FLASH: A finite element computer code for variably saturated flow
International Nuclear Information System (INIS)
Baca, R.G.; Magnuson, S.O.
1992-05-01
A numerical model was developed for use in performance assessment studies at the INEL. The numerical model, referred to as the FLASH computer code, is designed to simulate two-dimensional fluid flow in fractured-porous media. The code is specifically designed to model variably saturated flow in an arid site vadose zone and saturated flow in an unconfined aquifer. In addition, the code also has the capability to simulate heat conduction in the vadose zone. This report presents the following: description of the conceptual frame-work and mathematical theory; derivations of the finite element techniques and algorithms; computational examples that illustrate the capability of the code; and input instructions for the general use of the code. The FLASH computer code is aimed at providing environmental scientists at the INEL with a predictive tool for the subsurface water pathway. This numerical model is expected to be widely used in performance assessments for: (1) the Remedial Investigation/Feasibility Study process and (2) compliance studies required by the US Department of Energy Order 5820.2A
FLAME: A finite element computer code for contaminant transport n variably-saturated media
International Nuclear Information System (INIS)
Baca, R.G.; Magnuson, S.O.
1992-06-01
A numerical model was developed for use in performance assessment studies at the INEL. The numerical model referred to as the FLAME computer code, is designed to simulate subsurface contaminant transport in a variably-saturated media. The code can be applied to model two-dimensional contaminant transport in an and site vadose zone or in an unconfined aquifer. In addition, the code has the capability to describe transport processes in a porous media with discrete fractures. This report presents the following: description of the conceptual framework and mathematical theory, derivations of the finite element techniques and algorithms, computational examples that illustrate the capability of the code, and input instructions for the general use of the code. The development of the FLAME computer code is aimed at providing environmental scientists at the INEL with a predictive tool for the subsurface water pathway. This numerical model is expected to be widely used in performance assessments for: (1) the Remedial Investigation/Feasibility Study process and (2) compliance studies required by the US Department of energy Order 5820.2A
FLAME: A finite element computer code for contaminant transport n variably-saturated media
Energy Technology Data Exchange (ETDEWEB)
Baca, R.G.; Magnuson, S.O.
1992-06-01
A numerical model was developed for use in performance assessment studies at the INEL. The numerical model referred to as the FLAME computer code, is designed to simulate subsurface contaminant transport in a variably-saturated media. The code can be applied to model two-dimensional contaminant transport in an and site vadose zone or in an unconfined aquifer. In addition, the code has the capability to describe transport processes in a porous media with discrete fractures. This report presents the following: description of the conceptual framework and mathematical theory, derivations of the finite element techniques and algorithms, computational examples that illustrate the capability of the code, and input instructions for the general use of the code. The development of the FLAME computer code is aimed at providing environmental scientists at the INEL with a predictive tool for the subsurface water pathway. This numerical model is expected to be widely used in performance assessments for: (1) the Remedial Investigation/Feasibility Study process and (2) compliance studies required by the US Department of energy Order 5820.2A.
Computing element evolution towards Exascale and its impact on legacy simulation codes
International Nuclear Information System (INIS)
Colin de Verdiere, Guillaume J.L.
2015-01-01
In the light of the current race towards the Exascale, this article highlights the main features of the forthcoming computing elements that will be at the core of next generations of supercomputers. The market analysis, underlying this work, shows that computers are facing a major evolution in terms of architecture. As a consequence, it is important to understand the impacts of those evolutions on legacy codes or programming methods. The problems of dissipated power and memory access are discussed and will lead to a vision of what should be an exascale system. To survive, programming languages had to respond to the hardware evolutions either by evolving or with the creation of new ones. From the previous elements, we elaborate why vectorization, multithreading, data locality awareness and hybrid programming will be the key to reach the exascale, implying that it is time to start rewriting codes. (orig.)
Directory of Open Access Journals (Sweden)
Seyed Abolfazl Hosseini
2016-02-01
Full Text Available In the present paper, development of the three-dimensional (3D computational code based on Galerkin finite element method (GFEM for solving the multigroup forward/adjoint diffusion equation in both rectangular and hexagonal geometries is reported. Linear approximation of shape functions in the GFEM with unstructured tetrahedron elements is used in the calculation. Both criticality and fixed source calculations may be performed using the developed GFEM-3D computational code. An acceptable level of accuracy at a low computational cost is the main advantage of applying the unstructured tetrahedron elements. The unstructured tetrahedron elements generated with Gambit software are used in the GFEM-3D computational code through a developed interface. The forward/adjoint multiplication factor, forward/adjoint flux distribution, and power distribution in the reactor core are calculated using the power iteration method. Criticality calculations are benchmarked against the valid solution of the neutron diffusion equation for International Atomic Energy Agency (IAEA-3D and Water-Water Energetic Reactor (VVER-1000 reactor cores. In addition, validation of the calculations against the P1 approximation of the transport theory is investigated in relation to the liquid metal fast breeder reactor benchmark problem. The neutron fixed source calculations are benchmarked through a comparison with the results obtained from similar computational codes. Finally, an analysis of the sensitivity of calculations to the number of elements is performed.
SAFE: A computer code for the steady-state and transient thermal analysis of LMR fuel elements
International Nuclear Information System (INIS)
Hayes, S.L.
1993-12-01
SAFE is a computer code developed for both the steady-state and transient thermal analysis of single LMR fuel elements. The code employs a two-dimensional control-volume based finite difference methodology with fully implicit time marching to calculate the temperatures throughout a fuel element and its associated coolant channel for both the steady-state and transient events. The code makes no structural calculations or predictions whatsoever. It does, however, accept as input structural parameters within the fuel such as the distributions of porosity and fuel composition, as well as heat generation, to allow a thermal analysis to be performed on a user-specified fuel structure. The code was developed with ease of use in mind. An interactive input file generator and material property correlations internal to the code are available to expedite analyses using SAFE. This report serves as a complete design description of the code as well as a user's manual. A sample calculation made with SAFE is included to highlight some of the code's features. Complete input and output files for the sample problem are provided
Delta: An object-oriented finite element code architecture for massively parallel computers
Energy Technology Data Exchange (ETDEWEB)
Weatherby, J.R.; Schutt, J.A.; Peery, J.S.; Hogan, R.E.
1996-02-01
Delta is an object-oriented code architecture based on the finite element method which enables simulation of a wide range of engineering mechanics problems in a parallel processing environment. Written in C{sup ++}, Delta is a natural framework for algorithm development and for research involving coupling of mechanics from different Engineering Science disciplines. To enhance flexibility and encourage code reuse, the architecture provides a clean separation of the major aspects of finite element programming. Spatial discretization, temporal discretization, and the solution of linear and nonlinear systems of equations are each implemented separately, independent from the governing field equations. Other attractive features of the Delta architecture include support for constitutive models with internal variables, reusable ``matrix-free`` equation solvers, and support for region-to-region variations in the governing equations and the active degrees of freedom. A demonstration code built from the Delta architecture has been used in two-dimensional and three-dimensional simulations involving dynamic and quasi-static solid mechanics, transient and steady heat transport, and flow in porous media.
International Nuclear Information System (INIS)
Valach, M.; Zymak, J.; Svoboda, R.
1997-01-01
This paper presents the development status of the computer codes for the WWER fuel elements thermomechanical behavior modelling under high burnup conditions at the Nuclear Research Institute Rez. The accent is given on the analysis of the results from the parametric calculations, performed by the programmes PIN-W and RODQ2D, rather than on their detailed theoretical description. Several new optional correlations for the UO2 thermal conductivity with degradation effect caused by burnup were implemented into the both codes. Examples of performed calculations document differences between previous and new versions of both programmes. Some recommendations for further development of the codes are given in conclusion. (author). 6 refs, 9 figs
Wakefield Computations for the CLIC PETS using the Parallel Finite Element Time-Domain Code T3P
Energy Technology Data Exchange (ETDEWEB)
Candel, A; Kabel, A.; Lee, L.; Li, Z.; Ng, C.; Schussman, G.; Ko, K.; /SLAC; Syratchev, I.; /CERN
2009-06-19
In recent years, SLAC's Advanced Computations Department (ACD) has developed the high-performance parallel 3D electromagnetic time-domain code, T3P, for simulations of wakefields and transients in complex accelerator structures. T3P is based on advanced higher-order Finite Element methods on unstructured grids with quadratic surface approximation. Optimized for large-scale parallel processing on leadership supercomputing facilities, T3P allows simulations of realistic 3D structures with unprecedented accuracy, aiding the design of the next generation of accelerator facilities. Applications to the Compact Linear Collider (CLIC) Power Extraction and Transfer Structure (PETS) are presented.
Geochemical computer codes. A review
International Nuclear Information System (INIS)
Andersson, K.
1987-01-01
In this report a review of available codes is performed and some code intercomparisons are also discussed. The number of codes treating natural waters (groundwater, lake water, sea water) is large. Most geochemical computer codes treat equilibrium conditions, although some codes with kinetic capability are available. A geochemical equilibrium model consists of a computer code, solving a set of equations by some numerical method and a data base, consisting of thermodynamic data required for the calculations. There are some codes which treat coupled geochemical and transport modeling. Some of these codes solve the equilibrium and transport equations simultaneously while other solve the equations separately from each other. The coupled codes require a large computer capacity and have thus as yet limited use. Three code intercomparisons have been found in literature. It may be concluded that there are many codes available for geochemical calculations but most of them require a user that us quite familiar with the code. The user also has to know the geochemical system in order to judge the reliability of the results. A high quality data base is necessary to obtain a reliable result. The best results may be expected for the major species of natural waters. For more complicated problems, including trace elements, precipitation/dissolution, adsorption, etc., the results seem to be less reliable. (With 44 refs.) (author)
Energy Technology Data Exchange (ETDEWEB)
Saurwein, J.J.
1977-08-01
A system of computer codes has been developed to statistically reduce Peach Bottom fuel test element metrology data and to compare the material strains and fuel rod-fuel hole gaps computed from these data with HTGR design code predictions. The codes included in this system are STAT, STRAIN, GAPS, and DRWDIM. STAT statistically evaluates test element metrology data yielding fuel rod, fuel body, and sleeve irradiation-induced strains; fuel rod anisotropy; and additional data characterizing each analyzed fuel element. STRAIN compares test element fuel rod and fuel body irradiation-induced strains computed from metrology data with the corresponding design code predictions. GAPS compares test element fuel rod, fuel hole heat transfer gaps computed from metrology data with the corresponding design code predictions. DRWDIM plots the measured and predicted gaps and strains. Although specifically developed to expedite the analysis of Peach Bottom fuel test elements, this system can be applied, without extensive modification, to the analysis of Fort St. Vrain or other HTGR-type fuel test elements.
International Nuclear Information System (INIS)
Popescu, C.; Biro, L.; Iftode, I.; Turcu, I.
1975-10-01
The RAP-3A computer code is designed for calculating the main steady state thermo-hydraulic parameters of multirod fuel clusters with liquid metal cooling. The programme provides a double accuracy computation of temperatures and axial enthalpy distributions of pressure losses and axial heat flux distributions in fuel clusters before boiling conditions occur. Physical and mathematical models as well as a sample problem are presented. The code is written in FORTRAN-4 language and is running on a IBM-370/135 computer
Elements of algebraic coding systems
Cardoso da Rocha, Jr, Valdemar
2014-01-01
Elements of Algebraic Coding Systems is an introductory text to algebraic coding theory. In the first chapter, you'll gain inside knowledge of coding fundamentals, which is essential for a deeper understanding of state-of-the-art coding systems. This book is a quick reference for those who are unfamiliar with this topic, as well as for use with specific applications such as cryptography and communication. Linear error-correcting block codes through elementary principles span eleven chapters of the text. Cyclic codes, some finite field algebra, Goppa codes, algebraic decoding algorithms, and applications in public-key cryptography and secret-key cryptography are discussed, including problems and solutions at the end of each chapter. Three appendices cover the Gilbert bound and some related derivations, a derivation of the Mac- Williams' identities based on the probability of undetected error, and two important tools for algebraic decoding-namely, the finite field Fourier transform and the Euclidean algorithm f...
International Nuclear Information System (INIS)
Strickert, R.; Friedman, A.M.; Fried, S.
1979-04-01
A computer program (ARDISC, the Argonne Dispersion Code) is described which simulates the migration of nuclides in porous media and includes first order kinetic effects on the retention constants. The code allows for different absorption and desorption rates and solves the coupled migration equations by arithmetic reiterations. Input data needed are the absorption and desorption rates, equilibrium surface absorption coefficients, flow rates and volumes, and media porosities
International Nuclear Information System (INIS)
Delene, J.
1984-01-01
CONCEPT is a computer code that will provide conceptual capital investment cost estimates for nuclear and coal-fired power plants. The code can develop an estimate for construction at any point in time. Any unit size within the range of about 400 to 1300 MW electric may be selected. Any of 23 reference site locations across the United States and Canada may be selected. PWR, BWR, and coal-fired plants burning high-sulfur and low-sulfur coal can be estimated. Multiple-unit plants can be estimated. Costs due to escalation/inflation and interest during construction are calculated
International Nuclear Information System (INIS)
Rohmann, D.; Koehler, T.
1987-02-01
This is a description of the computer code FIT, written in FORTRAN-77 for a PDP 11/34. FIT is an interactive program to decude position, width and intensity of lines of X-ray spectra (max. length of 4K channels). The lines (max. 30 lines per fit) may have Gauss- or Voigt-profile, as well as exponential tails. Spectrum and fit can be displayed on a Tektronix terminal. (orig.) [de
Energy Technology Data Exchange (ETDEWEB)
Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.; Stuart, D.S.; Thompson, S.L. [Sandia National Labs., Albuquerque, NM (United States); Hodge, S.A.; Hyman, C.R.; Sanders, R.L. [Oak Ridge National Lab., TN (United States)
1995-03-01
MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.
International Nuclear Information System (INIS)
Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.; Stuart, D.S.; Thompson, S.L.; Hodge, S.A.; Hyman, C.R.; Sanders, R.L.
1995-03-01
MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR's phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package
Energy Technology Data Exchange (ETDEWEB)
Anderson, D.V.; Breazeal, J.; Finan, C.H.; Johnston, B.M.
1976-09-14
ABCXYZ is a computer code for obtaining the Cartesian components of the vector potential and the magnetic field on an observed grid from an arrangement of current-carrying wires. Arbitrary combinations of straight line segments, arcs, and loops are allowed in the specification of the currents. Arbitrary positions and orientations of the current-carrying elements are also allowed. Specification of the wire diameter permits the computation of well-defined fields, even in the interiors of the conductors. An optical feature generates magnetic field lines. Extensive graphical and printed output is available to the user including contour, grid-line, and field-line plots. 12 figures, 1 table.
Neural Elements for Predictive Coding
Directory of Open Access Journals (Sweden)
Stewart SHIPP
2016-11-01
Full Text Available Predictive coding theories of sensory brain function interpret the hierarchical construction of the cerebral cortex as a Bayesian, generative model capable of predicting the sensory data consistent with any given percept. Predictions are fed backwards in the hierarchy and reciprocated by prediction error in the forward direction, acting to modify the representation of the outside world at increasing levels of abstraction, and so to optimize the nature of perception over a series of iterations. This accounts for many ‘illusory’ instances of perception where what is seen (heard, etc is unduly influenced by what is expected, based on past experience. This simple conception, the hierarchical exchange of prediction and prediction error, confronts a rich cortical microcircuitry that is yet to be fully documented. This article presents the view that, in the current state of theory and practice, it is profitable to begin a two-way exchange: that predictive coding theory can support an understanding of cortical microcircuit function, and prompt particular aspects of future investigation, whilst existing knowledge of microcircuitry can, in return, influence theoretical development. As an example, a neural inference arising from the earliest formulations of predictive coding is that the source populations of forwards and backwards pathways should be completely separate, given their functional distinction; this aspect of circuitry – that neurons with extrinsically bifurcating axons do not project in both directions – has only recently been confirmed. Here, the computational architecture prescribed by a generalized (free-energy formulation of predictive coding is combined with the classic ‘canonical microcircuit’ and the laminar architecture of hierarchical extrinsic connectivity to produce a template schematic, that is further examined in the light of (a updates in the microcircuitry of primate visual cortex, and (b rapid technical advances made
Neural Elements for Predictive Coding.
Shipp, Stewart
2016-01-01
Predictive coding theories of sensory brain function interpret the hierarchical construction of the cerebral cortex as a Bayesian, generative model capable of predicting the sensory data consistent with any given percept. Predictions are fed backward in the hierarchy and reciprocated by prediction error in the forward direction, acting to modify the representation of the outside world at increasing levels of abstraction, and so to optimize the nature of perception over a series of iterations. This accounts for many 'illusory' instances of perception where what is seen (heard, etc.) is unduly influenced by what is expected, based on past experience. This simple conception, the hierarchical exchange of prediction and prediction error, confronts a rich cortical microcircuitry that is yet to be fully documented. This article presents the view that, in the current state of theory and practice, it is profitable to begin a two-way exchange: that predictive coding theory can support an understanding of cortical microcircuit function, and prompt particular aspects of future investigation, whilst existing knowledge of microcircuitry can, in return, influence theoretical development. As an example, a neural inference arising from the earliest formulations of predictive coding is that the source populations of forward and backward pathways should be completely separate, given their functional distinction; this aspect of circuitry - that neurons with extrinsically bifurcating axons do not project in both directions - has only recently been confirmed. Here, the computational architecture prescribed by a generalized (free-energy) formulation of predictive coding is combined with the classic 'canonical microcircuit' and the laminar architecture of hierarchical extrinsic connectivity to produce a template schematic, that is further examined in the light of (a) updates in the microcircuitry of primate visual cortex, and (b) rapid technical advances made possible by transgenic neural
Cupillard, Paul; Delavaud, Elise; Burgos, Gaël.; Festa, Geatano; Vilotte, Jean-Pierre; Capdeville, Yann; Montagner, Jean-Paul
2012-03-01
The spectral element method, which provides an accurate solution of the elastodynamic problem in heterogeneous media, is implemented in a code, called RegSEM, to compute seismic wave propagation at the regional scale. By regional scale we here mean distances ranging from about 1 km (local scale) to 90° (continental scale). The advantage of RegSEM resides in its ability to accurately take into account 3-D discontinuities such as the sediment-rock interface and the Moho. For this purpose, one version of the code handles local unstructured meshes and another version manages continental structured meshes. The wave equation can be solved in any velocity model, including anisotropy and intrinsic attenuation in the continental version. To validate the code, results from RegSEM are compared to analytical and semi-analytical solutions available in simple cases (e.g. explosion in PREM, plane wave in a hemispherical basin). In addition, realistic simulations of an earthquake in different tomographic models of Europe are performed. All these simulations show the great flexibility of the code and point out the large influence of the shallow layers on the propagation of seismic waves at the regional scale. RegSEM is written in Fortran 90 but it also contains a couple of C routines. It is an open-source software which runs on distributed memory architectures. It can give rise to interesting applications, such as testing regional tomographic models, developing tomography using either passive (i.e. noise correlations) or active (i.e. earthquakes) data, or improving our knowledge on effects linked with sedimentary basins.
Computer code abstract: NESTLE
International Nuclear Information System (INIS)
Turinsky, P.J.; Al-Chalabi, R.M.K.; Engrand, P.; Sarsour, H.N.; Faure, F.X.; Guo, W.
1995-01-01
NESTLE is a few-group neutron diffusion equation solver utilizing the nodal expansion method (NEM) for eigenvalue, adjoint, and fixed-source steady-state and transient problems. The NESTLE code solve the eigenvalue (criticality), eigenvalue adjoint, external fixed-source steady-state, and external fixed-source or eigenvalue initiated transient problems. The eigenvalue problem allows criticality searches to be completed, and the external fixed-source steady-state problem can search to achieve a specified power level. Transient problems model delayed neutrons via precursor groups. Several core properties can be input as time dependent. Two- or four-energy groups can be utilized, with all energy groups being thermal groups (i.e., upscatter exits) is desired. Core geometries modeled include Cartesian and hexagonal. Three-, two-, and one-dimensional models can be utilized with various symmetries. The thermal conditions predicted by the thermal-hydraulic model of the core are used to correct cross sections for temperature and density effects. Cross sections for temperature and density effects. Cross sections are parameterized by color, control rod state (i.e., in or out), and burnup, allowing fuel depletion to be modeled. Either a macroscopic or microscopic model may be employed
Collins, Earl R., Jr.
1990-01-01
Authorized users respond to changing challenges with changing passwords. Scheme for controlling access to computers defeats eavesdroppers and "hackers". Based on password system of challenge and password or sign, challenge, and countersign correlated with random alphanumeric codes in matrices of two or more dimensions. Codes stored on floppy disk or plug-in card and changed frequently. For even higher security, matrices of four or more dimensions used, just as cubes compounded into hypercubes in concurrent processing.
Translation of ARAC computer codes
International Nuclear Information System (INIS)
Takahashi, Kunio; Chino, Masamichi; Honma, Toshimitsu; Ishikawa, Hirohiko; Kai, Michiaki; Imai, Kazuhiko; Asai, Kiyoshi
1982-05-01
In 1981 we have translated the famous MATHEW, ADPIC and their auxiliary computer codes for CDC 7600 computer version to FACOM M-200's. The codes consist of a part of the Atmospheric Release Advisory Capability (ARAC) system of Lawrence Livermore National Laboratory (LLNL). The MATHEW is a code for three-dimensional wind field analysis. Using observed data, it calculates the mass-consistent wind field of grid cells by a variational method. The ADPIC is a code for three-dimensional concentration prediction of gases and particulates released to the atmosphere. It calculates concentrations in grid cells by the particle-in-cell method. They are written in LLLTRAN, i.e., LLNL Fortran language and are implemented on the CDC 7600 computers of LLNL. In this report, i) the computational methods of the MATHEW/ADPIC and their auxiliary codes, ii) comparisons of the calculated results with our JAERI particle-in-cell, and gaussian plume models, iii) translation procedures from the CDC version to FACOM M-200's, are described. Under the permission of LLNL G-Division, this report is published to keep the track of the translation procedures and to serve our JAERI researchers for comparisons and references of their works. (author)
Energy Technology Data Exchange (ETDEWEB)
Eyler, L.L.; Budden, M.J.
