bi-directional data link between the standardised calculation structure and the Gantt chart supports the on-line optimisation of the Gantt chart structure. Multi-option work is applied, i.e. decommissioning options, which cover all decommissioning scenarios to be considered, are evaluated individually and multi-attribute analysis is applied for selecting the optimal one. Methods of sensitivity analysis and evaluation of uncertainties of calculated costs were developed for support the decision making process and for
On the use of WIMS-7 for calculations on accelerator-driven systems
Energy Technology Data Exchange (ETDEWEB)
De Kruijff, W.J.M.; Freudenreich, W.J.M
1998-02-01
The WIMS-7 code package has successfully been applied for a simple benchmark of a lead-cooled accelerator-driven system (ADS). With WIMS-7 it is possible to model a fixed source and to calculate the multiplication in a subcritical system. The calculations have shown that WIMS-7 is capable of treating this benchmark of a homogenized lead-cooled system with a fast neutron spectrum. The results described in this report are very promising and stimulate further investigation of WIMS-7 to study ADS-applications and lead-cooled reactor cores. It is useful to have a more extensive validation of WIMS-7 for lead-cooled ADS. In this report we have only considered a simple homogenized system. In the near future the application of WIMS-7 will be twofold. First, WIMS-7 can be applied to calculate the neutron spectrum in an accelerator-driven system in order to perform transmutation studies with a burnup code. Second, WIMS-7 can be used to study in more detail the neutronics of accelerator-driven systems. This is useful in order to learn more about the physics of accelerator-driven systems. 6 refs.
On the use of WIMS-7 for calculations on accelerator-driven systems
International Nuclear Information System (INIS)
The WIMS-7 code package has successfully been applied for a simple benchmark of a lead-cooled accelerator-driven system (ADS). With WIMS-7 it is possible to model a fixed source and to calculate the multiplication in a subcritical system. The calculations have shown that WIMS-7 is capable of treating this benchmark of a homogenized lead-cooled system with a fast neutron spectrum. The results described in this report are very promising and stimulate further investigation of WIMS-7 to study ADS-applications and lead-cooled reactor cores. It is useful to have a more extensive validation of WIMS-7 for lead-cooled ADS. In this report we have only considered a simple homogenized system. In the near future the application of WIMS-7 will be twofold. First, WIMS-7 can be applied to calculate the neutron spectrum in an accelerator-driven system in order to perform transmutation studies with a burnup code. Second, WIMS-7 can be used to study in more detail the neutronics of accelerator-driven systems. This is useful in order to learn more about the physics of accelerator-driven systems. 6 refs
International Nuclear Information System (INIS)
The MAX phantom has been developed from existing segmented images of a male adult body, in order to achieve a representation as close as possible to the anatomical properties of the reference adult male specified by the ICRP. In computational dosimetry, MAX can simulate the geometry of a human body under exposure to ionizing radiations, internal or external, with the objective of calculating the equivalent dose in organs and tissues for occupational, medical or environmental purposes of the radiation protection. This study presents a methodology used to build a new computational exposure model MAX/EGS4: the geometric construction of the phantom; the development of the algorithm of one-directional, divergent, and isotropic radioactive sources; new methods for calculating the equivalent dose in the red bone marrow and in the skin, and the coupling of the MAX phantom with the EGS4 Monte Carlo code. Finally, some results of radiation protection, in the form of conversion coefficients between equivalent dose (or effective dose) and free air-kerma for external photon irradiation are presented and discussed. Comparing the results presented with similar data from other human phantoms it is possible to conclude that the coupling MAX/EGS4 is satisfactory for the calculation of the equivalent dose in radiation protection. (author)
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...
Aggarwal, K M; Lawson, K D
2016-01-01
There have been discussions in the recent literature regarding the accuracy of the available electron impact excitation rates (equivalently effective collision strengths $\\Upsilon$) for transitions in Be-like ions. In the present paper we demonstrate, once again, that earlier results for $\\Upsilon$ are indeed overestimated (by up to four orders of magnitude), for over 40\\% of transitions and over a wide range of temperatures. To do this we have performed two sets of calculations for N~IV, with two different model sizes consisting of 166 and 238 fine-structure energy levels. As in our previous work, for the determination of atomic structure the GRASP (General-purpose Relativistic Atomic Structure Package) is adopted and for the scattering calculations (the standard and parallelised versions of) the Dirac Atomic R-matrix Code ({\\sc darc}) are employed. Calculations for collision strengths and effective collision strengths have been performed over a wide range of energy (up to 45~Ryd) and temperature (up to 2.0$...