1985-03-01
The objective of this work is to assess prediction capabilities and features of the MAGNUM-2D computer code in relation to its intended use in the Basalt Waste Isolation Project (BWIP). This objective is accomplished through a code verification and benchmarking task. Results are documented which support correctness of prediction capabilities in areas of intended model application. 10 references, 43 figures, 11 tables.
Periodic Boundary Conditions in the ALEGRA Finite Element Code
International Nuclear Information System (INIS)
Aidun, John B.; Robinson, Allen C.; Weatherby, Joe R.
1999-01-01
This document describes the implementation of periodic boundary conditions in the ALEGRA finite element code. ALEGRA is an arbitrary Lagrangian-Eulerian multi-physics code with both explicit and implicit numerical algorithms. The periodic boundary implementation requires a consistent set of boundary input sets which are used to describe virtual periodic regions. The implementation is noninvasive to the majority of the ALEGRA coding and is based on the distributed memory parallel framework in ALEGRA. The technique involves extending the ghost element concept for interprocessor boundary communications in ALEGRA to additionally support on- and off-processor periodic boundary communications. The user interface, algorithmic details and sample computations are given
Computer code conversion using HISTORIAN
International Nuclear Information System (INIS)
Matsumoto, Kiyoshi; Kumakura, Toshimasa.
1990-09-01
When a computer program written for a computer A is converted for a computer B, in general, the A version source program is rewritten for B version. However, in this way of program conversion, the following inconvenient problems arise. 1) The original statements to be rewritten for B version are lost. 2) If the original statements of the A version rewritten for B version would remain as comment lines, the B version source program becomes quite large. 3) When update directives of the program are mailed from the organization which developed the program or when some modifications are needed for the program, it is difficult to point out the part to be updated or modified in the B version source program. To solve these problems, the conversion method using the general-purpose software management aid system, HISTORIAN, has been introduced. This conversion method makes a large computer code a easy-to-use program for use to update, modify or improve after the conversion. This report describes the planning and procedures of the conversion method and the MELPROG-PWR/MOD1 code conversion from the CRAY version to the JAERI FACOM version as an example. This report would provide useful information for those who develop or introduce large programs. (author)
A code for obtaining temperature distribution by finite element method
International Nuclear Information System (INIS)
Bloch, M.
1984-01-01
The ELEFIB Fortran language computer code using finite element method for calculating temperature distribution of linear and two dimensional problems, in permanent region or in the transient phase of heat transfer, is presented. The formulation of equations uses the Galerkin method. Some examples are shown and the results are compared with other papers. The comparative evaluation shows that the elaborated code gives good values. (M.C.K.) [pt
SPINLIE - new computer code for polarization calculation
International Nuclear Information System (INIS)
Eidelman, Yu.; Yakimenko, V.
1993-01-01
The application of the analytical method of spin calculation is described. The calculation of both orbital and spin motion is based on Lie operators technique. The computer code SPINLIE realizing this method is discussed. The input language of SPINLIE is compatible with that of MAD. In SPINLIE elements are described as open-quotes thick lensesclose quotes for spin motion as well as for orbital calculations. The explicit expressions of Lie operators were found for orbital and spin motion for elements of different type (bending magnets, quadrupoles, sextupoles, RF-cavities, solenoids, kickers). The rules of addition of spin transformations were obtained for the beam passing the collider structure. Good agreement was found for SPINLIE results for the linear spin resonances in comparison with the other codes (SITF, SMILE). The first results are presented for the calculation of nonlinear spin resonances
International Nuclear Information System (INIS)
Giuliani, Giovanni; Giuliani, Silvano.
1980-01-01
The FORTRAN IV subroutine SURF has been designed to help visualising the results of Finite Element computations. It drawns the axonometric projection of a surface generated in 3-dimensional space by a scalar function over a discretized plane domain. The most important characteristic of the routine is to remove the hidden lines and in this way it enables a clear vision of the details of the generated surface
Meinders, Vincent T.; van den Boogaard, Antonius H.; Huetink, Han
2003-01-01
To intensify the use of implicit finite element codes for solving large scale problems, the computation time of these codes has to be decreased drastically. A method is developed which decreases the computational time of implicit codes by factors. The method is based on introducing inertia effects
International Nuclear Information System (INIS)
Kedziur, F.
1982-07-01
The computer code CALIPSO was developed for the calculation of a hypothetical accident in an LMFBR (Liquid Metal Fast Breeder Reactor), where the failure of fuel pins is assumed. It calculates two-dimensionally the thermodynamics, fluiddynamics and changes in geometry of a single fuel pin and its coolant channel in a time period between failure of the pin and a state, at which the geometry is nearly destroyed. The determination of temperature profiles in the fuel pin cladding and the channel wall make it possible to take melting and freezing processes into account. Further features of CALIPSO are the variable channel cross section in order to model disturbances of the channel geometry as well as the calculation of two velocity fields including the consideration of virtual mass effects. The documented version of CALIPSO is especially suited for the calculation of the SIMBATH experiments carried out at the Kernforschungszentrum Karlsruhe, which simulate the above-mentioned accident. The report contains the complete documentation of the CALIPSO code: the modeling of the geometry, the equations used, the structure of the code and the solution procedure as well as the instructions for use with an application example. (orig.) [de
Computer codes for ventilation in nuclear facilities
International Nuclear Information System (INIS)
Mulcey, P.
1987-01-01
In this paper the authors present some computer codes, developed in the last years, for ventilation and radioprotection. These codes are used for safety analysis in the conception, exploitation and dismantlement of nuclear facilities. The authors present particularly: DACC1 code used for aerosol deposit in sampling circuit of radiation monitors; PIAF code used for modelization of complex ventilation system; CLIMAT 6 code used for optimization of air conditioning system [fr
The computer code EURDYN-1M (release 2). User's manual
International Nuclear Information System (INIS)
1982-01-01
EURDYN-1M is a finite element computer code developed at J.R.C. Ispra to compute the response of two-dimensional coupled fluid-structure configurations to transient dynamic loading for reactor safety studies. This report gives instructions for preparing input data to EURDYN-1M, release 2, and describes a test problem in order to illustrate both the input and the output of the code
Reactor safety computer code development at INEL
International Nuclear Information System (INIS)
Johnsen, G.W.
1985-01-01
This report provides a brief overview of the computer code development programs being conducted at EG and G Idaho, Inc. on behalf of US Nuclear Regulatory Commission and the Department of Energy, Idaho Operations Office. Included are descriptions of the codes being developed, their development status as of the date of this report, and resident code development expertise
Implatation of MC2 computer code
International Nuclear Information System (INIS)
Seehusen, J.; Nair, R.P.K.; Becceneri, J.C.
1981-01-01
The implantation of MC2 computer code in the CDC system is presented. The MC2 computer code calculates multigroup cross sections for tipical compositions of fast reactors. The multigroup constants are calculated using solutions of PI or BI approximations for determined buckling value as weighting function. (M.C.K.) [pt
Dual-code quantum computation model
Choi, Byung-Soo
2015-08-01
In this work, we propose the dual-code quantum computation model—a fault-tolerant quantum computation scheme which alternates between two different quantum error-correction codes. Since the chosen two codes have different sets of transversal gates, we can implement a universal set of gates transversally, thereby reducing the overall cost. We use code teleportation to convert between quantum states in different codes. The overall cost is decreased if code teleportation requires fewer resources than the fault-tolerant implementation of the non-transversal gate in a specific code. To analyze the cost reduction, we investigate two cases with different base codes, namely the Steane and Bacon-Shor codes. For the Steane code, neither the proposed dual-code model nor another variation of it achieves any cost reduction since the conventional approach is simple. For the Bacon-Shor code, the three proposed variations of the dual-code model reduce the overall cost. However, as the encoding level increases, the cost reduction decreases and becomes negative. Therefore, the proposed dual-code model is advantageous only when the encoding level is low and the cost of the non-transversal gate is relatively high.
Computation of the bounce-average code
International Nuclear Information System (INIS)
Cutler, T.A.; Pearlstein, L.D.; Rensink, M.E.
1977-01-01
The bounce-average computer code simulates the two-dimensional velocity transport of ions in a mirror machine. The code evaluates and bounce-averages the collision operator and sources along the field line. A self-consistent equilibrium magnetic field is also computed using the long-thin approximation. Optionally included are terms that maintain μ, J invariance as the magnetic field changes in time. The assumptions and analysis that form the foundation of the bounce-average code are described. When references can be cited, the required results are merely stated and explained briefly. A listing of the code is appended
A surface code quantum computer in silicon.
Hill, Charles D; Peretz, Eldad; Hile, Samuel J; House, Matthew G; Fuechsle, Martin; Rogge, Sven; Simmons, Michelle Y; Hollenberg, Lloyd C L
2015-10-01
The exceptionally long quantum coherence times of phosphorus donor nuclear spin qubits in silicon, coupled with the proven scalability of silicon-based nano-electronics, make them attractive candidates for large-scale quantum computing. However, the high threshold of topological quantum error correction can only be captured in a two-dimensional array of qubits operating synchronously and in parallel-posing formidable fabrication and control challenges. We present an architecture that addresses these problems through a novel shared-control paradigm that is particularly suited to the natural uniformity of the phosphorus donor nuclear spin qubit states and electronic confinement. The architecture comprises a two-dimensional lattice of donor qubits sandwiched between two vertically separated control layers forming a mutually perpendicular crisscross gate array. Shared-control lines facilitate loading/unloading of single electrons to specific donors, thereby activating multiple qubits in parallel across the array on which the required operations for surface code quantum error correction are carried out by global spin control. The complexities of independent qubit control, wave function engineering, and ad hoc quantum interconnects are explicitly avoided. With many of the basic elements of fabrication and control based on demonstrated techniques and with simulated quantum operation below the surface code error threshold, the architecture represents a new pathway for large-scale quantum information processing in silicon and potentially in other qubit systems where uniformity can be exploited.
Computer Code for Nanostructure Simulation
Filikhin, Igor; Vlahovic, Branislav
2009-01-01
Due to their small size, nanostructures can have stress and thermal gradients that are larger than any macroscopic analogue. These gradients can lead to specific regions that are susceptible to failure via processes such as plastic deformation by dislocation emission, chemical debonding, and interfacial alloying. A program has been developed that rigorously simulates and predicts optoelectronic properties of nanostructures of virtually any geometrical complexity and material composition. It can be used in simulations of energy level structure, wave functions, density of states of spatially configured phonon-coupled electrons, excitons in quantum dots, quantum rings, quantum ring complexes, and more. The code can be used to calculate stress distributions and thermal transport properties for a variety of nanostructures and interfaces, transport and scattering at nanoscale interfaces and surfaces under various stress states, and alloy compositional gradients. The code allows users to perform modeling of charge transport processes through quantum-dot (QD) arrays as functions of inter-dot distance, array order versus disorder, QD orientation, shape, size, and chemical composition for applications in photovoltaics and physical properties of QD-based biochemical sensors. The code can be used to study the hot exciton formation/relation dynamics in arrays of QDs of different shapes and sizes at different temperatures. It also can be used to understand the relation among the deposition parameters and inherent stresses, strain deformation, heat flow, and failure of nanostructures.
International Nuclear Information System (INIS)
Yerkess, A.
1984-01-01
SEURBNUK-2 has been designed to model the hydrodynamic development in time of a hypothetical core disrupture accident in a fast breeder reactor. SEURBNUK-2 is a two-dimensional, axisymmetric, eulerian, finite difference containment code. The numerical procedure adopted in SEURBNUK to solve the hydrodynamic equations is based on the semi-implicit ICE method. SEURBNUK has a full thin shell treatment for tanks of arbitrary shape and includes the effects of the compressibility of the fluid. Fluid flow through porous media and porous structures can also be accommodated. An important feature of SEURBNUK is that the thin shell equations are solved quite separately from those of the fluid, and the time step for the fluid flow calculation can be an integer multiple of that for calculating the shell motion. The interaction of the shell with the fluid is then considered as a modification to the coefficients in the implicit pressure equations, the modifications naturally depending on the behaviour of the thin shell section within the fluid cell. The code is limited to dealing with a single fluid, the coolant, whereas the bubble and the cover gas are treated as cavities of uniform pressure calculated via appropriate pressure-volume-energy relationships. This manual describes the input data specifications needed for the execution of SEURBNUK-2 calculations and nine sample problems of varying degrees of complexity highlight the code capabilities. After explaining the output facilities information is included to aid those unfamiliar with SEURBNUK-2 to avoid the common pit-falls experienced by novices
ANACROM - A computer code for chromatogram analysis
International Nuclear Information System (INIS)
Gouveia, A.S. de; Mesquita, C.H. de.
1981-01-01
The computer code was developed for automatic research of peaks and evaluation of chromatogram parameters as : center, height, area, medium - height width (FWHM) and the rate FWHM/center of each peak. (Author) [pt
Quantum computation with Turaev-Viro codes
International Nuclear Information System (INIS)
Koenig, Robert; Kuperberg, Greg; Reichardt, Ben W.
2010-01-01
For a 3-manifold with triangulated boundary, the Turaev-Viro topological invariant can be interpreted as a quantum error-correcting code. The code has local stabilizers, identified by Levin and Wen, on a qudit lattice. Kitaev's toric code arises as a special case. The toric code corresponds to an abelian anyon model, and therefore requires out-of-code operations to obtain universal quantum computation. In contrast, for many categories, such as the Fibonacci category, the Turaev-Viro code realizes a non-abelian anyon model. A universal set of fault-tolerant operations can be implemented by deforming the code with local gates, in order to implement anyon braiding. We identify the anyons in the code space, and present schemes for initialization, computation and measurement. This provides a family of constructions for fault-tolerant quantum computation that are closely related to topological quantum computation, but for which the fault tolerance is implemented in software rather than coming from a physical medium.
Cloud Computing for Complex Performance Codes.
Energy Technology Data Exchange (ETDEWEB)
Appel, Gordon John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Klein, Brandon Thorin [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Miner, John Gifford [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
2017-02-01
This report describes the use of cloud computing services for running complex public domain performance assessment problems. The work consisted of two phases: Phase 1 was to demonstrate complex codes, on several differently configured servers, could run and compute trivial small scale problems in a commercial cloud infrastructure. Phase 2 focused on proving non-trivial large scale problems could be computed in the commercial cloud environment. The cloud computing effort was successfully applied using codes of interest to the geohydrology and nuclear waste disposal modeling community.
Computer codes in particle transport physics
International Nuclear Information System (INIS)
Pesic, M.
2004-01-01
Simulation of transport and interaction of various particles in complex media and wide energy range (from 1 MeV up to 1 TeV) is very complicated problem that requires valid model of a real process in nature and appropriate solving tool - computer code and data library. A brief overview of computer codes based on Monte Carlo techniques for simulation of transport and interaction of hadrons and ions in wide energy range in three dimensional (3D) geometry is shown. Firstly, a short attention is paid to underline the approach to the solution of the problem - process in nature - by selection of the appropriate 3D model and corresponding tools - computer codes and cross sections data libraries. Process of data collection and evaluation from experimental measurements and theoretical approach to establishing reliable libraries of evaluated cross sections data is Ion g, difficult and not straightforward activity. For this reason, world reference data centers and specialized ones are acknowledged, together with the currently available, state of art evaluated nuclear data libraries, as the ENDF/B-VI, JEF, JENDL, CENDL, BROND, etc. Codes for experimental and theoretical data evaluations (e.g., SAMMY and GNASH) together with the codes for data processing (e.g., NJOY, PREPRO and GRUCON) are briefly described. Examples of data evaluation and data processing to generate computer usable data libraries are shown. Among numerous and various computer codes developed in transport physics of particles, the most general ones are described only: MCNPX, FLUKA and SHIELD. A short overview of basic application of these codes, physical models implemented with their limitations, energy ranges of particles and types of interactions, is given. General information about the codes covers also programming language, operation system, calculation speed and the code availability. An example of increasing computation speed of running MCNPX code using a MPI cluster compared to the code sequential option
Computer code development plant for SMART design
International Nuclear Information System (INIS)
Bae, Kyoo Hwan; Choi, S.; Cho, B.H.; Kim, K.K.; Lee, J.C.; Kim, J.P.; Kim, J.H.; Chung, M.; Kang, D.J.; Chang, M.H.
1999-03-01
In accordance with the localization plan for the nuclear reactor design driven since the middle of 1980s, various computer codes have been transferred into the korea nuclear industry through the technical transfer program from the worldwide major pressurized water reactor supplier or through the international code development program. These computer codes have been successfully utilized in reactor and reload core design works. As the results, design- related technologies have been satisfactorily accumulated. However, the activities for the native code development activities to substitute the some important computer codes of which usages are limited by the original technique owners have been carried out rather poorly. Thus, it is most preferentially required to secure the native techniques on the computer code package and analysis methodology in order to establish the capability required for the independent design of our own model of reactor. Moreover, differently from the large capacity loop-type commercial reactors, SMART (SYSTEM-integrated Modular Advanced ReacTor) design adopts a single reactor pressure vessel containing the major primary components and has peculiar design characteristics such as self-controlled gas pressurizer, helical steam generator, passive residual heat removal system, etc. Considering those peculiar design characteristics for SMART, part of design can be performed with the computer codes used for the loop-type commercial reactor design. However, most of those computer codes are not directly applicable to the design of an integral reactor such as SMART. Thus, they should be modified to deal with the peculiar design characteristics of SMART. In addition to the modification efforts, various codes should be developed in several design area. Furthermore, modified or newly developed codes should be verified their reliability through the benchmarking or the test for the object design. Thus, it is necessary to proceed the design according to the
Gender codes why women are leaving computing
Misa, Thomas J
2010-01-01
The computing profession is facing a serious gender crisis. Women are abandoning the computing field at an alarming rate. Fewer are entering the profession than anytime in the past twenty-five years, while too many are leaving the field in mid-career. With a maximum of insight and a minimum of jargon, Gender Codes explains the complex social and cultural processes at work in gender and computing today. Edited by Thomas Misa and featuring a Foreword by Linda Shafer, Chair of the IEEE Computer Society Press, this insightful collection of essays explores the persisting gender imbalance in computing and presents a clear course of action for turning things around.
International Nuclear Information System (INIS)
Merle-Szeremeta, A.; Thomassin, A.
1999-01-01
The Institute of Protection and Nuclear Safety (I.P.S.N.) has developed a computer code, FOCON96 1.0 to calculate the dosimetric consequences of atmospheric radioactive releases from nuclear installations after several years of usual operation. This communication describes the principal characteristics of FOCON96 1.0 and its functionalities. The principal elements of a comparison between FOCON96 1.0 and PC-CREAM ( European computer code developed by the N.R.P.B. and answering the same criteria) are given here. (N.C.)
Secular perturbation theory and computation of asteroid proper elements
Milani, Andrea; Knezevic, Zoran
1991-01-01
A new theory for the calculation of proper elements is presented. This theory defines an explicit algorithm applicable to any chosen set of orbits and accounts for the effect of shallow resonances on secular frequencies. The proper elements are computed with an iterative algorithm and the behavior of the iteration can be used to define a quality code.
Verification of structural analysis computer codes in nuclear engineering
International Nuclear Information System (INIS)
Zebeljan, Dj.; Cizelj, L.
1990-01-01
Sources of potential errors, which can take place during use of finite element method based computer programs, are described in the paper. The magnitude of errors was defined as acceptance criteria for those programs. Error sources are described as they are treated by 'National Agency for Finite Element Methods and Standards (NAFEMS)'. Specific verification examples are used from literature of Nuclear Regulatory Commission (NRC). Example of verification is made on PAFEC-FE computer code for seismic response analyses of piping systems by response spectrum method. (author)
The computer code SEURBNUK-2: Recent developments
International Nuclear Information System (INIS)
Staniforth, R.; Yerkees, A.
1979-01-01
The SEURBNUK-2 code is now being developed jointly by AEE Winfrith and JRC Ispra for use in Fast Reactor Containment Studies. To meet the needs of such studies and the needs of the COVA programme, a number of improvements and extensions of the code have been made. A selection of these changes and illustrations of their use are given in this paper. The structural capability of SEURBNUK-2 was originally limited to the treatment of thin shells, and shell junctions. Although this facility proved surprisingly useful, it was realised that a more versatile and powerful means of calculating the deformation of more complicated structural geometries would be required. The finite element code EURDYN which employs convected coordinates was adapted for the purpose, so that axially symmetric elements of the isoparametric, triangular and thin shell families could be used to model various parts of the reactor structure. The method of coupling this finite element code to the fluid motion is described and the use of this new version of the code is illustrated and the results compared with those obtained by the original code and by experiment. A feature of many reactor designs which is being modelled in the later COVA experiments is the perforated plate or porous structure. For fixed perforated plates and porous structures, the additional pressure drop and inertia effects can be included in the momentum equations by addition of suitable terms and the original technique of solution is unaltered. Details of the finite difference equations are given in this paper together with the results of check calculations which were performed to ensure the correct functioning of the code. (orig.)
LATTICE: an interactive lattice computer code
International Nuclear Information System (INIS)
Staples, J.
1976-10-01
LATTICE is a computer code which enables an interactive user to calculate the functions of a synchrotron lattice. This program satisfies the requirements at LBL for a simple interactive lattice program by borrowing ideas from both TRANSPORT and SYNCH. A fitting routine is included
Computer Security: is your code sane?
Stefan Lueders, Computer Security Team
2015-01-01
How many of us write code? Software? Programs? Scripts? How many of us are properly trained in this and how well do we do it? Do we write functional, clean and correct code, without flaws, bugs and vulnerabilities*? In other words: are our codes sane? Figuring out weaknesses is not that easy (see our quiz in an earlier Bulletin article). Therefore, in order to improve the sanity of your code, prevent common pit-falls, and avoid the bugs and vulnerabilities that can crash your code, or – worse – that can be misused and exploited by attackers, the CERN Computer Security team has reviewed its recommendations for checking the security compliance of your code. “Static Code Analysers” are stand-alone programs that can be run on top of your software stack, regardless of whether it uses Java, C/C++, Perl, PHP, Python, etc. These analysers identify weaknesses and inconsistencies including: employing undeclared variables; expressions resu...
MESHJET. A mesh generation package for finite element MHD equilibrium codes at JET
International Nuclear Information System (INIS)
Springmann, E.; Taroni, A.
1984-01-01
MESHJET is a fairly general package and can be used to generate meshes for any finite element code in two space dimensions. These finite element codes are widely used at JET. The first code is for the identification of the plasma boundary and internal flux surfaces from measurements of external fluxes and fields under the assumption that the plasma toroidal density can be represented within a given class of functions. The second code computes plasma equilibrium configurations taking into account a two-dimensional model of the transformer iron core in JET. (author)
Evaluation of the SCANAIR Computer Code
International Nuclear Information System (INIS)
Jernkvist, Lars Olof; Massih, Ali
2001-11-01
The SCANAIR computer code, version 3.2, has been evaluated from the standpoint of its capability to analyze, simulate and predict nuclear fuel behavior during severe power transients. SCANAIR calculates the thermal and mechanical behavior of a pressurized water reactor (PWR) fuel rod during a postulated reactivity initiated accident (RIA), and our evaluation indicates that SCANAIR is a state of the art computational tool for this purpose. Our evaluation starts by reviewing the basic theoretical models in SCANAIR, namely the governing equations for heat transfer, the mechanical response of fuel and clad, and the fission gas release behavior. The numerical methods used to solve the governing equations are briefly reviewed, and the range of applicability of the models and their limitations are discussed and illustrated with examples. Next, the main features of the SCANAIR user interface are delineated. The code requires an extensive amount of input data, in order to define burnup-dependent initial conditions to the simulated RIA. These data must be provided in a special format by a thermal-mechanical fuel rod analysis code. The user also has to supply the transient power history under RIA as input, which requires a code for neutronics calculation. The programming structure and documentation of the code are also addressed in our evaluation. SCANAIR is programmed in Fortran-77, and makes use of several general Fortran-77 libraries for handling input/output, data storage and graphical presentation of computed results. The documentation of SCANAIR and its helping libraries is generally of good quality. A drawback with SCANAIR in its present form, is that the code and its pre- and post-processors are tied to computers running the Unix or Linux operating systems. As part of our evaluation, we have performed a large number of computations with SCANAIR, some of which are documented in this report. The computations presented here include a hypothetical RIA in a high
Computer codes used in particle accelerator design: First edition
International Nuclear Information System (INIS)
1987-01-01
This paper contains a listing of more than 150 programs that have been used in the design and analysis of accelerators. Given on each citation are person to contact, classification of the computer code, publications describing the code, computer and language runned on, and a short description of the code. Codes are indexed by subject, person to contact, and code acronym
(Nearly) portable PIC code for parallel computers
International Nuclear Information System (INIS)
Decyk, V.K.
1993-01-01
As part of the Numerical Tokamak Project, the author has developed a (nearly) portable, one dimensional version of the GCPIC algorithm for particle-in-cell codes on parallel computers. This algorithm uses a spatial domain decomposition for the fields, and passes particles from one domain to another as the particles move spatially. With only minor changes, the code has been run in parallel on the Intel Delta, the Cray C-90, the IBM ES/9000 and a cluster of workstations. After a line by line translation into cmfortran, the code was also run on the CM-200. Impressive speeds have been achieved, both on the Intel Delta and the Cray C-90, around 30 nanoseconds per particle per time step. In addition, the author was able to isolate the data management modules, so that the physics modules were not changed much from their sequential version, and the data management modules can be used as open-quotes black boxes.close quotes
Implementing a modular system of computer codes
International Nuclear Information System (INIS)
Vondy, D.R.; Fowler, T.B.
1983-07-01
A modular computation system has been developed for nuclear reactor core analysis. The codes can be applied repeatedly in blocks without extensive user input data, as needed for reactor history calculations. The primary control options over the calculational paths and task assignments within the codes are blocked separately from other instructions, admitting ready access by user input instruction or directions from automated procedures and promoting flexible and diverse applications at minimum application cost. Data interfacing is done under formal specifications with data files manipulated by an informed manager. This report emphasizes the system aspects and the development of useful capability, hopefully informative and useful to anyone developing a modular code system of much sophistication. Overall, this report in a general way summarizes the many factors and difficulties that are faced in making reactor core calculations, based on the experience of the authors. It provides the background on which work on HTGR reactor physics is being carried out
Review and Evaluation of a Turbomachinery Throughflow Finite Element Code
1989-06-01
stare. and ZIP code) 7b Address (city, state, and ZIP code,, Monterev, CA 93943- 5000 Monterey, CA 93943- 50001 Sa Name of Funding Sponsoring...Rotor Tip Section ............................. 52 Figure 27. Computational Mesh for the Blade-to-Blade Solution .............. 53 Figure 28. Iso -pressure...computation, the results are presented here for Case 5 at a flow coefficient of 0.61. The computed iso -pressure lines on the axisvrmrnetric stream surface at
Evaluation of the FRAPCON-3 Computer Code
International Nuclear Information System (INIS)
Jernkvist, Lars Olof; Massih, Ali
2002-03-01
The FRAPCON-3 computer code has been evaluated with respect to its applicability, modeling capability, user friendliness, source code structure and supporting experimental database. The code is intended for thermo-mechanical analyses of light water reactor nuclear fuel rods under steady-state operational conditions and moderate power excursions. It is applicable to both boiling- and pressurized water reactor fuel rods with UO 2 fuel, ranging up to about 65 MWd/kg U in rod average burnup. The models and numerical methods in FRAPCON-3 are relatively simple, which makes the code transparent and also fairly easy to modify and extend for the user. The fundamental equations for heat transfer, structural analysis and fuel fission gas release are solved in one-dimensional (radial) and stationary (time-independent) form, and interaction between axial segments of the rod is confined to calculations of coolant axial flow and rod internal gas pressure. The code is fairly easy to use; fuel rod design data and time histories of fuel rod power and coolant inlet conditions are input via a single text file, and the corresponding calculated variation with time of important fuel rod parameters are printed to a single output file in textual form. The results can also be presented in graphical form through an interface to the general graphics program xmgr. FRAPCON-3 also provides the possibility to export calculated results to the transient fuel rod analysis code FRAPTRAN, where the data can be used as burnup-dependent initial conditions to a postulated transient. Most of the source code to FRAPCON-3 is written in Fortran-IV, which is an archaic, non-standard dialect of the Fortran programming language. Since Fortran-IV is not accepted by all compilers for the latest standard of the language, Fortran-95, there is a risk that the source code must be partly rewritten in the future. Documentation of the code comprises (i) a general code description, which briefly presents models
Evaluation of the FRAPCON-3 Computer Code
Energy Technology Data Exchange (ETDEWEB)
Jernkvist, Lars Olof; Massih, Ali [Quantum Technologies AB, Uppsala (Sweden)
2002-03-01
The FRAPCON-3 computer code has been evaluated with respect to its applicability, modeling capability, user friendliness, source code structure and supporting experimental database. The code is intended for thermo-mechanical analyses of light water reactor nuclear fuel rods under steady-state operational conditions and moderate power excursions. It is applicable to both boiling- and pressurized water reactor fuel rods with UO{sub 2} fuel, ranging up to about 65 MWd/kg U in rod average burnup. The models and numerical methods in FRAPCON-3 are relatively simple, which makes the code transparent and also fairly easy to modify and extend for the user. The fundamental equations for heat transfer, structural analysis and fuel fission gas release are solved in one-dimensional (radial) and stationary (time-independent) form, and interaction between axial segments of the rod is confined to calculations of coolant axial flow and rod internal gas pressure. The code is fairly easy to use; fuel rod design data and time histories of fuel rod power and coolant inlet conditions are input via a single text file, and the corresponding calculated variation with time of important fuel rod parameters are printed to a single output file in textual form. The results can also be presented in graphical form through an interface to the general graphics program xmgr. FRAPCON-3 also provides the possibility to export calculated results to the transient fuel rod analysis code FRAPTRAN, where the data can be used as burnup-dependent initial conditions to a postulated transient. Most of the source code to FRAPCON-3 is written in Fortran-IV, which is an archaic, non-standard dialect of the Fortran programming language. Since Fortran-IV is not accepted by all compilers for the latest standard of the language, Fortran-95, there is a risk that the source code must be partly rewritten in the future. Documentation of the code comprises (i) a general code description, which briefly presents models
An integrated radiation physics computer code system.
Steyn, J. J.; Harris, D. W.
1972-01-01
An integrated computer code system for the semi-automatic and rapid analysis of experimental and analytic problems in gamma photon and fast neutron radiation physics is presented. Such problems as the design of optimum radiation shields and radioisotope power source configurations may be studied. The system codes allow for the unfolding of complex neutron and gamma photon experimental spectra. Monte Carlo and analytic techniques are used for the theoretical prediction of radiation transport. The system includes a multichannel pulse-height analyzer scintillation and semiconductor spectrometer coupled to an on-line digital computer with appropriate peripheral equipment. The system is geometry generalized as well as self-contained with respect to material nuclear cross sections and the determination of the spectrometer response functions. Input data may be either analytic or experimental.
H/sup 0/ precessor computer code
Energy Technology Data Exchange (ETDEWEB)
van Dyck, O.B.; Floyd, R.A.
1981-05-01
A spin precessor using H/sup -/ to H/sup 0/ stripping, followed by small precession magnets, has been developed for the LAMPF 800-MeV polarized H/sup -/ beam. The performance of the system was studied with the computer code documented in this report. The report starts from the fundamental physics of a system of spins with hyperfine coupling in a magnetic field and contains many examples of beam behavior as calculated by the program.
Poisson/Superfish codes for personal computers
International Nuclear Information System (INIS)
Humphries, S.
1992-01-01
The Poisson/Superfish codes calculate static E or B fields in two-dimensions and electromagnetic fields in resonant structures. New versions for 386/486 PCs and Macintosh computers have capabilities that exceed the mainframe versions. Notable improvements are interactive graphical post-processors, improved field calculation routines, and a new program for charged particle orbit tracking. (author). 4 refs., 1 tab., figs
Validation Report for ISAAC Computer Code
International Nuclear Information System (INIS)
Kim, Dong Ha; Song, Yong Mann; Park, Soo Yong; Jin, Young Ho; Kim, See Darl; Kim, Sang Baik
2008-12-01
A fully integrated severe accident code ISAAC was developed to simulate the accident scenarios that could lead to a severe core damage and eventually to the containment failure in CANDU reactors. Three ways of validation were adopted in this report. The first approach is to show the ISAAC results for the typical severe core damage sequences. In general, the ISAAC computer code shows the reasonable results in terms of the thermal hydraulic behavior as well as fission product transport from the PHTS to the containment. As the second step, the ISAAC results are compared against those from CATHENA and MAAP4-CANDU. In spite of the modeling differences, the overall trend is similar to each other. Especially, the major severe accident phenomena and the accident progression are similar to MAAP4-CANDU, though ISAAC predicts the accident progression faster. Finally ISAAC results are compared with the experimental data. The ISAAC models provide a good agreement with the measured data. Still more efforts are needed to validate the code by the code-to-code comparison and the comparison against the experimental data available
Computing Challenges in Coded Mask Imaging
Skinner, Gerald
2009-01-01
This slide presaentation reviews the complications and challenges in developing computer systems for Coded Mask Imaging telescopes. The coded mask technique is used when there is no other way to create the telescope, (i.e., when there are wide fields of view, high energies for focusing or low energies for the Compton/Tracker Techniques and very good angular resolution.) The coded mask telescope is described, and the mask is reviewed. The coded Masks for the INTErnational Gamma-Ray Astrophysics Laboratory (INTEGRAL) instruments are shown, and a chart showing the types of position sensitive detectors used for the coded mask telescopes is also reviewed. Slides describe the mechanism of recovering an image from the masked pattern. The correlation with the mask pattern is described. The Matrix approach is reviewed, and other approaches to image reconstruction are described. Included in the presentation is a review of the Energetic X-ray Imaging Survey Telescope (EXIST) / High Energy Telescope (HET), with information about the mission, the operation of the telescope, comparison of the EXIST/HET with the SWIFT/BAT and details of the design of the EXIST/HET.
Finite element code development for modeling detonation of HMX composites
Duran, Adam V.; Sundararaghavan, Veera
2017-01-01
In this work, we present a hydrodynamics code for modeling shock and detonation waves in HMX. A stable efficient solution strategy based on a Taylor-Galerkin finite element (FE) discretization was developed to solve the reactive Euler equations. In our code, well calibrated equations of state for the solid unreacted material and gaseous reaction products have been implemented, along with a chemical reaction scheme and a mixing rule to define the properties of partially reacted states. A linear Gruneisen equation of state was employed for the unreacted HMX calibrated from experiments. The JWL form was used to model the EOS of gaseous reaction products. It is assumed that the unreacted explosive and reaction products are in both pressure and temperature equilibrium. The overall specific volume and internal energy was computed using the rule of mixtures. Arrhenius kinetics scheme was integrated to model the chemical reactions. A locally controlled dissipation was introduced that induces a non-oscillatory stabilized scheme for the shock front. The FE model was validated using analytical solutions for SOD shock and ZND strong detonation models. Benchmark problems are presented for geometries in which a single HMX crystal is subjected to a shock condition.
ELLIPT2D: A Flexible Finite Element Code Written Python
International Nuclear Information System (INIS)
Pletzer, A.; Mollis, J.C.
2001-01-01
The use of the Python scripting language for scientific applications and in particular to solve partial differential equations is explored. It is shown that Python's rich data structure and object-oriented features can be exploited to write programs that are not only significantly more concise than their counter parts written in Fortran, C or C++, but are also numerically efficient. To illustrate this, a two-dimensional finite element code (ELLIPT2D) has been written. ELLIPT2D provides a flexible and easy-to-use framework for solving a large class of second-order elliptic problems. The program allows for structured or unstructured meshes. All functions defining the elliptic operator are user supplied and so are the boundary conditions, which can be of Dirichlet, Neumann or Robbins type. ELLIPT2D makes extensive use of dictionaries (hash tables) as a way to represent sparse matrices.Other key features of the Python language that have been widely used include: operator over loading, error handling, array slicing, and the Tkinter module for building graphical use interfaces. As an example of the utility of ELLIPT2D, a nonlinear solution of the Grad-Shafranov equation is computed using a Newton iterative scheme. A second application focuses on a solution of the toroidal Laplace equation coupled to a magnetohydrodynamic stability code, a problem arising in the context of magnetic fusion research
Computer codes for the analysis of flask impact problems
International Nuclear Information System (INIS)
Neilson, A.J.
1984-09-01
This review identifies typical features of the design of transportation flasks and considers some of the analytical tools required for the analysis of impact events. Because of the complexity of the physical problem, it is unlikely that a single code will adequately deal with all the aspects of the impact incident. Candidate codes are identified on the basis of current understanding of their strengths and limitations. It is concluded that the HONDO-II, DYNA3D AND ABAQUS codes which ar already mounted on UKAEA computers will be suitable tools for use in the analysis of experiments conducted in the proposed AEEW programme and of general flask impact problems. Initial attention should be directed at the DYNA3D and ABAQUS codes with HONDO-II being reserved for situations where the three-dimensional elements of DYNA3D may provide uneconomic simulations in planar or axisymmetric geometries. Attention is drawn to the importance of access to suitable mesh generators to create the nodal coordinate and element topology data required by these structural analysis codes. (author)
SALE: Safeguards Analytical Laboratory Evaluation computer code
Energy Technology Data Exchange (ETDEWEB)
Carroll, D.J.; Bush, W.J.; Dolan, C.A.
1976-09-01
The Safeguards Analytical Laboratory Evaluation (SALE) program implements an industry-wide quality control and evaluation system aimed at identifying and reducing analytical chemical measurement errors. Samples of well-characterized materials are distributed to laboratory participants at periodic intervals for determination of uranium or plutonium concentration and isotopic distributions. The results of these determinations are statistically-evaluated, and each participant is informed of the accuracy and precision of his results in a timely manner. The SALE computer code which produces the report is designed to facilitate rapid transmission of this information in order that meaningful quality control will be provided. Various statistical techniques comprise the output of the SALE computer code. Assuming an unbalanced nested design, an analysis of variance is performed in subroutine NEST resulting in a test of significance for time and analyst effects. A trend test is performed in subroutine TREND. Microfilm plots are obtained from subroutine CUMPLT. Within-laboratory standard deviations are calculated in the main program or subroutine VAREST, and between-laboratory standard deviations are calculated in SBLV. Other statistical tests are also performed. Up to 1,500 pieces of data for each nuclear material sampled by 75 (or fewer) laboratories may be analyzed with this code. The input deck necessary to run the program is shown, and input parameters are discussed in detail. Printed output and microfilm plot output are described. Output from a typical SALE run is included as a sample problem.
SALE: Safeguards Analytical Laboratory Evaluation computer code
International Nuclear Information System (INIS)
Carroll, D.J.; Bush, W.J.; Dolan, C.A.
1976-09-01
The Safeguards Analytical Laboratory Evaluation (SALE) program implements an industry-wide quality control and evaluation system aimed at identifying and reducing analytical chemical measurement errors. Samples of well-characterized materials are distributed to laboratory participants at periodic intervals for determination of uranium or plutonium concentration and isotopic distributions. The results of these determinations are statistically-evaluated, and each participant is informed of the accuracy and precision of his results in a timely manner. The SALE computer code which produces the report is designed to facilitate rapid transmission of this information in order that meaningful quality control will be provided. Various statistical techniques comprise the output of the SALE computer code. Assuming an unbalanced nested design, an analysis of variance is performed in subroutine NEST resulting in a test of significance for time and analyst effects. A trend test is performed in subroutine TREND. Microfilm plots are obtained from subroutine CUMPLT. Within-laboratory standard deviations are calculated in the main program or subroutine VAREST, and between-laboratory standard deviations are calculated in SBLV. Other statistical tests are also performed. Up to 1,500 pieces of data for each nuclear material sampled by 75 (or fewer) laboratories may be analyzed with this code. The input deck necessary to run the program is shown, and input parameters are discussed in detail. Printed output and microfilm plot output are described. Output from a typical SALE run is included as a sample problem
A zero-dimensional EXTRAP computer code
International Nuclear Information System (INIS)
Karlsson, P.
1982-10-01
A zero-dimensional computer code has been designed for the EXTRAP experiment to predict the density and the temperature and their dependence upon paramenters such as the plasma current and the filling pressure of neutral gas. EXTRAP is a Z-pinch immersed in a vacuum octupole field and could be either linear or toroidal. In this code the density and temperature are assumed to be constant from the axis up to a breaking point from where they decrease linearly in the radial direction out to the plasma radius. All quantities, however, are averaged over the plasma volume thus giving the zero-dimensional character of the code. The particle, momentum and energy one-fluid equations are solved including the effects of the surrounding neutral gas and oxygen impurities. The code shows that the temperature and density are very sensitive to the shape of the plasma, flatter profiles giving higher temperatures and densities. The temperature, however, is not strongly affected for oxygen concentration less than 2% and is well above the radiation barrier even for higher concentrations. (Author)
SKEMA - A computer code to estimate atmospheric dispersion
International Nuclear Information System (INIS)
Sacramento, A.M. do.
1985-01-01
This computer code is a modified version of DWNWND code, developed in Oak Ridge National Laboratory. The Skema code makes an estimative of concentration in air of a material released in atmosphery, by ponctual source. (C.M.) [pt
A new 3-D integral code for computation of accelerator magnets
International Nuclear Information System (INIS)
Turner, L.R.; Kettunen, L.
1991-01-01
For computing accelerator magnets, integral codes have several advantages over finite element codes; far-field boundaries are treated automatically, and computed field in the bore region satisfy Maxwell's equations exactly. A new integral code employing edge elements rather than nodal elements has overcome the difficulties associated with earlier integral codes. By the use of field integrals (potential differences) as solution variables, the number of unknowns is reduced to one less than the number of nodes. Two examples, a hollow iron sphere and the dipole magnet of Advanced Photon Source injector synchrotron, show the capability of the code. The CPU time requirements are comparable to those of three-dimensional (3-D) finite-element codes. Experiments show that in practice it can realize much of the potential CPU time saving that parallel processing makes possible. 8 refs., 4 figs., 1 tab
Recent progress of an integrated implosion code and modeling of element physics
International Nuclear Information System (INIS)
Nagatomo, H.; Takabe, H.; Mima, K.; Ohnishi, N.; Sunahara, A.; Takeda, T.; Nishihara, K.; Nishiguchu, A.; Sawada, K.