Aggarwal, K. M.; Keenan, F. P.; Lawson, K. D.
2016-10-01
There have been discussions in the recent literature regarding the accuracy of the available electron impact excitation rates (equivalently effective collision strengths Υ) for transitions in Be-like ions. In the present paper we demonstrate, once again, that earlier results for Υ are indeed overestimated (by up to four orders of magnitude), for over 40 per cent of transitions and over a wide range of temperatures. To do this we have performed two sets of calculations for N IV, with two different model sizes consisting of 166 and 238 fine-structure energy levels. As in our previous work, for the determination of atomic structure the GRASP (General-purpose Relativistic Atomic Structure Package) is adopted and for the scattering calculations (the standard and parallelised versions of) the Dirac Atomic R-matrix Code (DARC) are employed. Calculations for collision strengths and effective collision strengths have been performed over a wide range of energy (up to 45 Ryd) and temperature (up to 2.0 × 106 K), useful for applications in a variety of plasmas. Corresponding results for energy levels, lifetimes and A-values for all E1, E2, M1 and M2 transitions among 238 levels of N IV are also reported.
New Parallel Interference Cancellation for Convolutionally Coded CDMA Systems
Institute of Scientific and Technical Information of China (English)
Xu Guo-xiong; Gan Liang-cai; Huang Tian-xi
2004-01-01
Based on BCJR algorithm proposed by Bahl et al and linear soft decision feedback, a reduced-complexity parallel interference cancellation (simplified PIC) for convolutionally coded DS CDMA systems is proposed. By computer simulation, we compare the simplified PIC with the exact PIC. It shows that the simplified PIC can achieve the performance close to the exact PIC if the mean values of coded symbols are linearly computed in terms of the sum of initial a prior log-likelihood rate (LLR) and updated a prior LLR, while a significant performance loss will occur if the mean values of coded symbols are linearly computed in terms of the updated a prior LLR only. Meanwhile, we also compare the simplified PIC with MF receiver and conventional PICs. The simulation results show that the simplified PIC dominantly outperforms the MF receiver and conventional PICs, at signal-noise rate (SNR) of 7 dB, for example, the bit error rate is about 10-4 for the simplified PIC, which is far below that of matched-filter receiver and conventional PIC.
System Design Considerations In Bar-Code Laser Scanning
Barkan, Eric; Swartz, Jerome
1984-08-01
The unified transfer function approach to the design of laser barcode scanner signal acquisition hardware is considered. The treatment of seemingly disparate system areas such as the optical train, the scanning spot, the electrical filter circuits, the effects of noise, and printing errors is presented using linear systems theory. Such important issues as determination of depth of modulation, filter specification, tolerancing of optical components, and optimi-zation of system performance in the presence of noise are discussed. The concept of effective spot size to allow for impact of optical system and analog processing circuitry upon depth of modulation is introduced. Considerations are limited primarily to Gaussian spot profiles, but also apply to more general cases. Attention is paid to realistic bar-code symbol models and to implications with respect to printing tolerances.
Multiple Description Coding for Closed Loop Systems over Erasure Channels
DEFF Research Database (Denmark)
Østergaard, Jan; Quevedo, Daniel
2013-01-01
) and the decoder (plant). The feedback channel from the decoder to the encoder is assumed noiseless. Since the forward channel is digital, we need to employ quantization.We combine two techniques to enhance the reliability of the system. First, in order to guarantee that the system remains stable during packet......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....... In particular, we transmit M redundant packets, which are constructed such that when receiving any J packets, the current control signal as well as J-1 future control signals can be reliably reconstructed at the decoder. We prove stability subject to quantization constraints, random dropouts, and delays...
International Nuclear Information System (INIS)
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
International Nuclear Information System (INIS)
This report revises the original issuance of NUREG-0017, ''Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE-Code)'' (April 1976), to incorporate more recent operating data now available as well as the results of a number of in-plant measurement programs at operating pressurized water reactors. The PWR-GALE Code is a computerized mathematical model for calculating the releases of radioactive material in gaseous and liquid effluents (i.e., the gaseous and liquid source terms). The US Nuclear Regulatory Commission uses the PWR-GALE Code to determine conformance with the requirements of Appendix I to 10 CFR Part 50
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.
EquiFACS: The Equine Facial Action Coding System.
Wathan, Jen; Burrows, Anne M; Waller, Bridget M; McComb, Karen
2015-01-01
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.
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.