2001-01-01
Physics of the inertial fusion is based on a variety of elements such as compressible hydrodynamics, radiation transport, non-ideal equation of state, non-LTE atomic process, and relativistic laser plasma interaction. In addition, implosion process is not in stationary state and fluid dynamics, energy transport and instabilities should be solved simultaneously. In order to study such complex physics, an integrated implosion code including all physics important in the implosion process should be developed. The details of physics elements should be studied and the resultant numerical modeling should be installed in the integrated code so that the implosion can be simulated with available computer within realistic CPU time. Therefore, this task can be basically separated into two parts. One is to integrate all physics elements into a code, which is strongly related to the development of hydrodynamic equation solver. We have developed 2-D integrated implosion code which solves mass, momentum, electron energy, ion energy, equation of states, laser ray-trace, laser absorption radiation, surface tracing and so on. The reasonable results in simulating Rayleigh-Taylor instability and cylindrical implosion are obtained using this code. The other is code development on each element physics and verification of these codes. We had progress in developing a nonlocal electron transport code and 2 and 3 dimension radiation hydrodynamic code. (author)
The MESORAD dose assessment model: Computer code
International Nuclear Information System (INIS)
Ramsdell, J.V.; Athey, G.F.; Bander, T.J.; Scherpelz, R.I.
1988-10-01
MESORAD is a dose equivalent model for emergency response applications that is designed to be run on minicomputers. It has been developed by the Pacific Northwest Laboratory for use as part of the Intermediate Dose Assessment System in the US Nuclear Regulatory Commission Operations Center in Washington, DC, and the Emergency Management System in the US Department of Energy Unified Dose Assessment Center in Richland, Washington. This volume describes the MESORAD computer code and contains a listing of the code. The technical basis for MESORAD is described in the first volume of this report (Scherpelz et al. 1986). A third volume of the documentation planned. That volume will contain utility programs and input and output files that can be used to check the implementation of MESORAD. 18 figs., 4 tabs
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.
Computer code for quantitative ALARA evaluations
International Nuclear Information System (INIS)
Voilleque, P.G.
1984-01-01
A FORTRAN computer code has been developed to simplify the determination of whether dose reduction actions meet the as low as is reasonably achievable (ALARA) criterion. The calculations are based on the methodology developed for the Atomic Industrial Forum. The code is used for analyses of eight types of dose reduction actions, characterized as follows: reduce dose rate, reduce job frequency, reduce productive working time, reduce crew size, increase administrative dose limit for the task, and increase the workers' time utilization and dose utilization through (a) improved working conditions, (b) basic skill training, or (c) refresher training for special skills. For each type of action, two analysis modes are available. The first is a generic analysis in which the program computes potential benefits (in dollars) for a range of possible improvements, e.g., for a range of lower dose rates. Generic analyses are most useful in the planning stage and for evaluating the general feasibility of alternative approaches. The second is a specific analysis in which the potential annual benefits of a specific level of improvement and the annual implementation cost are compared. The potential benefits reflect savings in operational and societal costs that can be realized if occupational radiation doses are reduced. Because the potential benefits depend upon many variables which characterize the job, the workplace, and the workers, there is no unique relationship between the potential dollar savings and the dose savings. The computer code permits rapid quantitative analyses of alternatives and is a tool that supplements the health physicist's professional judgment. The program output provides a rational basis for decision-making and a record of the assumptions employed
MQRAD, a computer code for synchrotron radiation from quadrupole magnets
International Nuclear Information System (INIS)
Morimoto, Teruhisa.
1984-01-01
The computer code, MQRAD, is developed for the calculation of the synchrotron radiation from the particles passing through quadrupole magnets at the straight section of the electron-positron colliding machine. This code computes the distributions of photon numbers and photon energies at any given points on the beam orbit. In this code, elements such as the quadrupole magnets and the drift spaces can be divided into many sub-elements in order to obtain the results with good accuracy. The synchrotron radiation produced by inserted quadrupole magnets at the interaction region of the electron-positron collider is one of the main background sources to the detector. The masking system against the synchrotron radiation at TRISTAN is very important because of the relatively high beam energy and the long straight section, which are 30 GeV and 100 meters, respectively. MQRAD has been used to design the masking system of the TOPAZ detector and the result is presented here as an example. (author)
A framework for developing finite element codes for multi-disciplinary applications.
Dadvand, Pooyan
2007-01-01
The world of computing simulation has experienced great progresses in recent years and requires more exigent multidisciplinary challenges to satisfy the new upcoming demands. Increasing the importance of solving multi-disciplinary problems makes developers put more attention to these problems and deal with difficulties involved in developing software in this area. Conventional finite element codes have several difficulties in dealing with multi-disciplinary problems. Many of these codes are d...
Development of three-dimensional transport code by the double finite element method
International Nuclear Information System (INIS)
Fujimura, Toichiro
1985-01-01
Development of a three-dimensional neutron transport code by the double finite element method is described. Both of the Galerkin and variational methods are adopted to solve the problem, and then the characteristics of them are compared. Computational results of the collocation method, developed as a technique for the vaviational one, are illustrated in comparison with those of an Ssub(n) code. (author)
APPLICATION FOR DESIGN OF STRUCTURAL ELEMENT USING VISUAL BASIC CODING
T. Thenmozhi; K. Nithya; M. Arun Kumar; M. Ravichandran
2017-01-01
The increasing reliance of engineers on computer software in the performance of their tasks requires engineers, the future professional engineers, must be knowledgeable of sound engineering concepts, updated on the latest computer technology used in the industry and aware of the limitations and capabilities of the computer in solving engineering problems. Computer Methods in Civil Engineering to developed structural design program for design of structural element using Visual Basic. By creati...
Development of Probabilistic Internal Dosimetry Computer Code
Energy Technology Data Exchange (ETDEWEB)
Noh, Siwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kwon, Tae-Eun [Korea Institute of Radiological and Medical Sciences, Seoul (Korea, Republic of); Lee, Jai-Ki [Korean Association for Radiation Protection, Seoul (Korea, Republic of)
2017-02-15
Internal radiation dose assessment involves biokinetic models, the corresponding parameters, measured data, and many assumptions. Every component considered in the internal dose assessment has its own uncertainty, which is propagated in the intake activity and internal dose estimates. For research or scientific purposes, and for retrospective dose reconstruction for accident scenarios occurring in workplaces having a large quantity of unsealed radionuclides, such as nuclear power plants, nuclear fuel cycle facilities, and facilities in which nuclear medicine is practiced, a quantitative uncertainty assessment of the internal dose is often required. However, no calculation tools or computer codes that incorporate all the relevant processes and their corresponding uncertainties, i.e., from the measured data to the committed dose, are available. Thus, the objective of the present study is to develop an integrated probabilistic internal-dose-assessment computer code. First, the uncertainty components in internal dosimetry are identified, and quantitative uncertainty data are collected. Then, an uncertainty database is established for each component. In order to propagate these uncertainties in an internal dose assessment, a probabilistic internal-dose-assessment system that employs the Bayesian and Monte Carlo methods. Based on the developed system, we developed a probabilistic internal-dose-assessment code by using MATLAB so as to estimate the dose distributions from the measured data with uncertainty. Using the developed code, we calculated the internal dose distribution and statistical values (e.g. the 2.5{sup th}, 5{sup th}, median, 95{sup th}, and 97.5{sup th} percentiles) for three sample scenarios. On the basis of the distributions, we performed a sensitivity analysis to determine the influence of each component on the resulting dose in order to identify the major component of the uncertainty in a bioassay. The results of this study can be applied to various
Development of probabilistic internal dosimetry computer code
Noh, Siwan; Kwon, Tae-Eun; Lee, Jai-Ki
2017-02-01
Internal radiation dose assessment involves biokinetic models, the corresponding parameters, measured data, and many assumptions. Every component considered in the internal dose assessment has its own uncertainty, which is propagated in the intake activity and internal dose estimates. For research or scientific purposes, and for retrospective dose reconstruction for accident scenarios occurring in workplaces having a large quantity of unsealed radionuclides, such as nuclear power plants, nuclear fuel cycle facilities, and facilities in which nuclear medicine is practiced, a quantitative uncertainty assessment of the internal dose is often required. However, no calculation tools or computer codes that incorporate all the relevant processes and their corresponding uncertainties, i.e., from the measured data to the committed dose, are available. Thus, the objective of the present study is to develop an integrated probabilistic internal-dose-assessment computer code. First, the uncertainty components in internal dosimetry are identified, and quantitative uncertainty data are collected. Then, an uncertainty database is established for each component. In order to propagate these uncertainties in an internal dose assessment, a probabilistic internal-dose-assessment system that employs the Bayesian and Monte Carlo methods. Based on the developed system, we developed a probabilistic internal-dose-assessment code by using MATLAB so as to estimate the dose distributions from the measured data with uncertainty. Using the developed code, we calculated the internal dose distribution and statistical values ( e.g. the 2.5th, 5th, median, 95th, and 97.5th percentiles) for three sample scenarios. On the basis of the distributions, we performed a sensitivity analysis to determine the influence of each component on the resulting dose in order to identify the major component of the uncertainty in a bioassay. The results of this study can be applied to various situations. In cases of
40 CFR 194.23 - Models and computer codes.
2010-07-01
... 40 Protection of Environment 24 2010-07-01 2010-07-01 false Models and computer codes. 194.23... General Requirements § 194.23 Models and computer codes. (a) Any compliance application shall include: (1... obtain stable solutions; (iv) Computer models accurately implement the numerical models; i.e., computer...
Impact of new computing systems on finite element computations
International Nuclear Information System (INIS)
Noor, A.K.; Fulton, R.E.; Storaasi, O.O.
1983-01-01
Recent advances in computer technology that are likely to impact finite element computations are reviewed. The characteristics of supersystems, highly parallel systems, and small systems (mini and microcomputers) are summarized. The interrelations of numerical algorithms and software with parallel architectures are discussed. A scenario is presented for future hardware/software environment and finite element systems. A number of research areas which have high potential for improving the effectiveness of finite element analysis in the new environment are identified
Business model elements impacting cloud computing adoption
DEFF Research Database (Denmark)
Bogataj, Kristina; Pucihar, Andreja; Sudzina, Frantisek
The paper presents a proposed research framework for identification of business model elements impacting Cloud Computing Adoption. We provide a definition of main Cloud Computing characteristics, discuss previous findings on factors impacting Cloud Computing Adoption, and investigate technology...... adoption theories, such as Diffusion of Innovations, Technology Acceptance Model, Unified Theory of Acceptance and Use of Technology. Further on, at research model for identification of Cloud Computing Adoption factors from a business model perspective is presented. The following business model building...
ICAN Computer Code Adapted for Building Materials
Murthy, Pappu L. N.
1997-01-01
The NASA Lewis Research Center has been involved in developing composite micromechanics and macromechanics theories over the last three decades. These activities have resulted in several composite mechanics theories and structural analysis codes whose applications range from material behavior design and analysis to structural component response. One of these computer codes, the Integrated Composite Analyzer (ICAN), is designed primarily to address issues related to designing polymer matrix composites and predicting their properties - including hygral, thermal, and mechanical load effects. Recently, under a cost-sharing cooperative agreement with a Fortune 500 corporation, Master Builders Inc., ICAN was adapted to analyze building materials. The high costs and technical difficulties involved with the fabrication of continuous-fiber-reinforced composites sometimes limit their use. Particulate-reinforced composites can be thought of as a viable alternative. They are as easily processed to near-net shape as monolithic materials, yet have the improved stiffness, strength, and fracture toughness that is characteristic of continuous-fiber-reinforced composites. For example, particlereinforced metal-matrix composites show great potential for a variety of automotive applications, such as disk brake rotors, connecting rods, cylinder liners, and other hightemperature applications. Building materials, such as concrete, can be thought of as one of the oldest materials in this category of multiphase, particle-reinforced materials. The adaptation of ICAN to analyze particle-reinforced composite materials involved the development of new micromechanics-based theories. A derivative of the ICAN code, ICAN/PART, was developed and delivered to Master Builders Inc. as a part of the cooperative activity.
Computation of Asteroid Proper Elements: Recent Advances
Knežević, Z.
2017-12-01
The recent advances in computation of asteroid proper elements are briefly reviewed. Although not representing real breakthroughs in computation and stability assessment of proper elements, these advances can still be considered as important improvements offering solutions to some practical problems encountered in the past. The problem of getting unrealistic values of perihelion frequency for very low eccentricity orbits is solved by computing frequencies using the frequency-modified Fourier transform. The synthetic resonant proper elements adjusted to a given secular resonance helped to prove the existence of Astraea asteroid family. The preliminary assessment of stability with time of proper elements computed by means of the analytical theory provides a good indication of their poorer performance with respect to their synthetic counterparts, and advocates in favor of ceasing their regular maintenance; the final decision should, however, be taken on the basis of more comprehensive and reliable direct estimate of their individual and sample average deviations from constancy.
Computation of asteroid proper elements: Recent advances
Directory of Open Access Journals (Sweden)
Knežević Z.
2017-01-01
Full Text Available The recent advances in computation of asteroid proper elements are briefly reviewed. Although not representing real breakthroughs in computation and stability assessment of proper elements, these advances can still be considered as important improvements offering solutions to some practical problems encountered in the past. The problem of getting unrealistic values of perihelion frequency for very low eccentricity orbits is solved by computing frequencies using the frequencymodified Fourier transform. The synthetic resonant proper elements adjusted to a given secular resonance helped to prove the existence of Astraea asteroid family. The preliminary assessment of stability with time of proper elements computed by means of the analytical theory provides a good indication of their poorer performance with respect to their synthetic counterparts, and advocates in favor of ceasing their regular maintenance; the final decision should, however, be taken on the basis of more comprehensive and reliable direct estimate of their individual and sample average deviations from constancy.
ADORAVA - A computer code to sum random variables
International Nuclear Information System (INIS)
Fleming, P.V.; Oliveira, L.F.S. de; Senna, V.; Salles, M.R.
1985-06-01
The ADORAVA computer code was carried out aiming to determine the moments of random variable sum distribution when moments are known. The ADORAVA computer code was developed to be applied in probabilistic safety analysis, more specifically for uncertainty propagation in fault trees. The description of ADORAVA algorithm, input, examples and the output of compiled code are presented. (M.C.K.) [pt
Basic elements of computational statistics
Härdle, Wolfgang Karl; Okhrin, Yarema
2017-01-01
This textbook on computational statistics presents tools and concepts of univariate and multivariate statistical data analysis with a strong focus on applications and implementations in the statistical software R. It covers mathematical, statistical as well as programming problems in computational statistics and contains a wide variety of practical examples. In addition to the numerous R sniplets presented in the text, all computer programs (quantlets) and data sets to the book are available on GitHub and referred to in the book. This enables the reader to fully reproduce as well as modify and adjust all examples to their needs. The book is intended for advanced undergraduate and first-year graduate students as well as for data analysts new to the job who would like a tour of the various statistical tools in a data analysis workshop. The experienced reader with a good knowledge of statistics and programming might skip some sections on univariate models and enjoy the various mathematical roots of multivariate ...
ECON-KG: A Code for Computation of Electrical Conductivity Using Density Functional Theory
2017-10-01
materials can be obtained using first-principles quantum mechanical techniques. One approach for computing the conductivity from first principles is...conductivity. Therefore, the Quantum Espresso source code was modified so that momentum matrix elements between all bands are now computed and written to...executable from the Quantum Espresso suite. 2) Using the converged bands, compute the momentum matrix elements using the locally modified (as described
FEAST: a two-dimensional non-linear finite element code for calculating stresses
International Nuclear Information System (INIS)
Tayal, M.
1986-06-01
The computer code FEAST calculates stresses, strains, and displacements. The code is two-dimensional. That is, either plane or axisymmetric calculations can be done. The code models elastic, plastic, creep, and thermal strains and stresses. Cracking can also be simulated. The finite element method is used to solve equations describing the following fundamental laws of mechanics: equilibrium; compatibility; constitutive relations; yield criterion; and flow rule. FEAST combines several unique features that permit large time-steps in even severely non-linear situations. The features include a special formulation for permitting many finite elements to simultaneously cross the boundary from elastic to plastic behaviour; accomodation of large drops in yield-strength due to changes in local temperature and a three-step predictor-corrector method for plastic analyses. These features reduce computing costs. Comparisons against twenty analytical solutions and against experimental measurements show that predictions of FEAST are generally accurate to ± 5%
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.
An Object-Oriented Approach to Writing Computational Electromagnetics Codes
Zimmerman, Martin; Mallasch, Paul G.
1996-01-01
Presently, most computer software development in the Computational Electromagnetics (CEM) community employs the structured programming paradigm, particularly using the Fortran language. Other segments of the software community began switching to an Object-Oriented Programming (OOP) paradigm in recent years to help ease design and development of highly complex codes. This paper examines design of a time-domain numerical analysis CEM code using the OOP paradigm, comparing OOP code and structured programming code in terms of software maintenance, portability, flexibility, and speed.
Hamor-2: a computer code for LWR inventory calculation
International Nuclear Information System (INIS)
Guimaraes, L.N.F.; Marzo, M.A.S.
1985-01-01
A method for calculating the accuracy inventory of LWR reactors is presented. This method uses the Hamor-2 computer code. Hamor-2 is obtained from the coupling of two other computer codes Hammer-Techion and Origen-2 for testing Hamor-2, its results were compared to concentration values measured from activides of two PWR reactors; Kernkraftwerk Obrighein (KWO) and H.B. Robinson (HBR). These actinides are U 235 , U 236 , U 238 , Pu 239 , Pu 241 and PU 242 . The computer code Hammor-2 shows better results than the computer code Origem-2, when both are compared with experimental results. (E.G.) [pt
DEFF Research Database (Denmark)
Henrichsen, Søren Randrup; Lindgaard, Esben; Lund, Erik
2015-01-01
In this work optimum stiffness design of laminated composite structures is performed using the commercially available programs ANSYS and MATLAB. Within these programs a Free Material Optimization algorithm is implemented based on an optimality condition and a heuristic update scheme. The heuristic...... update scheme is needed because commercially available finite element analysis software is used. When using a commercial finite element analysis code it is not straight forward to implement a computationally efficient gradient based optimization algorithm. Examples considered in this work are a clamped......, where full access to the finite element analysis core is granted. This comparison displays the possibility of using commercially available programs for stiffness design of laminated composite structures....
Computer codes validation for conditions of core voiding
International Nuclear Information System (INIS)
Delja, A.; Hawley, P.
2011-01-01
Void generation during a Loss of Coolant Accident (LOCA) in a core of a CANDU reactor is of specific importance because of its strong coupling with reactor neutronics. The use of dynamic behaviour and computer code capability to predict void generation accurately in the temporal and spatial domain of the reactor core is fundamental for the determination of CANDU safety. The Canadian industry has used the RD-14M test facilities for its code validation. The validation exercises for the Canadian computer codes TUF and CATHENA were performed some years ago. Recently, the CNSC has gained access to the USNRC computer code TRACE. This has provided an opportunity to explore the use of this code in CANDU related applications. As a part of regulatory assessment and resolving identified Generic Issues (GI), and in an effort to build independent thermal hydraulic computer codes assessment capability within the CNSC, preliminary validation exercises were performed using the TRACE computer code for an evaluation of the void generation phenomena. The paper presents a preliminary assessment of the TRACE computer code for an RD-14M channel voiding test. It is also a validation exercise of void generation for the TRACE computer code. The accuracy of the obtained results is discussed and compared with previous validation assessments that were done using the CATHENA and TUF codes. (author)
FINELM: a multigroup finite element diffusion code. Part I
International Nuclear Information System (INIS)
Davierwalla, D.M.
1980-12-01
The author presents a two dimensional code for multigroup diffusion using the finite element method. It was realized that the extensive connectivity which contributes significantly to the accuracy, results in a matrix which, although symmetric and positive definite, is wide band and possesses an irregular profile. Hence, it was decided to introduce sparsity techniques into the code. The introduction of the R-Z geometry lead to a great deal of changes in the code since the rotational invariance of the removal matrices in X-Y geometry did not carry over in R-Z geometry. Rectangular elements were introduced to remedy the inability of the triangles to model essentially one dimensional problems such as slab geometry. The matter is discussed briefly in the text in the section on benchmark problems. This report is restricted to the general theory of the triangular elements and to the sparsity techniques viz. incomplete disections. The latter makes the size of the problem that can be handled independent of core memory and dependent only on disc storage capacity which is virtually unlimited. (Auth.)
International Nuclear Information System (INIS)
Soares, P.A.; Sirimarco, L.F.
1984-01-01
SACI-2 is a computer code created to study the dynamic behaviour of a PWR nuclear power plant. To evaluate the quality of its results, SACI-2 was used to recalculate commissioning tests done in BIBLIS-A nuclear power plant and to calculate postulated transients for Angra-2 reactor. The results of SACI-2 computer code from BIBLIS-A showed as much good agreement as those calculated with the KWU Loop 7 computer code for Angra-2. (E.G.) [pt
Finite Element Framework for Computational Fluid Dynamics in FEBio.
Ateshian, Gerard A; Shim, Jay J; Maas, Steve A; Weiss, Jeffrey A
2018-02-01
The mechanics of biological fluids is an important topic in biomechanics, often requiring the use of computational tools to analyze problems with realistic geometries and material properties. This study describes the formulation and implementation of a finite element framework for computational fluid dynamics (CFD) in FEBio, a free software designed to meet the computational needs of the biomechanics and biophysics communities. This formulation models nearly incompressible flow with a compressible isothermal formulation that uses a physically realistic value for the fluid bulk modulus. It employs fluid velocity and dilatation as essential variables: The virtual work integral enforces the balance of linear momentum and the kinematic constraint between fluid velocity and dilatation, while fluid density varies with dilatation as prescribed by the axiom of mass balance. Using this approach, equal-order interpolations may be used for both essential variables over each element, contrary to traditional mixed formulations that must explicitly satisfy the inf-sup condition. The formulation accommodates Newtonian and non-Newtonian viscous responses as well as inviscid fluids. The efficiency of numerical solutions is enhanced using Broyden's quasi-Newton method. The results of finite element simulations were verified using well-documented benchmark problems as well as comparisons with other free and commercial codes. These analyses demonstrated that the novel formulation introduced in FEBio could successfully reproduce the results of other codes. The analogy between this CFD formulation and standard finite element formulations for solid mechanics makes it suitable for future extension to fluid-structure interactions (FSIs).
International Nuclear Information System (INIS)
Weaver, W.L.
1989-01-01
RELAP5/MOD3 is a pressurized water reactor (PWR) system analysis code being developed jointly by the U.S. Nuclear Regulatory Commission (USNRC) and a consortium consisting of several of the countries and domestic organizations that are members of the International Code Assessment and Applications Program (ICAP). The mission of the RELAP5/MOD3 code development program is to develop a code version suitable for the analysis of all transients and postulated accidents in PWR systems including both large and small break loss of coolant accidents (LOCA's) as well as the full range of operational transients. Although the emphasis of the RELAP5/MOD3 development is on large break LOCA, improvements to existing code models, based on the results of assessments against small break LOCA and operational transient test data, are also being made. This paper discusses the new code models as well as improvements to existing models. 13 refs., 6 figs., 1 tab
The Intertwining of Transposable Elements and Non-Coding RNAs
Directory of Open Access Journals (Sweden)
Nicholas Delihas
2013-06-01
Full Text Available Growing evidence shows a close association of transposable elements (TE with non-coding RNAs (ncRNA, and a significant number of small ncRNAs originate from TEs. Further, ncRNAs linked with TE sequences participate in a wide-range of regulatory functions. Alu elements in particular are critical players in gene regulation and molecular pathways. Alu sequences embedded in both long non-coding RNAs (lncRNA and mRNAs form the basis of targeted mRNA decay via short imperfect base-pairing. Imperfect pairing is prominent in most ncRNA/target RNA interactions and found throughout all biological kingdoms. The piRNA-Piwi complex is multifunctional, but plays a major role in protection against invasion by transposons. This is an RNA-based genetic immune system similar to the one found in prokaryotes, the CRISPR system. Thousands of long intergenic non-coding RNAs (lincRNAs are associated with endogenous retrovirus LTR transposable elements in human cells. These TEs can provide regulatory signals for lincRNA genes. A surprisingly large number of long circular ncRNAs have been discovered in human fibroblasts. These serve as “sponges” for miRNAs. Alu sequences, encoded in introns that flank exons are proposed to participate in RNA circularization via Alu/Alu base-pairing. Diseases are increasingly found to have a TE/ncRNA etiology. A single point mutation in a SINE/Alu sequence in a human long non-coding RNA leads to brainstem atrophy and death. On the other hand, genomic clusters of repeat sequences as well as lncRNAs function in epigenetic regulation. Some clusters are unstable, which can lead to formation of diseases such as facioscapulohumeral muscular dystrophy. The future may hold more surprises regarding diseases associated with ncRNAs andTEs.
Control rod computer code IAMCOS: general theory and numerical methods
International Nuclear Information System (INIS)
West, G.
1982-11-01
IAMCOS is a computer code for the description of mechanical and thermal behavior of cylindrical control rods for fast breeders. This code version was applied, tested and modified from 1979 to 1981. In this report are described the basic model (02 version), theoretical definitions and computation methods [fr
Implantation of FRAPCON-2 code in HB computer
International Nuclear Information System (INIS)
Silva, C.F. da.
1987-05-01
The modifications carried out for implanting FRAPCON-2 computer code in the HB DPS-T7 computer are presented. The FRAPCON-2 code calculates thermo-mechanical response during long period of burnup in stationary state for fuel rods of PWR type reactors. (M.C.K.)
The SEDA computer code and its utilization for Angra 1
International Nuclear Information System (INIS)
Fernandes Filho, T.L.
1988-11-01
The implementation of SEDA 2.0 computer code, developed at Ezeiza Atomic Center, Argentine for Angra 1 reactor is described. The SEDA code gives an estimate for radiological consequences of nuclear accidents with release of radiactive materials for the environment. This code is now available for an IBM PC-XT. The computer environment, the files used, data, the programining structure and the models used are presented. The input data and results for two sample case are described. (author) [pt
Adams, M. L.; Padovan, J.; Fertis, D. G.
1981-01-01
NASA-sponsored research on engine dynamic simulation using general finite element nonlinear time transient computer codes available on the open market is reviewed. The approach taken was to develop software packages to model engine components which are not typically found on dynamical structures and are therefore not already computer codes. The software package developed for squeeze-film bearing dampers is outlined, and the results of a parametric study of damper pressure for a variety of specified circular orbits are presented for both long-bearing and short-bearing solutions. The data from a four-degree-of-freedom rotor-damper-stator model under conditions of small rotor unbalance through large rotor unbalance are also given.
Coded Cooperative Computation for Internet of Things
Keshtkarjahromi, Yasaman; Seferoglu, Hulya
2018-01-01
Cooperative computation is a promising approach for localized data processing for Internet of Things (IoT), where computationally intensive tasks in a device could be divided into sub-tasks, and offloaded to other devices or servers in close proximity. However, exploiting the potential of cooperative computation is challenging mainly due to the heterogeneous nature of IoT devices. Indeed, IoT devices may have different and time-varying computing power and energy resources, and could be mobile...
The failure mechanisms of HTR coated particle fuel and computer code
International Nuclear Information System (INIS)
Yang Lin; Liu Bing; Shao Youlin; Liang Tongxiang; Tang Chunhe
2010-01-01
The basic constituent unit of fuel element in HTR is ceramic coated particle fuel. And the performance of coated particle fuel determines the safety of HTR. In addition to the traditional detection of radiation experiments, establishing computer code is of great significance to the research. This paper mainly introduces the structure and the failure mechanism of TRISO-coated particle fuel, as well as a few basic assumptions,principles and characteristics of some existed main overseas codes. Meanwhile, this paper has proposed direction of future research by comparing the advantages and disadvantages of several computer codes. (authors)
Parallel fast multipole boundary element method applied to computational homogenization
Ptaszny, Jacek
2018-01-01
In the present work, a fast multipole boundary element method (FMBEM) and a parallel computer code for 3D elasticity problem is developed and applied to the computational homogenization of a solid containing spherical voids. The system of equation is solved by using the GMRES iterative solver. The boundary of the body is dicretized by using the quadrilateral serendipity elements with an adaptive numerical integration. Operations related to a single GMRES iteration, performed by traversing the corresponding tree structure upwards and downwards, are parallelized by using the OpenMP standard. The assignment of tasks to threads is based on the assumption that the tree nodes at which the moment transformations are initialized can be partitioned into disjoint sets of equal or approximately equal size and assigned to the threads. The achieved speedup as a function of number of threads is examined.
PORPST: A statistical postprocessor for the PORMC computer code
International Nuclear Information System (INIS)
Eslinger, P.W.; Didier, B.T.
1991-06-01
This report describes the theory underlying the PORPST code and gives details for using the code. The PORPST code is designed to do statistical postprocessing on files written by the PORMC computer code. The data written by PORMC are summarized in terms of means, variances, standard deviations, or statistical distributions. In addition, the PORPST code provides for plotting of the results, either internal to the code or through use of the CONTOUR3 postprocessor. Section 2.0 discusses the mathematical basis of the code, and Section 3.0 discusses the code structure. Section 4.0 describes the free-format point command language. Section 5.0 describes in detail the commands to run the program. Section 6.0 provides an example program run, and Section 7.0 provides the references. 11 refs., 1 fig., 17 tabs
Reducing Computational Overhead of Network Coding with Intrinsic Information Conveying
DEFF Research Database (Denmark)
Heide, Janus; Zhang, Qi; Pedersen, Morten V.
is RLNC (Random Linear Network Coding) and the goal is to reduce the amount of coding operations both at the coding and decoding node, and at the same time remove the need for dedicated signaling messages. In a traditional RLNC system, coding operation takes up significant computational resources and adds...... to the overall energy consumption, which is particular problematic for mobile battery-driven devices. In RLNC coding is performed over a FF (Finite Field). We propose to divide this field into sub fields, and let each sub field signify some information or state. In order to embed the information correctly...
Codes of Ethics for Computing at Russian Institutions and Universities.
Pourciau, Lester J.; Spain, Victoria, Ed.
1997-01-01
To determine the degree to which Russian institutions and universities have formulated and promulgated codes of ethics or policies for acceptable computer use, the author examined Russian institution and university home pages. Lists home pages examined, 10 commandments for computer ethics from the Computer Ethics Institute, and a policy statement…
Improvements on binary coding using parallel computing
Fuentes, P A; Quintas, D G
2011-01-01
The error-correcting codes have many applications in fields related to communications. This paper tackles some partition algorithms to optimize the data encoding. These algorithms are based on sliding windows and allow for a parallel implementation. We analyse them and then expound a comparative study between the different partition methods that we propose.
Optimization of KINETICS Chemical Computation Code
Donastorg, Cristina
2012-01-01
NASA JPL has been creating a code in FORTRAN called KINETICS to model the chemistry of planetary atmospheres. Recently there has been an effort to introduce Message Passing Interface (MPI) into the code so as to cut down the run time of the program. There has been some implementation of MPI into KINETICS; however, the code could still be more efficient than it currently is. One way to increase efficiency is to send only certain variables to all the processes when an MPI subroutine is called and to gather only certain variables when the subroutine is finished. Therefore, all the variables that are used in three of the main subroutines needed to be investigated. Because of the sheer amount of code that there is to comb through this task was given as a ten-week project. I have been able to create flowcharts outlining the subroutines, common blocks, and functions used within the three main subroutines. From these flowcharts I created tables outlining the variables used in each block and important information about each. All this information will be used to determine how to run MPI in KINETICS in the most efficient way possible.
Development Of The Computer Code For Comparative Neutron Activation Analysis
International Nuclear Information System (INIS)
Purwadi, Mohammad Dhandhang
2001-01-01
The qualitative and quantitative chemical analysis with Neutron Activation Analysis (NAA) is an importance utilization of a nuclear research reactor, and this should be accelerated and promoted in application and its development to raise the utilization of the reactor. The application of Comparative NAA technique in GA Siwabessy Multi Purpose Reactor (RSG-GAS) needs special (not commercially available yet) soft wares for analyzing the spectrum of multiple elements in the analysis at once. The application carried out using a single spectrum software analyzer, and comparing each result manually. This method really degrades the quality of the analysis significantly. To solve the problem, a computer code was designed and developed for comparative NAA. Spectrum analysis in the code is carried out using a non-linear fitting method. Before the spectrum analyzed, it was passed to the numerical filter which improves the signal to noise ratio to do the deconvolution operation. The software was developed using the G language and named as PASAN-K The testing result of the developed software was benchmark with the IAEA spectrum and well operated with less than 10 % deviation
High burnup models in computer code fair
International Nuclear Information System (INIS)
Dutta, B.K.; Swami Prasad, P.; Kushwaha, H.S.; Mahajan, S.C.; Kakodar, A.
1997-01-01
An advanced fuel analysis code FAIR has been developed for analyzing the behavior of fuel rods of water cooled reactors under severe power transients and high burnups. The code is capable of analyzing fuel pins of both collapsible clad, as in PHWR and free standing clad as in LWR. The main emphasis in the development of this code is on evaluating the fuel performance at extended burnups and modelling of the fuel rods for advanced fuel cycles. For this purpose, a number of suitable models have been incorporated in FAIR. For modelling the fission gas release, three different models are implemented, namely Physically based mechanistic model, the standard ANS 5.4 model and the Halden model. Similarly the pellet thermal conductivity can be modelled by the MATPRO equation, the SIMFUEL relation or the Halden equation. The flux distribution across the pellet is modelled by using the model RADAR. For modelling pellet clad interaction (PCMI)/ stress corrosion cracking (SCC) induced failure of sheath, necessary routines are provided in FAIR. The validation of the code FAIR is based on the analysis of fuel rods of EPRI project ''Light water reactor fuel rod modelling code evaluation'' and also the analytical simulation of threshold power ramp criteria of fuel rods of pressurized heavy water reactors. In the present work, a study is carried out by analysing three CRP-FUMEX rods to show the effect of various combinations of fission gas release models and pellet conductivity models, on the fuel analysis parameters. The satisfactory performance of FAIR may be concluded through these case studies. (author). 12 refs, 5 figs
Simulation of Aircraft Landing Gears with a Nonlinear Dynamic Finite Element Code
Lyle, Karen H.; Jackson, Karen E.; Fasanella, Edwin L.
2000-01-01
Recent advances in computational speed have made aircraft and spacecraft crash simulations using an explicit, nonlinear, transient-dynamic, finite element analysis code more feasible. This paper describes the development of a simple landing gear model, which accurately simulates the energy absorbed by the gear without adding substantial complexity to the model. For a crash model, the landing gear response is approximated with a spring where the force applied to the fuselage is computed in a user-written subroutine. Helicopter crash simulations using this approach are compared with previously acquired experimental data from a full-scale crash test of a composite helicopter.
Talking about Code: Integrating Pedagogical Code Reviews into Early Computing Courses
Hundhausen, Christopher D.; Agrawal, Anukrati; Agarwal, Pawan
2013-01-01
Given the increasing importance of soft skills in the computing profession, there is good reason to provide students withmore opportunities to learn and practice those skills in undergraduate computing courses. Toward that end, we have developed an active learning approach for computing education called the "Pedagogical Code Review"…
Code 672 observational science branch computer networks
Hancock, D. W.; Shirk, H. G.
1988-01-01
In general, networking increases productivity due to the speed of transmission, easy access to remote computers, ability to share files, and increased availability of peripherals. Two different networks within the Observational Science Branch are described in detail.
Lattice Boltzmann method fundamentals and engineering applications with computer codes
Mohamad, A A
2014-01-01
Introducing the Lattice Boltzmann Method in a readable manner, this book provides detailed examples with complete computer codes. It avoids the most complicated mathematics and physics without scarifying the basic fundamentals of the method.
Theoretical calculation possibilities of the computer code HAMMER
International Nuclear Information System (INIS)
Onusic Junior, J.
1978-06-01
With the aim to know the theoretical calculation possibilities of the computer code HAMMER, developed at Savanah River Laboratory, a analysis of the crytical cells assembly of the kind utilized in PWR reactors is made. (L.F.S.) [pt
Study of nuclear computer code maintenance and management system
International Nuclear Information System (INIS)
Ryu, Chang Mo; Kim, Yeon Seung; Eom, Heung Seop; Lee, Jong Bok; Kim, Ho Joon; Choi, Young Gil; Kim, Ko Ryeo
1989-01-01
Software maintenance is one of the most important problems since late 1970's.We wish to develop a nuclear computer code system to maintenance and manage KAERI's nuclear software. As a part of this system, we have developed three code management programs for use on CYBER and PC systems. They are used in systematic management of computer code in KAERI. The first program is embodied on the CYBER system to rapidly provide information on nuclear codes to the users. The second and the third programs were embodied on the PC system for the code manager and for the management of data in korean language, respectively. In the requirement analysis, we defined each code, magnetic tape, manual and abstract information data. In the conceptual design, we designed retrieval, update, and output functions. In the implementation design, we described the technical considerations of database programs, utilities, and directions for the use of databases. As a result of this research, we compiled the status of nuclear computer codes which belonged KAERI until September, 1988. Thus, by using these three database programs, we could provide the nuclear computer code information to the users more rapidly. (Author)
Nuclear model computer codes available from the NEA Data Bank
International Nuclear Information System (INIS)
Sartori, E.
1989-01-01
A library of computer codes for nuclear model calculations, a subset of a library covering the different aspects of reactor physics and technology applications has been established at the NEA Data Bank. These codes are listed and classified according to the model used in the text. Copies of the programs can be obtained from the NEA Data Bank. (author). 8 refs
User manual of FRAPCON-I computer code
International Nuclear Information System (INIS)
Chia, C.T.
1985-11-01
The manual for using the FRAPCON-I code implanted by Reactor Department of Brazilian-CNEN to convert IBM FORTRAN in FORTRAN 77 of Honeywell Bull computer is presented. The FRAPCON-I code describes the behaviour of fuel rods of PWR type reactors at stationary state during long periods of burnup. (M.C.K.)
Computer codes for level 1 probabilistic safety assessment
International Nuclear Information System (INIS)
1990-06-01
Probabilistic Safety Assessment (PSA) entails several laborious tasks suitable for computer codes assistance. This guide identifies these tasks, presents guidelines for selecting and utilizing computer codes in the conduct of the PSA tasks and for the use of PSA results in safety management and provides information on available codes suggested or applied in performing PSA in nuclear power plants. The guidance is intended for use by nuclear power plant system engineers, safety and operating personnel, and regulators. Large efforts are made today to provide PC-based software systems and PSA processed information in a way to enable their use as a safety management tool by the nuclear power plant overall management. Guidelines on the characteristics of software needed for management to prepare a software that meets their specific needs are also provided. Most of these computer codes are also applicable for PSA of other industrial facilities. The scope of this document is limited to computer codes used for the treatment of internal events. It does not address other codes available mainly for the analysis of external events (e.g. seismic analysis) flood and fire analysis. Codes discussed in the document are those used for probabilistic rather than for phenomenological modelling. It should be also appreciated that these guidelines are not intended to lead the user to selection of one specific code. They provide simply criteria for the selection. Refs and tabs
Reasoning with Computer Code: a new Mathematical Logic
Pissanetzky, Sergio
2013-01-01
A logic is a mathematical model of knowledge used to study how we reason, how we describe the world, and how we infer the conclusions that determine our behavior. The logic presented here is natural. It has been experimentally observed, not designed. It represents knowledge as a causal set, includes a new type of inference based on the minimization of an action functional, and generates its own semantics, making it unnecessary to prescribe one. This logic is suitable for high-level reasoning with computer code, including tasks such as self-programming, objectoriented analysis, refactoring, systems integration, code reuse, and automated programming from sensor-acquired data. A strong theoretical foundation exists for the new logic. The inference derives laws of conservation from the permutation symmetry of the causal set, and calculates the corresponding conserved quantities. The association between symmetries and conservation laws is a fundamental and well-known law of nature and a general principle in modern theoretical Physics. The conserved quantities take the form of a nested hierarchy of invariant partitions of the given set. The logic associates elements of the set and binds them together to form the levels of the hierarchy. It is conjectured that the hierarchy corresponds to the invariant representations that the brain is known to generate. The hierarchies also represent fully object-oriented, self-generated code, that can be directly compiled and executed (when a compiler becomes available), or translated to a suitable programming language. The approach is constructivist because all entities are constructed bottom-up, with the fundamental principles of nature being at the bottom, and their existence is proved by construction. The new logic is mathematically introduced and later discussed in the context of transformations of algorithms and computer programs. We discuss what a full self-programming capability would really mean. We argue that self
Two-phase computer codes for zero-gravity applications
International Nuclear Information System (INIS)
Krotiuk, W.J.
1986-10-01
This paper discusses the problems existing in the development of computer codes which can analyze the thermal-hydraulic behavior of two-phase fluids especially in low gravity nuclear reactors. The important phenomenon affecting fluid flow and heat transfer in reduced gravity is discussed. The applicability of using existing computer codes for space applications is assessed. Recommendations regarding the use of existing earth based fluid flow and heat transfer correlations are made and deficiencies in these correlations are identified
Development of computer code in PNC, 7
International Nuclear Information System (INIS)
Shibata, Tomofumi
1990-01-01
The phenomenon of bending vibrations of rotating shafts in perhaps the most common problem that is discussed by a vibration engineer. This paper presents a Transfer Matrix's general formulation for synchronous unbalance response effects. All the transfer Matrix are reformed comparing with using complex type elements matrix. (author)
Analysis of piping systems by finite element method using code SAP-IV
International Nuclear Information System (INIS)
Cizelj, L.; Ogrizek, D.
1987-01-01
Due to extensive and multiple use of the computer code SAP-IV we have decided to install it on VAX 11/750 machine. Installation required a large quantity of programming due to great discrepancies between the CDC (the original program version) and the VAX. Testing was performed basically in the field of pipe elements, based on a comparison between results obtained with the codes PSAFE2, DOCIJEV, PIPESD and SAP -V. Besides, the model of reactor pressure vessel with 3-D thick shell elements was done. The capabilities show good agreement with the results of other programs mentioned above. Along with the package installation, the graphical postprocessors being developed for mesh plotting. (author)
Computer code ANISN multiplying media and shielding calculation II. Code description (input/output)
International Nuclear Information System (INIS)
Maiorino, J.R.
1990-01-01
The user manual of the ANISN computer code describing input and output subroutines is presented. ANISN code was developed to solve one-dimensional transport equation for neutron or gamma rays in slab, sphere or cylinder geometry with general anisotropic scattering. The solution technique is the discrete ordinate method. (M.C.K.)
A restructuring of CF package for MIDAS computer code
International Nuclear Information System (INIS)
Park, S. H.; Kim, K. R.; Kim, D. H.; Cho, S. W.
2004-01-01
CF package, which evaluates user-specified 'control functions' and applies them to define or control various aspects of computation, has been restructured for the MIDAS computer code. MIDAS is being developed as an integrated severe accident analysis code with a user-friendly graphical user interface and modernized data structure. To do this, data transferring methods of current MELCOR code are modified and adopted into the CF package. The data structure of the current MELCOR code using FORTRAN77 causes a difficult grasping of meaning of the variables as well as waste of memory, difficulty is more over because its data is location information of other package's data due to characteristics of CF package. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type, which lead to an efficient memory treatment and an easy understanding of the code. Restructuring of the CF package addressed in this paper includes module development, subroutine modification, and treats MELGEN, which generates data file, as well as MELCOR, which is processing a calculation. The verification has been done by comparing the results of the modified code with those from the existing code. As the trends are similar to each other, it hints that the same approach could be extended to the entire code package. It is expected that code restructuring will accelerate the code domestication thanks to direct understanding of each variable and easy implementation of modified or newly developed models
The archaeology of computer codes - illustrated on the basis of the code SABINE
International Nuclear Information System (INIS)
Sdouz, G.