International Nuclear Information System (INIS)
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
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
To calculating on electromagnetic drive for nuclear reactor control systems
International Nuclear Information System (INIS)
Consideration is being given to the results of calculating a magnetic system for a linear polyphase pecking motor (LPPM) with magnetic flux, passing along an armature (longitudinal flux). The considered LPPM are applied in control rod drives at high temperature gas cooled reactors. Calculation algorithm and flowsheet of the PSTAT1 program (FORTRAN, ES computer), realizing this algorithm are described. Applicability of the considered method for calculating unsaturated magnet systems is concluded on the basis of comparison of the obtained data with experimental results
Zelingher, Julian; Ash, Nachman
2013-05-01
The IsraeLi healthcare system has undergone major processes for the adoption of health information technologies (HIT), and enjoys high Levels of utilization in hospital and ambulatory care. Coding is an essential infrastructure component of HIT, and ts purpose is to represent data in a simplified and common format, enhancing its manipulation by digital systems. Proper coding of data enables efficient identification, storage, retrieval and communication of data. UtiLization of uniform coding systems by different organizations enables data interoperability between them, facilitating communication and integrating data elements originating in different information systems from various organizations. Current needs in Israel for heaLth data coding include recording and reporting of diagnoses for hospitalized patients, outpatients and visitors of the Emergency Department, coding of procedures and operations, coding of pathology findings, reporting of discharge diagnoses and causes of death, billing codes, organizational data warehouses and national registries. New national projects for cLinicaL data integration, obligatory reporting of quality indicators and new Ministry of Health (MOH) requirements for HIT necessitate a high Level of interoperability that can be achieved only through the adoption of uniform coding. Additional pressures were introduced by the USA decision to stop the maintenance of the ICD-9-CM codes that are also used by Israeli healthcare, and the adoption of ICD-10-C and ICD-10-PCS as the main coding system for billing purpose. The USA has also mandated utilization of SNOMED-CT as the coding terminology for the ELectronic Health Record problem list, and for reporting quality indicators to the CMS. Hence, the Israeli MOH has recently decided that discharge diagnoses will be reported using ICD-10-CM codes, and SNOMED-CT will be used to code the cLinical information in the EHR. We reviewed the characteristics, strengths and weaknesses of these two coding
Coded aper ture compressive imaging array applied for surveillance systems
Institute of Scientific and Technical Information of China (English)
Jing Chen; Yongtian Wang; Hanxiao Wu
2013-01-01
This paper proposes an application of compressive imaging systems to the problem of wide-area video surveil ance systems. A paral el coded aperture compressive imaging sys-tem and a corresponding motion target detection algorithm in video using compressive image data are developed. Coded masks with random Gaussian, Toeplitz and random binary are utilized to simulate the compressive image respectively. For compres-sive images, a mixture of the Gaussian distribution is applied to the compressed image field to model the background. A simple threshold test in compressive sampling image is used to declare motion objects. Foreground image retrieval from underdetermined measurement using the total variance optimization algorithm is explored. The signal-to-noise ratio (SNR) is employed to evalu-ate the image quality recovered from the compressive sampling signals, and receiver operation characteristic (ROC) curves are used to quantify the performance of the motion detection algo-rithm. Experimental results demonstrate that the low dimensional compressed imaging representation is sufficient to determine spa-tial motion targets. Compared with the random Gaussian and Toeplitz mask, motion detection algorithms using the random bi-nary phase mask can yield better detection results. However using the random Gaussian and Toeplitz phase mask can achieve high resolution reconstructed images.
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.
Technique for Calculating Solution Derivatives With Respect to Geometry Parameters in a CFD Code
Mathur, Sanjay
2011-01-01
A solution has been developed to the challenges of computation of derivatives with respect to geometry, which is not straightforward because these are not typically direct inputs to the computational fluid dynamics (CFD) solver. To overcome these issues, a procedure has been devised that can be used without having access to the mesh generator, while still being applicable to all types of meshes. The basic approach is inspired by the mesh motion algorithms used to deform the interior mesh nodes in a smooth manner when the surface nodes, for example, are in a fluid structure interaction problem. The general idea is to model the mesh edges and nodes as constituting a spring-mass system. Changes to boundary node locations are propagated to interior nodes by allowing them to assume their new equilibrium positions, for instance, one where the forces on each node are in balance. The main advantage of the technique is that it is independent of the volumetric mesh generator, and can be applied to structured, unstructured, single- and multi-block meshes. It essentially reduces the problem down to defining the surface mesh node derivatives with respect to the geometry parameters of interest. For analytical geometries, this is quite straightforward. In the more general case, one would need to be able to interrogate the underlying parametric CAD (computer aided design) model and to evaluate the derivatives either analytically, or by a finite difference technique. Because the technique is based on a partial differential equation (PDE), it is applicable not only to forward mode problems (where derivatives of all the output quantities are computed with respect to a single input), but it could also be extended to the adjoint problem, either by using an analytical adjoint of the PDE or a discrete analog.