1987-02-01
Computer codes used by the physics group of the Institute for Reactor Safety are stored on back-up-tapes. However during the last years both the computer and the system have been changed. For new tasks these programmes have to be available. A new procedure is necessary to find and to activate a stored programme. This procedure is illustrated on the basis of the code SABINE. (Author)
Computer codes incorporating pre-equilibrium decay
International Nuclear Information System (INIS)
Prince, A.
1980-01-01
After establishing the need to describe the high-energy particle spectrum which is evident in the experimental data, the various models used in the interpretation are presented. This includes the following: a) Cascade Model; b) Fermi-Gas Relaxation Model; c) Exciton Model; d) Hybrid and Geometry-Dependent Model. The codes description and preparation of input data for STAPRE was presented (Dr. Strohmaier). A simulated output was employed for a given input and comparison with experimental data substantiated the rather sophisticated treatment. (author)
Adaptation of HAMMER computer code to CYBER 170/750 computer
International Nuclear Information System (INIS)
Pinheiro, A.M.B.S.; Nair, R.P.K.
1982-01-01
The adaptation of HAMMER computer code to CYBER 170/750 computer is presented. The HAMMER code calculates cell parameters by multigroup transport theory and reactor parameters by few group diffusion theory. The auxiliary programs, the carried out modifications and the use of HAMMER system adapted to CYBER 170/750 computer are described. (M.C.K.) [pt
Verdon: code of mechanical and thermal behavior of fuel element
International Nuclear Information System (INIS)
Courtois, C.; Truffert, J.
1979-01-01
Verdon code must be used for analysis and simulation of mechanical, two-dimensional, thermal and physico-chemical behavior of fuel oxide pin in steady-state and transient conditions. Calculation can be done in plane or axisymmetric geometry. Radial, one dimensional, thermal analysis works with finite differences. It takes into account the fissile material's evolution (radial redistribution, flux deepening ...) and the main fuel physico-chemical properties (conductivity, migration, fission gas release ...). Only thermal consequences of fuel mechanical behavior: fuel-cladding gap width, crack formation, creep ... are submitted to a two dimensional analysis. Mechanical analysis works in two dimensional, finite elements, plane or axisymmetric geometry. The mesh represents a part of fuel and cladding pin [fr
User's manual for the vertical axis wind turbine performance computer code darter
Klimas, P. C.; French, R. E.
1980-05-01
The computer code DARTER (DARrieus, Turbine, Elemental Reynolds number) is an aerodynamic performance/loads prediction scheme based upon the conservation of momentum principle. It is the latest evolution in a sequence which began with a model developed by Templin of NRC, Canada and progressed through the Sandia National Laboratories developed SIMOSS (SImple MOmentum, Single Streamtube) and DART (Darrieus Turbine) to DARTER.
The community project COSA: comparison of geo-mechanical computer codes for salt
International Nuclear Information System (INIS)
Lowe, M.J.S.; Knowles, N.C.
1986-01-01
Two benchmark problems related to waste disposal in salt were tackled by ten European organisations using twelve rock-mechanics finite element computer codes. The two problems represented increasing complexity with first a hypothetical verification and then the simulation of a laboratory experiment. The project allowed to ascertain a shapshot of the current combined expertise of European organisations in the modelling of salt behaviour
Case studies in Gaussian process modelling of computer codes
International Nuclear Information System (INIS)
Kennedy, Marc C.; Anderson, Clive W.; Conti, Stefano; O'Hagan, Anthony
2006-01-01
In this paper we present a number of recent applications in which an emulator of a computer code is created using a Gaussian process model. Tools are then applied to the emulator to perform sensitivity analysis and uncertainty analysis. Sensitivity analysis is used both as an aid to model improvement and as a guide to how much the output uncertainty might be reduced by learning about specific inputs. Uncertainty analysis allows us to reflect output uncertainty due to unknown input parameters, when the finished code is used for prediction. The computer codes themselves are currently being developed within the UK Centre for Terrestrial Carbon Dynamics
Morse-code-controlled computer aid for the nonvocal quadriplegic.
Shannon, D A; Staewen, W S; Miller, J T; Cohen, B S
1981-01-01
A computer-based communications system was established for a quadriplegic patient who suffered a brainstem infarction that left him with normal intelligence, but without speech and without functional motor control except for the thumb of his right hand. An arm brace was developed to support the hand and thumb so it could control two switches that generated Morse code signals. An interface was also devised to translate Morse code into ASCII characters to access a personal computer. Additional codes were added to represent keyboard control functions. This device allows the patient to communicate with others and control, to some extent, his environment.
Computer codes for beam dynamics analysis of cyclotronlike accelerators
Smirnov, V.
2017-12-01
Computer codes suitable for the study of beam dynamics in cyclotronlike (classical and isochronous cyclotrons, synchrocyclotrons, and fixed field alternating gradient) accelerators are reviewed. Computer modeling of cyclotron segments, such as the central zone, acceleration region, and extraction system is considered. The author does not claim to give a full and detailed description of the methods and algorithms used in the codes. Special attention is paid to the codes already proven and confirmed at the existing accelerating facilities. The description of the programs prepared in the worldwide known accelerator centers is provided. The basic features of the programs available to users and limitations of their applicability are described.
Multitasking the code ARC3D. [for computational fluid dynamics
Barton, John T.; Hsiung, Christopher C.
1986-01-01
The CRAY multitasking system was developed in order to utilize all four processors and sharply reduce the wall clock run time. This paper describes the techniques used to modify the computational fluid dynamics code ARC3D for this run and analyzes the achieved speedup. The ARC3D code solves either the Euler or thin-layer N-S equations using an implicit approximate factorization scheme. Results indicate that multitask processing can be used to achieve wall clock speedup factors of over three times, depending on the nature of the program code being used. Multitasking appears to be particularly advantageous for large-memory problems running on multiple CPU computers.
Code system to compute radiation dose in human phantoms
International Nuclear Information System (INIS)
Ryman, J.C.; Cristy, M.; Eckerman, K.F.; Davis, J.L.; Tang, J.S.; Kerr, G.D.
1986-01-01
Monte Carlo photon transport code and a code using Monte Carlo integration of a point kernel have been revised to incorporate human phantom models for an adult female, juveniles of various ages, and a pregnant female at the end of the first trimester of pregnancy, in addition to the adult male used earlier. An analysis code has been developed for deriving recommended values of specific absorbed fractions of photon energy. The computer code system and calculational method are described, emphasizing recent improvements in methods
Thermohydraulic analysis of nuclear power plant accidents by computer codes
International Nuclear Information System (INIS)
Petelin, S.; Stritar, A.; Istenic, R.; Gregoric, M.; Jerele, A.; Mavko, B.
1982-01-01
RELAP4/MOD6, BRUCH-D-06, CONTEMPT-LT-28, RELAP5/MOD1 and COBRA-4-1 codes were successful y implemented at the CYBER 172 computer in Ljubljana. Input models of NPP Krsko for the first three codes were prepared. Because of the high computer cost only one analysis of double ended guillotine break of the cold leg of NPP Krsko by RELAP4 code has been done. BRUCH code is easier and cheaper for use. Several analysis have been done. Sensitivity study was performed with CONTEMPT-LT-28 for double ended pump suction break. These codes are intended to be used as a basis for independent safety analyses. (author)
Computational structural analysis and finite element methods
Kaveh, A
2014-01-01
Graph theory gained initial prominence in science and engineering through its strong links with matrix algebra and computer science. Moreover, the structure of the mathematics is well suited to that of engineering problems in analysis and design. The methods of analysis in this book employ matrix algebra, graph theory and meta-heuristic algorithms, which are ideally suited for modern computational mechanics. Efficient methods are presented that lead to highly sparse and banded structural matrices. The main features of the book include: application of graph theory for efficient analysis; extension of the force method to finite element analysis; application of meta-heuristic algorithms to ordering and decomposition (sparse matrix technology); efficient use of symmetry and regularity in the force method; and simultaneous analysis and design of structures.
Theory Manual for ISAAC Computer Code
International Nuclear Information System (INIS)
Park, Soo Yong; Kim, Dong Ha; Song, Yong Mann; Jin, Young Ho; Kim, See Darl; Choi, Young
2008-12-01
Major models adopted in ISAAC are introduced briefly. The primary heat transport system of two independent figure-of-eight loops are represented in ISAAC. All four steam generators and 4 pumps are also modeled individually. PHTS model tracks the masses and energy in one gas space and in multiple water pools. The pressurizer model is similar to the PWR model in MAAP4, except two independent surge lines which are connected to each PHTS loop. The two-region steam generator model was newly implemented into the ISAAC code. In each region, masses and energies of water, steam, and non-condensable gases are tracked. Then, the pressure, water temperature, and gas temperature in each region are calculated based on the masses and energies, using non-equilibrium thermodynamic model. The core heatup module calculates the thermal-hydraulic response and fission product transport, including the rates-of-change of dynamic variables, within the core region. The key quantities of calandria tank model include thermal-hydraulic variables, modeling of corium, water, and calandria tank wall heat transfer, corium debris bed in the bottom of the tank, failure mechanisms, and fission product transport. ISAAC also models main safety features in Wolsong plants as well as fission product behavior. As described, ISAAC has a fundamental models to capture the main phenomena during the severe accident in Wolsong plants
An improved UO2 thermal conductivity model in the ELESTRES computer code
International Nuclear Information System (INIS)
Chassie, G.G.; Tochaie, M.; Xu, Z.
2010-01-01
This paper describes the improved UO 2 thermal conductivity model for use in the ELESTRES (ELEment Simulation and sTRESses) computer code. The ELESTRES computer code models the thermal, mechanical and microstructural behaviour of a CANDU® fuel element under normal operating conditions. The main purpose of the code is to calculate fuel temperatures, fission gas release, internal gas pressure, fuel pellet deformation, and fuel sheath strains for fuel element design and assessment. It is also used to provide initial conditions for evaluating fuel behaviour during high temperature transients. The thermal conductivity of UO 2 fuel is one of the key parameters that affect ELESTRES calculations. The existing ELESTRES thermal conductivity model has been assessed and improved based on a large amount of thermal conductivity data from measurements of irradiated and un-irradiated UO 2 fuel with different densities. The UO 2 thermal conductivity data cover 90% to 99% theoretical density of UO 2 , temperature up to 3027 K, and burnup up to 1224 MW·h/kg U. The improved thermal conductivity model, which is recommended for a full implementation in the ELESTRES computer code, has reduced the ELESTRES code prediction biases of temperature, fission gas release, and fuel sheath strains when compared with the available experimental data. This improved thermal conductivity model has also been checked with a test version of ELESTRES over the full ranges of fuel temperature, fuel burnup, and fuel density expected in CANDU fuel. (author)
A compendium of computer codes in fault tree analysis
International Nuclear Information System (INIS)
Lydell, B.
1981-03-01
In the past ten years principles and methods for a unified system reliability and safety analysis have been developed. Fault tree techniques serve as a central feature of unified system analysis, and there exists a specific discipline within system reliability concerned with the theoretical aspects of fault tree evaluation. Ever since the fault tree concept was established, computer codes have been developed for qualitative and quantitative analyses. In particular the presentation of the kinetic tree theory and the PREP-KITT code package has influenced the present use of fault trees and the development of new computer codes. This report is a compilation of some of the better known fault tree codes in use in system reliability. Numerous codes are available and new codes are continuously being developed. The report is designed to address the specific characteristics of each code listed. A review of the theoretical aspects of fault tree evaluation is presented in an introductory chapter, the purpose of which is to give a framework for the validity of the different codes. (Auth.)
The computer code EURDYN-1MP (release 1) for plane fluid-structure systems: user's manual
International Nuclear Information System (INIS)
Giuliani, S.
1984-01-01
EURDYN-1MP is a finite element computer code developed at J.R.C. Ispra to compute the response of two-dimensional, coupled fluid-structure configurations to transient dynamic loading. The structural domain is assumed to be in a situation of plane strain, but axial motion of the fluid is considered. This report gives instructions for preparing input data to EURDYN-1MP, release 1, and describes a test problem in order to illustrate both the input and the output of the code
Users manual for CAFE-3D : a computational fluid dynamics fire code.
Energy Technology Data Exchange (ETDEWEB)
Khalil, Imane; Lopez, Carlos; Suo-Anttila, Ahti Jorma (Alion Science and Technology, Albuquerque, NM)
2005-03-01
The Container Analysis Fire Environment (CAFE) computer code has been developed to model all relevant fire physics for predicting the thermal response of massive objects engulfed in large fires. It provides realistic fire thermal boundary conditions for use in design of radioactive material packages and in risk-based transportation studies. The CAFE code can be coupled to commercial finite-element codes such as MSC PATRAN/THERMAL and ANSYS. This coupled system of codes can be used to determine the internal thermal response of finite element models of packages to a range of fire environments. This document is a user manual describing how to use the three-dimensional version of CAFE, as well as a description of CAFE input and output parameters. Since this is a user manual, only a brief theoretical description of the equations and physical models is included.
Users manual for CAFE-3D : a computational fluid dynamics fire code
International Nuclear Information System (INIS)
Khalil, Imane; Lopez, Carlos; Suo-Anttila, Ahti Jorma
2005-01-01
The Container Analysis Fire Environment (CAFE) computer code has been developed to model all relevant fire physics for predicting the thermal response of massive objects engulfed in large fires. It provides realistic fire thermal boundary conditions for use in design of radioactive material packages and in risk-based transportation studies. The CAFE code can be coupled to commercial finite-element codes such as MSC PATRAN/THERMAL and ANSYS. This coupled system of codes can be used to determine the internal thermal response of finite element models of packages to a range of fire environments. This document is a user manual describing how to use the three-dimensional version of CAFE, as well as a description of CAFE input and output parameters. Since this is a user manual, only a brief theoretical description of the equations and physical models is included
Computer Security: better code, fewer problems
Stefan Lueders, Computer Security Team
2016-01-01
The origin of many security incidents is negligence or unintentional mistakes made by web developers or programmers. In the rush to complete the work, due to skewed priorities, or just to ignorance, basic security principles can be omitted or forgotten. The resulting vulnerabilities lie dormant until the evil side spots them and decides to hit hard. Computer security incidents in the past have put CERN’s reputation at risk due to websites being defaced with negative messages about the Organization, hash files of passwords being extracted, restricted data exposed… And it all started with a little bit of negligence! If you check out the Top 10 web development blunders, you will see that the most prevalent mistakes are: Not filtering input, e.g. accepting “<“ or “>” in input fields even if only a number is expected. Not validating that input: you expect a birth date? So why accept letters? &...
Development of computer code in PNC, 8
International Nuclear Information System (INIS)
Ohhira, Mitsuru
1990-01-01
Private buildings applied base isolation system, are on the practical stage now. So, under Construction and Maintenance Management Office, we are doing an application study of base isolation system to nuclear fuel facilities. On the process of this study, we have developed Dynamic Analysis Program-Base Isolation System (DAP-BS) which is able to run a 32-bit personal computer. Using this program, we can analyze a 3-dimensional structure, and evaluate the various properties of base isolation parts that are divided into maximum 16 blocks. And from the results of some simulation analyses, we thought that DAP-BS had good reliability and marketability. So, we put DAP-BS on the market. (author)
Turbulence statistics in a spectral element code: a toolbox for High-Fidelity Simulations
Energy Technology Data Exchange (ETDEWEB)
Vinuesa, Ricardo [KTH Mechanics, Stockholm (Sweden); Swedish e-Science Research Center (SeRC), Stockholm (Sweden); Fick, Lambert [Argonne National Lab. (ANL), Argonne, IL (United States); Negi, Prabal [KTH Mechanics, Stockholm (Sweden); Swedish e-Science Research Center (SeRC), Stockholm (Sweden); Marin, Oana [Argonne National Lab. (ANL), Argonne, IL (United States); Merzari, Elia [Argonne National Lab. (ANL), Argonne, IL (United States); Schlatter, Phillip [KTH Mechanics, Stockholm (Sweden); Swedish e-Science Research Center (SeRC), Stockholm (Sweden)
2017-02-01
In the present document we describe a toolbox for the spectral-element code Nek5000, aimed at computing turbulence statistics. The toolbox is presented for a small test case, namely a square duct with L_{x} = 2h, L_{y} = 2h and L_{z} = 4h, where x, y and z are the horizontal, vertical and streamwise directions, respectively. The number of elements in the xy-plane is 16 X 16 = 256, and the number of elements in z is 4, leading to a total of 1,204 spectral elements. A polynomial order of N = 5 is chosen, and the mesh is generated using the Nek5000 tool genbox. The toolbox presented here allows to compute mean-velocity components, the Reynolds-stress tensor as well as turbulent kinetic energy (TKE) and Reynolds-stress budgets. Note that the present toolbox allows to compute turbulence statistics in turbulent flows with one homogeneous direction (where the statistics are based on time-averaging as well as averaging in the homogeneous direction), as well as in fully three-dimensional flows (with no periodic directions, where only time-averaging is considered).
RADTRAN: a computer code to analyze transportation of radioactive material
International Nuclear Information System (INIS)
Taylor, J.M.; Daniel, S.L.
1977-04-01
A computer code is presented which predicts the environmental impact of any specific scheme of radioactive material transportation. Results are presented in terms of annual latent cancer fatalities and annual early fatility probability resulting from exposure, during normal transportation or transport accidents. The code is developed in a generalized format to permit wide application including normal transportation analysis; consideration of alternatives; and detailed consideration of specific sectors of industry
Computer codes for problems of isotope and radiation research
International Nuclear Information System (INIS)
Remer, M.
1986-12-01
A survey is given of computer codes for problems in isotope and radiation research. Altogether 44 codes are described as titles with abstracts. 17 of them are in the INIS scope and are processed individually. The subjects are indicated in the chapter headings: 1) analysis of tracer experiments, 2) spectrum calculations, 3) calculations of ion and electron trajectories, 4) evaluation of gamma irradiation plants, and 5) general software
Computer code for calculating personnel doses due to tritium exposures
International Nuclear Information System (INIS)
Graham, C.L.; Parlagreco, J.R.
1977-01-01
This report describes a computer code written in LLL modified Fortran IV that can be used on a CDC 7600 for calculating personnel doses due to internal exposures to tritium. The code is capable of handling various exposure situations and is also capable of detecting a large variety of data input errors that would lead to errors in the dose assessment. The critical organ is the body water
User's manual for the NEFTRAN II computer code
International Nuclear Information System (INIS)
Olague, N.E.; Campbell, J.E.; Leigh, C.D.; Longsine, D.E.
1991-02-01
This document describes the NEFTRAN II (NEtwork Flow and TRANsport in Time-Dependent Velocity Fields) computer code and is intended to provide the reader with sufficient information to use the code. NEFTRAN II was developed as part of a performance assessment methodology for storage of high-level nuclear waste in unsaturated, welded tuff. NEFTRAN II is a successor to the NEFTRAN and NWFT/DVM computer codes and contains several new capabilities. These capabilities include: (1) the ability to input pore velocities directly to the transport model and bypass the network fluid flow model, (2) the ability to transport radionuclides in time-dependent velocity fields, (3) the ability to account for the effect of time-dependent saturation changes on the retardation factor, and (4) the ability to account for time-dependent flow rates through the source regime. In addition to these changes, the input to NEFTRAN II has been modified to be more convenient for the user. This document is divided into four main sections consisting of (1) a description of all the models contained in the code, (2) a description of the program and subprograms in the code, (3) a data input guide and (4) verification and sample problems. Although NEFTRAN II is the fourth generation code, this document is a complete description of the code and reference to past user's manuals should not be necessary. 19 refs., 33 figs., 25 tabs
Compendium of computer codes for the safety analysis of LMFBR's
International Nuclear Information System (INIS)
1975-06-01
A high level of mathematical sophistication is required in the safety analysis of LMFBR's to adequately meet the demands for realism and confidence in all areas of accident consequence evaluation. The numerical solution procedures associated with these analyses are generally so complex and time consuming as to necessitate their programming into computer codes. These computer codes have become extremely powerful tools for safety analysis, combining unique advantages in accuracy, speed and cost. The number, diversity and complexity of LMFBR safety codes in the U. S. has grown rapidly in recent years. It is estimated that over 100 such codes exist in various stages of development throughout the country. It is inevitable that such a large assortment of codes will require rigorous cataloguing and abstracting to aid individuals in identifying what is available. It is the purpose of this compendium to provide such a service through the compilation of code summaries which describe and clarify the status of domestic LMFBR safety codes. (U.S.)
Independent peer review of nuclear safety computer codes
International Nuclear Information System (INIS)
Boyack, B.E.; Jenks, R.P.
1993-01-01
A structured, independent computer code peer-review process has been developed to assist the US Nuclear Regulatory Commission (NRC) and the US Department of Energy in their nuclear safety missions. This paper describes a structured process of independent code peer review, benefits associated with a code-independent peer review, as well as the authors' recent peer-review experience. The NRC adheres to the principle that safety of plant design, construction, and operation are the responsibility of the licensee. Nevertheless, NRC staff must have the ability to independently assess plant designs and safety analyses submitted by license applicants. According to Ref. 1, open-quotes this requires that a sound understanding be obtained of the important physical phenomena that may occur during transients in operating power plants.close quotes The NRC concluded that computer codes are the principal products to open-quotes understand and predict plant response to deviations from normal operating conditionsclose quotes and has developed several codes for that purpose. However, codes cannot be used blindly; they must be assessed and found adequate for the purposes they are intended. A key part of the qualification process can be accomplished through code peer reviews; this approach has been adopted by the NRC
Computer code MLCOSP for multiple-correlation and spectrum analysis with a hybrid computer
International Nuclear Information System (INIS)
Oguma, Ritsuo; Fujii, Yoshio; Usui, Hozumi; Watanabe, Koichi
1975-10-01
Usage of the computer code MLCOSP(Multiple Correlation and Spectrum) developed is described for a hybrid computer installed in JAERI Functions of the hybrid computer and its terminal devices are utilized ingeniously in the code to reduce complexity of the data handling which occurrs in analysis of the multivariable experimental data and to perform the analysis in perspective. Features of the code are as follows; Experimental data can be fed to the digital computer through the analog part of the hybrid computer by connecting with a data recorder. The computed results are displayed in figures, and hardcopies are taken when necessary. Series-messages to the code are shown on the terminal, so man-machine communication is possible. And further the data can be put in through a keyboard, so case study according to the results of analysis is possible. (auth.)
High performance computer code for molecular dynamics simulations
International Nuclear Information System (INIS)
Levay, I.; Toekesi, K.
2007-01-01
Complete text of publication follows. Molecular Dynamics (MD) simulation is a widely used technique for modeling complicated physical phenomena. Since 2005 we are developing a MD simulations code for PC computers. The computer code is written in C++ object oriented programming language. The aim of our work is twofold: a) to develop a fast computer code for the study of random walk of guest atoms in Be crystal, b) 3 dimensional (3D) visualization of the particles motion. In this case we mimic the motion of the guest atoms in the crystal (diffusion-type motion), and the motion of atoms in the crystallattice (crystal deformation). Nowadays, it is common to use Graphics Devices in intensive computational problems. There are several ways to use this extreme processing performance, but never before was so easy to programming these devices as now. The CUDA (Compute Unified Device) Architecture introduced by nVidia Corporation in 2007 is a very useful for every processor hungry application. A Unified-architecture GPU include 96-128, or more stream processors, so the raw calculation performance is 576(!) GFLOPS. It is ten times faster, than the fastest dual Core CPU [Fig.1]. Our improved MD simulation software uses this new technology, which speed up our software and the code run 10 times faster in the critical calculation code segment. Although the GPU is a very powerful tool, it has a strongly paralleled structure. It means, that we have to create an algorithm, which works on several processors without deadlock. Our code currently uses 256 threads, shared and constant on-chip memory, instead of global memory, which is 100 times slower than others. It is possible to implement the total algorithm on GPU, therefore we do not need to download and upload the data in every iteration. On behalf of maximal throughput, every thread run with the same instructions
Prodeto, a computer code for probabilistic fatigue design
Energy Technology Data Exchange (ETDEWEB)
Braam, H. [ECN-Solar and Wind Energy, Petten (Netherlands); Christensen, C.J.; Thoegersen, M.L. [Risoe National Lab., Roskilde (Denmark); Ronold, K.O. [Det Norske Veritas, Hoevik (Norway)
1999-03-01
A computer code for structural relibility analyses of wind turbine rotor blades subjected to fatigue loading is presented. With pre-processors that can transform measured and theoretically predicted load series to load range distributions by rain-flow counting and with a family of generic distribution models for parametric representation of these distribution this computer program is available for carying through probabilistic fatigue analyses of rotor blades. (au)
Groundwater flow analyses in preliminary site investigations. Modelling strategy and computer codes
International Nuclear Information System (INIS)
Taivassalo, V.; Koskinen, L.; Meling, K.
1994-02-01
The analyses of groundwater flow comprised a part of the preliminary site investigations which were carried out by Teollisuuden Voima Oy (TVO) for five areas in Finland during 1987 -1992. The main objective of the flow analyses was to characterize groundwater flow at the sites. The flow simulations were also used to identify and study uncertainties and inadequacies which are inherent in the results of earlier modelling phases. The flow analyses were performed for flow conditions similar to the present conditions. The modelling approach was based on the concept of an equivalent continuum. Each fracture zone and the rock matrix among the zones was, however, considered separately as a hydrogeologic unit. The numerical calculations were carried out with a computer code package, FEFLOW. The code is based upon the finite element method. With the code two- and one-dimensional elements can also be used by way of embedding them in a three-dimensional element mesh. A set of new algorithms was developed and employed to create element meshes for FEFLOW. The most useful program in the preliminary site investigations was PAAWI, which adds two-dimensional elements for fracture zones to an existing three-dimensional element mesh. The new algorithms reduced significantly the time required to create spatial discretization for complex geometries. Three element meshes were created for each site. The boundaries of the regional models coincide with those of the flow models. (55 refs., 40 figs., 1 tab.)
Protect Heterogeneous Environment Distributed Computing from Malicious Code Assignment
Directory of Open Access Journals (Sweden)
V. S. Gorbatov
2011-09-01
Full Text Available The paper describes the practical implementation of the protection system of heterogeneous environment distributed computing from malicious code for the assignment. A choice of technologies, development of data structures, performance evaluation of the implemented system security are conducted.
Method for quantitative assessment of nuclear safety computer codes
International Nuclear Information System (INIS)
Dearien, J.A.; Davis, C.B.; Matthews, L.J.
1979-01-01
A procedure has been developed for the quantitative assessment of nuclear safety computer codes and tested by comparison of RELAP4/MOD6 predictions with results from two Semiscale tests. This paper describes the developed procedure, the application of the procedure to the Semiscale tests, and the results obtained from the comparison
Atmospheric dispersion of radioactive releases: Computer code DIASPORA
International Nuclear Information System (INIS)
Synodinou, B.M.; Bartzis, J.M.
1982-05-01
The computer code DIASPORA is presented. Air and ground concentrations of an airborne radioactive material released from an elevated continuous point source are calculated using Gaussian plume models. Dry and wet deposition as well as plume rise effects are taken into consideration. (author)
Connecting Neural Coding to Number Cognition: A Computational Account
Prather, Richard W.
2012-01-01
The current study presents a series of computational simulations that demonstrate how the neural coding of numerical magnitude may influence number cognition and development. This includes behavioral phenomena cataloged in cognitive literature such as the development of numerical estimation and operational momentum. Though neural research has…
Methods and computer codes for probabilistic sensitivity and uncertainty analysis
International Nuclear Information System (INIS)
Vaurio, J.K.
1985-01-01
This paper describes the methods and applications experience with two computer codes that are now available from the National Energy Software Center at Argonne National Laboratory. The purpose of the SCREEN code is to identify a group of most important input variables of a code that has many (tens, hundreds) input variables with uncertainties, and do this without relying on judgment or exhaustive sensitivity studies. Purpose of the PROSA-2 code is to propagate uncertainties and calculate the distributions of interesting output variable(s) of a safety analysis code using response surface techniques, based on the same runs used for screening. Several applications are discussed, but the codes are generic, not tailored to any specific safety application code. They are compatible in terms of input/output requirements but also independent of each other, e.g., PROSA-2 can be used without first using SCREEN if a set of important input variables has first been selected by other methods. Also, although SCREEN can select cases to be run (by random sampling), a user can select cases by other methods if he so prefers, and still use the rest of SCREEN for identifying important input variables
A comparison of two three-dimensional shell-element transient electromagnetics codes
International Nuclear Information System (INIS)
Yugo, J.J.; Williamson, D.E.
1992-01-01
Electromagnetic forces due to eddy currents strongly influence the design of components for the next generation of fusion devices. An effort has been made to benchmark two computer programs used to generate transient electromagnetic loads: SPARK and EddyCuFF. Two simple transient field problems were analyzed, both of which had been previously analyzed by the SPARK code with results recorded in the literature. A third problem that uses an ITER inboard blanket benchmark model was analyzed as well. This problem was driven with a self-consistent, distributed multifilament plasma model generated by an axisymmetric physics code. The benchmark problems showed good agreement between the two shell-element codes. Variations in calculated eddy currents of 1--3% have been found for similar, finely meshed models. A difference of 8% was found in induced current and 20% in force for a coarse mesh and complex, multifilament field driver. Because comparisons were made to results obtained from literature, model preparation and code execution times were not evaluated
Power throttling of collections of computing elements
Bellofatto, Ralph E [Ridgefield, CT; Coteus, Paul W [Yorktown Heights, NY; Crumley, Paul G [Yorktown Heights, NY; Gara, Alan G [Mount Kidsco, NY; Giampapa, Mark E [Irvington, NY; Gooding,; Thomas, M [Rochester, MN; Haring, Rudolf A [Cortlandt Manor, NY; Megerian, Mark G [Rochester, MN; Ohmacht, Martin [Yorktown Heights, NY; Reed, Don D [Mantorville, MN; Swetz, Richard A [Mahopac, NY; Takken, Todd [Brewster, NY
2011-08-16
An apparatus and method for controlling power usage in a computer includes a plurality of computers communicating with a local control device, and a power source supplying power to the local control device and the computer. A plurality of sensors communicate with the computer for ascertaining power usage of the computer, and a system control device communicates with the computer for controlling power usage of the computer.
Computed radiography simulation using the Monte Carlo code MCNPX
International Nuclear Information System (INIS)
Correa, S.C.A.; Souza, E.M.; Silva, A.X.; Lopes, R.T.
2009-01-01
Simulating x-ray images has been of great interest in recent years as it makes possible an analysis of how x-ray images are affected owing to relevant operating parameters. In this paper, a procedure for simulating computed radiographic images using the Monte Carlo code MCNPX is proposed. The sensitivity curve of the BaFBr image plate detector as well as the characteristic noise of a 16-bit computed radiography system were considered during the methodology's development. The results obtained confirm that the proposed procedure for simulating computed radiographic images is satisfactory, as it allows obtaining results comparable with experimental data. (author)
Computed radiography simulation using the Monte Carlo code MCNPX
Energy Technology Data Exchange (ETDEWEB)
Correa, S.C.A. [Programa de Engenharia Nuclear/COPPE, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil); Centro Universitario Estadual da Zona Oeste (CCMAT)/UEZO, Av. Manuel Caldeira de Alvarenga, 1203, Campo Grande, 23070-200, Rio de Janeiro, RJ (Brazil); Souza, E.M. [Programa de Engenharia Nuclear/COPPE, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil); Silva, A.X., E-mail: ademir@con.ufrj.b [PEN/COPPE-DNC/Poli CT, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil); Cassiano, D.H. [Instituto de Radioprotecao e Dosimetria/CNEN Av. Salvador Allende, s/n, Recreio, 22780-160, Rio de Janeiro, RJ (Brazil); Lopes, R.T. [Programa de Engenharia Nuclear/COPPE, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil)
2010-09-15
Simulating X-ray images has been of great interest in recent years as it makes possible an analysis of how X-ray images are affected owing to relevant operating parameters. In this paper, a procedure for simulating computed radiographic images using the Monte Carlo code MCNPX is proposed. The sensitivity curve of the BaFBr image plate detector as well as the characteristic noise of a 16-bit computed radiography system were considered during the methodology's development. The results obtained confirm that the proposed procedure for simulating computed radiographic images is satisfactory, as it allows obtaining results comparable with experimental data.
Fuel rod computations. The COMETHE code in its CEA version
International Nuclear Information System (INIS)
Lenepveu, Dominique.
1976-01-01
The COMETHE code (COde d'evolution MEcanique et THermique) is intended for computing the irradiation behavior of water reactor fuel pins. It is concerned with steadily operated cylindrical pins, containing fuel pellet stacks (UO 2 or PuO 2 ). The pin consists in five different axial zones: two expansion chambers, two blankets, and a central core that may be divided into several stacks parted by plugs. As far as computation is concerned, the pin is divided into slices (maximum 15) in turn divided into rings (maximum 50). Information are obtained for each slice: the radial temperature distribution, heat transfer coefficients, thermal flux at the pin surface, changes in geometry according to temperature conditions, and specific burn-up. The physical models involved take account for: heat transfer, fission gas release, fuel expansion, and creep of the can. Results computed with COMETHE are compared with those from ELP and EPEL irradiation experiments [fr
Development and application of computational aerothermodynamics flowfield computer codes
Venkatapathy, Ethiraj
1993-01-01
Computations are presented for one-dimensional, strong shock waves that are typical of those that form in front of a reentering spacecraft. The fluid mechanics and thermochemistry are modeled using two different approaches. The first employs traditional continuum techniques in solving the Navier-Stokes equations. The second-approach employs a particle simulation technique (the direct simulation Monte Carlo method, DSMC). The thermochemical models employed in these two techniques are quite different. The present investigation presents an evaluation of thermochemical models for nitrogen under hypersonic flow conditions. Four separate cases are considered. The cases are governed, respectively, by the following: vibrational relaxation; weak dissociation; strong dissociation; and weak ionization. In near-continuum, hypersonic flow, the nonequilibrium thermochemical models employed in continuum and particle simulations produce nearly identical solutions. Further, the two approaches are evaluated successfully against available experimental data for weakly and strongly dissociating flows.
International Nuclear Information System (INIS)
Yamada, Tomonori
2010-01-01
The safety requirement of nuclear power plant attracts much attention nowadays. With the growing computing power, numerical simulation is one of key technologies to meet this safety requirement. Center for Computational Science and e-Systems of Japan Atomic Energy Agency has been developing a finite element analysis code for assembled structure to accurately evaluate the structural integrity of nuclear power plant in its entirety under seismic events. Because nuclear power plant is very huge assembled structure with tens of millions of mechanical components, the finite element model of each component is assembled into one structure and non-conforming meshes of mechanical components are bonded together inside the code. The main technique to bond these mechanical components is triple sparse matrix multiplication with multiple point constrains and global stiffness matrix. In our code, this procedure is conducted in a component by component manner, so that the working memory size and computing time for this multiplication are available on the current computing environment. As an illustrative example, seismic simulation of a real nuclear reactor of High Temperature engineering Test Reactor, which is located at the O-arai research and development center of JAEA, with 80 major mechanical components was conducted. Consequently, our code successfully simulated detailed elasto-plastic deformation of nuclear reactor and its computational performance was investigated. (author)
RADTRAN 5: A computer code for transportation risk analysis
International Nuclear Information System (INIS)
Neuhauser, K.S.; Kanipe, F.L.
1991-01-01
RADTRAN 5 is a computer code developed at Sandia National Laboratories (SNL) in Albuquerque, NM, to estimate radiological and nonradiological risks of radioactive materials transportation. RADTRAN 5 is written in ANSI Standard FORTRAN 77 and contains significant advances in the methodology for route-specific analysis first developed by SNL for RADTRAN 4 (Neuhauser and Kanipe, 1992). Like the previous RADTRAN codes, RADTRAN 5 contains two major modules for incident-free and accident risk amlysis, respectively. All commercially important transportation modes may be analyzed with RADTRAN 5: highway by combination truck; highway by light-duty vehicle; rail; barge; ocean-going ship; cargo air; and passenger air
An improved thermal model for the computer code NAIAD
International Nuclear Information System (INIS)
Rainbow, M.T.
1982-12-01
An improved thermal model, based on the concept of heat slabs, has been incorporated as an option into the thermal hydraulic computer code NAIAD. The heat slabs are one-dimensional thermal conduction models with temperature independent thermal properties which may be internal and/or external to the fluid. Thermal energy may be added to or removed from the fluid via heat slabs and passed across the external boundary of external heat slabs at a rate which is a linear function of the external surface temperatures. The code input for the new option has been restructured to simplify data preparation. A full description of current input requirements is presented
Integrated computer codes for nuclear power plant severe accident analysis
International Nuclear Information System (INIS)
Jordanov, I.; Khristov, Y.
1995-01-01
This overview contains a description of the Modular Accident Analysis Program (MAAP), ICARE computer code and Source Term Code Package (STCP). STCP is used to model TMLB sample problems for Zion Unit 1 and WWER-440/V-213 reactors. Comparison is made of STCP implementation on VAX and IBM systems. In order to improve accuracy, a double precision version of MARCH-3 component of STCP is created and the overall thermal hydraulics is modelled. Results of modelling the containment pressure, debris temperature, hydrogen mass are presented. 5 refs., 10 figs., 2 tabs
Aeschliman, D. P.; Oberkampf, W. L.; Blottner, F. G.
Verification, calibration, and validation (VCV) of Computational Fluid Dynamics (CFD) codes is an essential element of the code development process. The exact manner in which code VCV activities are planned and conducted, however, is critically important. It is suggested that the way in which code validation, in particular, is often conducted--by comparison to published experimental data obtained for other purposes--is in general difficult and unsatisfactory, and that a different approach is required. This paper describes a proposed methodology for CFD code VCV that meets the technical requirements and is philosophically consistent with code development needs. The proposed methodology stresses teamwork and cooperation between code developers and experimentalists throughout the VCV process, and takes advantage of certain synergisms between CFD and experiment. A novel approach to uncertainty analysis is described which can both distinguish between and quantify various types of experimental error, and whose attributes are used to help define an appropriate experimental design for code VCV experiments. The methodology is demonstrated with an example of laminar, hypersonic, near perfect gas, 3-dimensional flow over a sliced sphere/cone of varying geometrical complexity.
Theoretical Atomic Physics code development IV: LINES, A code for computing atomic line spectra
International Nuclear Information System (INIS)
Abdallah, J. Jr.; Clark, R.E.H.
1988-12-01
A new computer program, LINES, has been developed for simulating atomic line emission and absorption spectra using the accurate fine structure energy levels and transition strengths calculated by the (CATS) Cowan Atomic Structure code. Population distributions for the ion stages are obtained in LINES by using the Local Thermodynamic Equilibrium (LTE) model. LINES is also useful for displaying the pertinent atomic data generated by CATS. This report describes the use of LINES. Both CATS and LINES are part of the Theoretical Atomic PhysicS (TAPS) code development effort at Los Alamos. 11 refs., 9 figs., 1 tab
Computer codes for evaluation of control room habitability (HABIT)
International Nuclear Information System (INIS)
Stage, S.A.
1996-06-01
This report describes the Computer Codes for Evaluation of Control Room Habitability (HABIT). HABIT is a package of computer codes designed to be used for the evaluation of control room habitability in the event of an accidental release of toxic chemicals or radioactive materials. Given information about the design of a nuclear power plant, a scenario for the release of toxic chemicals or radionuclides, and information about the air flows and protection systems of the control room, HABIT can be used to estimate the chemical exposure or radiological dose to control room personnel. HABIT is an integrated package of several programs that previously needed to be run separately and required considerable user intervention. This report discusses the theoretical basis and physical assumptions made by each of the modules in HABIT and gives detailed information about the data entry windows. Sample runs are given for each of the modules. A brief section of programming notes is included. A set of computer disks will accompany this report if the report is ordered from the Energy Science and Technology Software Center. The disks contain the files needed to run HABIT on a personal computer running DOS. Source codes for the various HABIT routines are on the disks. Also included are input and output files for three demonstration runs
Convergence acceleration of the Proteus computer code with multigrid methods
Demuren, A. O.; Ibraheem, S. O.
1995-01-01
This report presents the results of a study to implement convergence acceleration techniques based on the multigrid concept in the two-dimensional and three-dimensional versions of the Proteus computer code. The first section presents a review of the relevant literature on the implementation of the multigrid methods in computer codes for compressible flow analysis. The next two sections present detailed stability analysis of numerical schemes for solving the Euler and Navier-Stokes equations, based on conventional von Neumann analysis and the bi-grid analysis, respectively. The next section presents details of the computational method used in the Proteus computer code. Finally, the multigrid implementation and applications to several two-dimensional and three-dimensional test problems are presented. The results of the present study show that the multigrid method always leads to a reduction in the number of iterations (or time steps) required for convergence. However, there is an overhead associated with the use of multigrid acceleration. The overhead is higher in 2-D problems than in 3-D problems, thus overall multigrid savings in CPU time are in general better in the latter. Savings of about 40-50 percent are typical in 3-D problems, but they are about 20-30 percent in large 2-D problems. The present multigrid method is applicable to steady-state problems and is therefore ineffective in problems with inherently unstable solutions.
War of ontology worlds: mathematics, computer code, or Esperanto?
Rzhetsky, Andrey; Evans, James A
2011-09-01
The use of structured knowledge representations-ontologies and terminologies-has become standard in biomedicine. Definitions of ontologies vary widely, as do the values and philosophies that underlie them. In seeking to make these views explicit, we conducted and summarized interviews with a dozen leading ontologists. Their views clustered into three broad perspectives that we summarize as mathematics, computer code, and Esperanto. Ontology as mathematics puts the ultimate premium on rigor and logic, symmetry and consistency of representation across scientific subfields, and the inclusion of only established, non-contradictory knowledge. Ontology as computer code focuses on utility and cultivates diversity, fitting ontologies to their purpose. Like computer languages C++, Prolog, and HTML, the code perspective holds that diverse applications warrant custom designed ontologies. Ontology as Esperanto focuses on facilitating cross-disciplinary communication, knowledge cross-referencing, and computation across datasets from diverse communities. We show how these views align with classical divides in science and suggest how a synthesis of their concerns could strengthen the next generation of biomedical ontologies.
War of Ontology Worlds: Mathematics, Computer Code, or Esperanto?
Rzhetsky, Andrey; Evans, James A.
2011-01-01
The use of structured knowledge representations—ontologies and terminologies—has become standard in biomedicine. Definitions of ontologies vary widely, as do the values and philosophies that underlie them. In seeking to make these views explicit, we conducted and summarized interviews with a dozen leading ontologists. Their views clustered into three broad perspectives that we summarize as mathematics, computer code, and Esperanto. Ontology as mathematics puts the ultimate premium on rigor and logic, symmetry and consistency of representation across scientific subfields, and the inclusion of only established, non-contradictory knowledge. Ontology as computer code focuses on utility and cultivates diversity, fitting ontologies to their purpose. Like computer languages C++, Prolog, and HTML, the code perspective holds that diverse applications warrant custom designed ontologies. Ontology as Esperanto focuses on facilitating cross-disciplinary communication, knowledge cross-referencing, and computation across datasets from diverse communities. We show how these views align with classical divides in science and suggest how a synthesis of their concerns could strengthen the next generation of biomedical ontologies. PMID:21980276
A computer program for structural analysis of fuel elements
International Nuclear Information System (INIS)
Hayashi, I.M.V.; Perrotta, J.A.