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.
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.
Systemization of burnup sensitivity analysis code (2) (Contract research)
International Nuclear Information System (INIS)
Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant economic efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristic is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons: the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion
International Nuclear Information System (INIS)
The integral pressurized water reactor (PWR) concept, which incorporates the nuclear steam supply systems within the reactor vessel, is one of the innovative reactor types with high potential for near term deployment. An International Collaborative Standard Problem (ICSP) on Integral PWR Design, Natural Circulation Flow Stability and Thermohydraulic Coupling of Primary System and Containment during Accidents was established in 2010. Oregon State University, which made available the use of its experimental facility built to demonstrate the feasibility of the Multi-application Small Light Water Reactor (MASLWR) design, and sixteen institutes from seven Member States participated in this ICSP. The objective of the ICSP is to assess computer codes for reactor system design and safety analysis. This objective is achieved through the production of experimental data and computer code simulation of experiments. A loss of feedwater transient with subsequent automatic depressurization system blowdown and long term cooling was selected as the reference event since many different modes of natural circulation phenomena, including the coupling of primary system, high pressure containment and cooling pool are expected to occur during this transient. The power maneuvering transient is also tested to examine the stability of natural circulation during the single and two phase conditions. The ICSP was conducted in three phases: pre-test (with designed initial and boundary conditions established before the experiment was conducted), blind (with real initial and boundary conditions after the experiment was conducted) and open simulation (after the observation of real experimental data). Most advanced thermohydraulic system analysis codes such as TRACE, RELAPS and MARS have been assessed against experiments conducted at the MASLWR test facility. The ICSP has provided all participants with the opportunity to evaluate the strengths and weaknesses of their system codes in the transient
International Nuclear Information System (INIS)
A modular code system RADHEAT-V4 has been developed for performing precisely neutron and photon transport analyses, and shielding safety evaluations. The system consists of the functional modules for producing coupled multi-group neutron and photon cross section sets, for analyzing the neutron and photon transport, and for calculating the atom displacement and the energy deposition due to radiations in nuclear reactor or shielding material. A precise method named Direct Angular Representation (DAR) has been developed for eliminating an error associated with the method of the finite Legendre expansion in evaluating angular distributions of cross sections and radiation fluxes. The DAR method implemented in the code system has been described in detail. To evaluate the accuracy and applicability of the code system, some test calculations on strong anisotropy problems have been performed. From the results, it has been concluded that RADHEAT-V4 is successfully applicable to evaluating shielding problems accurately for fission and fusion reactors and radiation sources. The method employed in the code system is very effective in eliminating negative values and oscillations of angular fluxes in a medium having an anisotropic source or strong streaming. Definitions of the input data required in various options of the code system and the sample problems are also presented. (author)
Development of a tritium transport analysis code for the LMFBR system
Energy Technology Data Exchange (ETDEWEB)
Iizawa, Katsuyuki; Torii, Tatsuo [Japan Nuclear Cycle Development Inst., Tsuruga Head Office, Tsuruga, Fukui (Japan)
2001-03-01
A tritium transport analysis code for the LMFBR system, TTT code, has been developed and validated using data from a power rising test conducted at Monju in 1995. The behavior of tritium during future long-term full power operation of Monju has been estimated. The TTT code was created from the tritium and hydrogen transport model devised by R. Kumar and ANL. Actual data from some plants has been used to improve the code. In this study, we used data from Monju to increase the accuracy of the calculated to measured ratio, the C/E ratio. As a result of the study, we were able to: 1. show that the calculated tritium concentration distribution and the change in the primary and secondary sodium, steam and water correlated sufficiently closely with the measured, C/E ratio of 1.1; 2. propose a transport model between sodium and the cover gas system taking into account the mechanisms affecting the partial pressure difference and the isotopic exchange of H and H3; 3. examine the considerable effect of the hydrogen source within the sodium cooling system of Monju on tritium behavior and clarify the characteristics at the initial stage of plant; 4. estimate the tritium transport and distribution for the long-term full power operation of Monju. The tritium release from the core will be 7,400 TBq during 30 years of operation. The primary and secondary cold trap will capture 99% of this and 1% or less will be released to the environment as gaseous radioactive waste from stack and its drainage water from SG; and 5. compare the best fitted tritium source rates from cores in Phenix and Monju and estimate the major release from Monju's helium bond closed type control rods. (author)