1988-01-01
It's presented the code ELCOM for the matrix analysis of tubular structures coupled by rigid spacers, typical of PWR's fuel elements. The code ELCOM makes a static structural analysis, where the displacements and internal forces are obtained for each structure at the joints with the spacers, and also, the natural frequencies and vibrational modes of an equivalent integrated structure are obtained. The ELCOM result is compared to a PWR fuel element structural analysis obtained in published paper. (author) [pt
COYOTE: a finite element computer program for nonlinear heat conduction problems
International Nuclear Information System (INIS)
Gartling, D.K.
1978-06-01
COYOTE is a finite element computer program designed for the solution of two-dimensional, nonlinear heat conduction problems. The theoretical and mathematical basis used to develop the code is described. Program capabilities and complete user instructions are presented. Several example problems are described in detail to demonstrate the use of the program
Benchmarking of computer codes and approaches for modeling exposure scenarios
International Nuclear Information System (INIS)
Seitz, R.R.; Rittmann, P.D.; Wood, M.I.; Cook, J.R.
1994-08-01
The US Department of Energy Headquarters established a performance assessment task team (PATT) to integrate the activities of DOE sites that are preparing performance assessments for the disposal of newly generated low-level waste. The PATT chartered a subteam with the task of comparing computer codes and exposure scenarios used for dose calculations in performance assessments. This report documents the efforts of the subteam. Computer codes considered in the comparison include GENII, PATHRAE-EPA, MICROSHIELD, and ISOSHLD. Calculations were also conducted using spreadsheets to provide a comparison at the most fundamental level. Calculations and modeling approaches are compared for unit radionuclide concentrations in water and soil for the ingestion, inhalation, and external dose pathways. Over 30 tables comparing inputs and results are provided
FRANTIC: a computer code for time dependent unavailability analysis
International Nuclear Information System (INIS)
Vesely, W.E.; Goldberg, F.F.
1977-03-01
The FRANTIC computer code evaluates the time dependent and average unavailability for any general system model. The code is written in FORTRAN IV for the IBM 370 computer. Non-repairable components, monitored components, and periodically tested components are handled. One unique feature of FRANTIC is the detailed, time dependent modeling of periodic testing which includes the effects of test downtimes, test overrides, detection inefficiencies, and test-caused failures. The exponential distribution is used for the component failure times and periodic equations are developed for the testing and repair contributions. Human errors and common mode failures can be included by assigning an appropriate constant probability for the contributors. The output from FRANTIC consists of tables and plots of the system unavailability along with a breakdown of the unavailability contributions. Sensitivity studies can be simply performed and a wide range of tables and plots can be obtained for reporting purposes. The FRANTIC code represents a first step in the development of an approach that can be of direct value in future system evaluations. Modifications resulting from use of the code, along with the development of reliability data based on operating reactor experience, can be expected to provide increased confidence in its use and potential application to the licensing process
Computational radiology and imaging with the MCNP Monte Carlo code
Energy Technology Data Exchange (ETDEWEB)
Estes, G.P.; Taylor, W.M.
1995-05-01
MCNP, a 3D coupled neutron/photon/electron Monte Carlo radiation transport code, is currently used in medical applications such as cancer radiation treatment planning, interpretation of diagnostic radiation images, and treatment beam optimization. This paper will discuss MCNP`s current uses and capabilities, as well as envisioned improvements that would further enhance MCNP role in computational medicine. It will be demonstrated that the methodology exists to simulate medical images (e.g. SPECT). Techniques will be discussed that would enable the construction of 3D computational geometry models of individual patients for use in patient-specific studies that would improve the quality of care for patients.
Computer code for qualitative analysis of gamma-ray spectra
International Nuclear Information System (INIS)
Yule, H.P.
1979-01-01
Computer code QLN1 provides complete analysis of gamma-ray spectra observed with Ge(Li) detectors and is used at both the National Bureau of Standards and the Environmental Protection Agency. It locates peaks, resolves multiplets, identifies component radioisotopes, and computes quantitative results. The qualitative-analysis (or component identification) algorithms feature thorough, self-correcting steps which provide accurate isotope identification in spite of errors in peak centroids, energy calibration, and other typical problems. The qualitative-analysis algorithm is described in this paper
Bragg optics computer codes for neutron scattering instrument design
International Nuclear Information System (INIS)
Popovici, M.; Yelon, W.B.; Berliner, R.R.; Stoica, A.D.
1997-01-01
Computer codes for neutron crystal spectrometer design, optimization and experiment planning are described. Phase space distributions, linewidths and absolute intensities are calculated by matrix methods in an extension of the Cooper-Nathans resolution function formalism. For modeling the Bragg reflection on bent crystals the lamellar approximation is used. Optimization is done by satisfying conditions of focusing in scattering and in real space, and by numerically maximizing figures of merit. Examples for three-axis and two-axis spectrometers are given
Methods for the development of large computer codes under LTSS
International Nuclear Information System (INIS)
Sicilian, J.M.
1977-06-01
TRAC is a large computer code being developed by Group Q-6 for the analysis of the transient thermal hydraulic behavior of light-water nuclear reactors. A system designed to assist the development of TRAC is described. The system consists of a central HYDRA dataset, R6LIB, containing files used in the development of TRAC, and a file maintenance program, HORSE, which facilitates the use of this dataset
WAMCUT, a computer code for fault tree evaluation. Final report
International Nuclear Information System (INIS)
Erdmann, R.C.
1978-06-01
WAMCUT is a code in the WAM family which produces the minimum cut sets (MCS) for a given fault tree. The MCS are useful as they provide a qualitative evaluation of a system, as well as providing a means of determining the probability distribution function for the top of the tree. The program is very efficient and will produce all the MCS in a very short computer time span. 22 figures, 4 tables
International Nuclear Information System (INIS)
Bujan, A.; Adamik, V.; Misak, J.
1986-01-01
A brief description is presented of the expansion of the SICHTA-83 computer code for the analysis of the thermal history of the fuel channel for large LOCAs by modelling the mechanical behaviour of fuel element cladding. The new version of the code has a more detailed treatment of heat transfer in the fuel-cladding gap because it also respects the mechanical (plastic) deformations of the cladding and the fuel-cladding interaction (magnitude of contact pressure). Also respected is the change in pressure of the gas filling of the fuel element, the mechanical criterion is considered of a failure of the cladding and the degree is considered of the blockage of the through-flow cross section for coolant flow in the fuel channel. The LOCA WWER-440 model computation provides a comparison of the new SICHTA-85/MOD 1 code with the results of the original 83 version of SICHTA. (author)
INGEN: a general-purpose mesh generator for finite element codes
International Nuclear Information System (INIS)
Cook, W.A.
1979-05-01
INGEN is a general-purpose mesh generator for two- and three-dimensional finite element codes. The basic parts of the code are surface and three-dimensional region generators that use linear-blending interpolation formulas. These generators are based on an i, j, k index scheme that is used to number nodal points, construct elements, and develop displacement and traction boundary conditions. This code can generate truss elements (2 modal points); plane stress, plane strain, and axisymmetry two-dimensional continuum elements (4 to 8 nodal points); plate elements (4 to 8 nodal points); and three-dimensional continuum elements (8 to 21 nodal points). The traction loads generated are consistent with the element generated. The expansion--contraction option is of special interest. This option makes it possible to change an existing mesh such that some regions are refined and others are made coarser than the original mesh. 9 figures
SP_Ace: a new code to derive stellar parameters and elemental abundances
Boeche, C.; Grebel, E. K.
2016-03-01
Context. Ongoing and future massive spectroscopic surveys will collect large numbers (106-107) of stellar spectra that need to be analyzed. Highly automated software is needed to derive stellar parameters and chemical abundances from these spectra. Aims: We developed a new method of estimating the stellar parameters Teff, log g, [M/H], and elemental abundances. This method was implemented in a new code, SP_Ace (Stellar Parameters And Chemical abundances Estimator). This is a highly automated code suitable for analyzing the spectra of large spectroscopic surveys with low or medium spectral resolution (R = 2000-20 000). Methods: After the astrophysical calibration of the oscillator strengths of 4643 absorption lines covering the wavelength ranges 5212-6860 Å and 8400-8924 Å, we constructed a library that contains the equivalent widths (EW) of these lines for a grid of stellar parameters. The EWs of each line are fit by a polynomial function that describes the EW of the line as a function of the stellar parameters. The coefficients of these polynomial functions are stored in a library called the "GCOG library". SP_Ace, a code written in FORTRAN95, uses the GCOG library to compute the EWs of the lines, constructs models of spectra as a function of the stellar parameters and abundances, and searches for the model that minimizes the χ2 deviation when compared to the observed spectrum. The code has been tested on synthetic and real spectra for a wide range of signal-to-noise and spectral resolutions. Results: SP_Ace derives stellar parameters such as Teff, log g, [M/H], and chemical abundances of up to ten elements for low to medium resolution spectra of FGK-type stars with precision comparable to the one usually obtained with spectra of higher resolution. Systematic errors in stellar parameters and chemical abundances are presented and identified with tests on synthetic and real spectra. Stochastic errors are automatically estimated by the code for all the parameters
International Nuclear Information System (INIS)
Kimura, Hideo; Takahashi, Tomoyuki; Shima, Shigeki; Matsuzuru, Hideo
1992-11-01
The computer code system GSRW (Generic Safety assessment code for geologic disposal of Radioactive Waste) was developed as in interim version of safety assessment methodology for geologic disposal of high-level radioactive waste. Scenarios used here are based on normal evolution scenarios which assume that the performance of a disposal system is not affected by probabilistic events. The code consists of three parts. The first part evaluates a source term from a disposal facility which consists mainly of a vitrified waste, a metallic container and a buffer zone. Two kinds of source term models are provided: Model 1 which simulate the dissolution of silicate component of glass and the diffusive transport of radionuclides in the buffere zone, and Model 2 which assumes that the concentration of a radionuclide is limited by the solubility of its specific chemical form at the interface between the buffer and a vitrified wastes. The second part analyses the transport of radionuclides in the geosphere, which is based on analytical solutions or numerical solutions of a mass transport equation involving the advection, dispersion, linear sorption and decay chain. The third part assesses the transport of radionuclides in the biosphere and the resulting radiological consequences to the man, which is based on a dynamic compartment model for the biosphere and a dose factor method for dose calculations. This report describes mathematical models used, the structure of the code system, and user information and instructions for execution of the code. (author)
Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis
International Nuclear Information System (INIS)
Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E.; Tills, J.
1997-12-01
The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions
Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis
Energy Technology Data Exchange (ETDEWEB)
Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E. [Sandia National Labs., Albuquerque, NM (United States); Tills, J. [J. Tills and Associates, Inc., Sandia Park, NM (United States)
1997-12-01
The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.
GATO Code Modification to Compute Plasma Response to External Perturbations
Turnbull, A. D.; Chu, M. S.; Ng, E.; Li, X. S.; James, A.
2006-10-01
It has become increasingly clear that the plasma response to an external nonaxiymmetric magnetic perturbation cannot be neglected in many situations of interest. This response can be described as a linear combination of the eigenmodes of the ideal MHD operator. The eigenmodes of the system can be obtained numerically with the GATO ideal MHD stability code, which has been modified for this purpose. A key requirement is the removal of inadmissible continuum modes. For Finite Hybrid Element codes such as GATO, a prerequisite for this is their numerical restabilization by addition of small numerical terms to δ,to cancel the analytic numerical destabilization. In addition, robustness of the code was improved and the solution method speeded up by use of the SuperLU package to facilitate calculation of the full set of eigenmodes in a reasonable time. To treat resonant plasma responses, the finite element basis has been extended to include eigenfunctions with finite jumps at rational surfaces. Some preliminary numerical results for DIII-D equilibria will be given.
A directory of computer codes suitable for stress analysis of HLW containers - Compas project
International Nuclear Information System (INIS)
1989-01-01
This document reports the work carried out for the Compas project which looked at the capabilities of various computer codes for the stress analysis of high-level nuclear-waste containers and overpacks. The report concentrates on codes used by the project partners, but also includes a number of the major commercial finite element codes. The report falls into two parts. The first part of the report describes the capabilities of the codes. This includes details of the solution methods used in the codes, the types of analysis which they can carry out and the interfacing with pre - and post - processing packages. This is the more comprehensive section of the report. The second part of the report looks at the performance of a selection of the codes (those used by the project partners). This look at how the codes perform in a number of test problems which require calculations typical of those encountered in the design and analysis of high-level waste containers and overpacks
Interface between computational fluid dynamics (CFD) and plant analysis computer codes
International Nuclear Information System (INIS)
Coffield, R.D.; Dunckhorst, F.F.; Tomlinson, E.T.; Welch, J.W.
1993-01-01
Computational fluid dynamics (CFD) can provide valuable input to the development of advanced plant analysis computer codes. The types of interfacing discussed in this paper will directly contribute to modeling and accuracy improvements throughout the plant system and should result in significant reduction of design conservatisms that have been applied to such analyses in the past
MAPVAR - A Computer Program to Transfer Solution Data Between Finite Element Meshes
Energy Technology Data Exchange (ETDEWEB)
Wellman, G.W.
1999-03-01
MAPVAR, as was the case with its precursor programs, MERLIN and MERLIN II, is designed to transfer solution results from one finite element mesh to another. MAPVAR draws heavily from the structure and coding of MERLIN II, but it employs a new finite element data base, EXODUS II, and offers enhanced speed and new capabilities not available in MERLIN II. In keeping with the MERLIN II documentation, the computational algorithms used in MAPVAR are described. User instructions are presented. Example problems are included to demonstrate the operation of the code and the effects of various input options.
Improved Flow Modeling in Transient Reactor Safety Analysis Computer Codes
International Nuclear Information System (INIS)
Holowach, M.J.; Hochreiter, L.E.; Cheung, F.B.
2002-01-01
A method of accounting for fluid-to-fluid shear in between calculational cells over a wide range of flow conditions envisioned in reactor safety studies has been developed such that it may be easily implemented into a computer code such as COBRA-TF for more detailed subchannel analysis. At a given nodal height in the calculational model, equivalent hydraulic diameters are determined for each specific calculational cell using either laminar or turbulent velocity profiles. The velocity profile may be determined from a separate CFD (Computational Fluid Dynamics) analysis, experimental data, or existing semi-empirical relationships. The equivalent hydraulic diameter is then applied to the wall drag force calculation so as to determine the appropriate equivalent fluid-to-fluid shear caused by the wall for each cell based on the input velocity profile. This means of assigning the shear to a specific cell is independent of the actual wetted perimeter and flow area for the calculational cell. The use of this equivalent hydraulic diameter for each cell within a calculational subchannel results in a representative velocity profile which can further increase the accuracy and detail of heat transfer and fluid flow modeling within the subchannel when utilizing a thermal hydraulics systems analysis computer code such as COBRA-TF. Utilizing COBRA-TF with the flow modeling enhancement results in increased accuracy for a coarse-mesh model without the significantly greater computational and time requirements of a full-scale 3D (three-dimensional) transient CFD calculation. (authors)
Johnson, Ryan; Kercher, Andrew; Schwer, Douglas; Corrigan, Andrew; Kailasanath, Kazhikathra
2017-11-01
This presentation focuses on the development of a Discontinuous Galerkin (DG) method for application to chemically reacting flows. The in-house code, called Propel, was developed by the Laboratory of Computational Physics and Fluid Dynamics at the Naval Research Laboratory. It was designed specifically for developing advanced multi-dimensional algorithms to run efficiently on new and innovative architectures such as GPUs. For these results, Propel solves for convection and diffusion simultaneously with detailed transport and thermodynamics. Chemistry is currently solved in a time-split approach using Strang-splitting with finite element DG time integration of chemical source terms. Results presented here show canonical unsteady reacting flow cases, such as co-flow and splitter plate, and we report performance for higher order DG on CPU and GPUs.
International Nuclear Information System (INIS)
Viswanathan, H.S.
1995-01-01
The finite element code FEHMN is a three-dimensional finite element heat and mass transport simulator that can handle complex stratigraphy and nonlinear processes such as vadose zone flow, heat flow and solute transport. Scientists at LANL have been developed hydrologic flow and transport models of the Yucca Mountain site using FEHMN. Previous FEHMN simulations have used an equivalent K d model to model solute transport. In this thesis, FEHMN is modified making it possible to simulate the transport of a species with a rigorous chemical model. Including the rigorous chemical equations into FEHMN simulations should provide for more representative transport models for highly reactive chemical species. A fully kinetic formulation is chosen for the FEHMN reactive transport model. Several methods are available to computationally implement a fully kinetic formulation. Different numerical algorithms are investigated in order to optimize computational efficiency and memory requirements of the reactive transport model. The best algorithm of those investigated is then incorporated into FEHMN. The algorithm chosen requires for the user to place strongly coupled species into groups which are then solved for simultaneously using FEHMN. The complete reactive transport model is verified over a wide variety of problems and is shown to be working properly. The simulations demonstrate that gas flow and carbonate chemistry can significantly affect 14 C transport at Yucca Mountain. The simulations also provide that the new capabilities of FEHMN can be used to refine and buttress already existing Yucca Mountain radionuclide transport studies
The computer code SEURBNUK/EURDYN. Pt. 2
International Nuclear Information System (INIS)
Yerkess, A.; Broadhouse, B.J.; Smith, B.L.
1987-01-01
SEURBNUK-2 is a two-dimensional, axisymmetric, Eulerian, finite difference containment code. The numerical procedure adopted in SEURBNUK to solve the hydrodynamic equations is based on the semi-implicit ICE method which itself is an extension of the MAC algorithm. SEURBNUK has a finite difference thin shell treatment for vessels and internal structures of arbitrary shape and includes the effects of the compressibility of the fluid. Fluid flow through porous media and porous structures can also be accommodated. SEURBNUK/EURDYN is an extension of SEURBNUK-2 in which the finite difference thin shell treatment is replaced by a finite element calculation for both thin or thick structures. This has been achieved by coupling the shell elements and axisymmetric triangular elements. Within the code, the equations of motion for the structures are solved quite separately from those for the fluid, and the timestep for the fluid can be an integer multiple of that for the structures. The interaction of the structures with the fluid is then considered as a modification to the coefficients in the pressure equations, the modifications naturally depending on the behaviour of the structures within the fluid cell. The code is limited to dealing with a single fluid, the coolant, and the bubble and the cover gas are treated as cavities of uniform pressure calculated via appropriate pressure-volume-energy relationships. This manual describes the input data specifications needed for the execution of SEURBNUK/EURDYN calculations. After explaining the output facilities information is included to aid users to avoid some common pit-falls
The use of the MCNP code for the quantitative analysis of elements in geological formations
Energy Technology Data Exchange (ETDEWEB)
Cywicka-Jakiel, T.; Woynicka, U. [The Henryk Niewodniczanski Institute of Nuclear Physics, Krakow (Poland); Zorski, T. [University of Mining and Metallurgy, Faculty of Geology, Geophysics and Environmental Protection, Krakow (Poland)
2003-07-01
The Monte Carlo modelling calculations using the MCNP code have been performed, which support the spectrometric neutron-gamma (SNGL) borehole logging. The SNGL enables the lithology identification through the quantitative analysis of the elements in geological formations and thus can be very useful for the oil and gas industry as well as for prospecting of the potential host rocks for radioactive waste disposal. In the SNGL experiment, gamma-rays induced by the neutron interactions with the nuclei of the rock elements are detected using the gamma-ray probe of complex mechanical and electronic construction. The probe has to be calibrated for a wide range of the elemental concentrations, to assure the proper quantitative analysis. The Polish Calibration Station in Zielona Gora is equipped with a limited number of calibration standards. An extension of the experimental calibration and the evaluation of the effect of the so-called side effects (for example the borehole and formation salinity variation) on the accuracy of the SNGL method can be done by the use of the MCNP code. The preliminary MCNP results showing the effect of the borehole and formation fluids salinity variations on the accuracy of silicon (Si), calcium (Ca) and iron (Fe) content determination are presented in the paper. The main effort has been focused on a modelling of the complex SNGL probe situated in a fluid filled borehole, surrounded by a geological formation. Track length estimate of the photon flux from the (n,gamma) interactions as a function of gamma-rays energy was used. Calculations were run on the PC computer with AMD Athlon 1.33 GHz processor. Neutron and photon cross-sections libraries were taken from the MCNP4c package and based mainly on the ENDF/B-6, ENDF/B-5 and MCPLIB02 data. The results of simulated experiment are in conformity with results of the real experiment performed with the use of the main lithology models (sandstones, limestones and dolomite). (authors)
The use of the MCNP code for the quantitative analysis of elements in geological formations
International Nuclear Information System (INIS)
Cywicka-Jakiel, T.; Woynicka, U.; Zorski, T.
2003-01-01
The Monte Carlo modelling calculations using the MCNP code have been performed, which support the spectrometric neutron-gamma (SNGL) borehole logging. The SNGL enables the lithology identification through the quantitative analysis of the elements in geological formations and thus can be very useful for the oil and gas industry as well as for prospecting of the potential host rocks for radioactive waste disposal. In the SNGL experiment, gamma-rays induced by the neutron interactions with the nuclei of the rock elements are detected using the gamma-ray probe of complex mechanical and electronic construction. The probe has to be calibrated for a wide range of the elemental concentrations, to assure the proper quantitative analysis. The Polish Calibration Station in Zielona Gora is equipped with a limited number of calibration standards. An extension of the experimental calibration and the evaluation of the effect of the so-called side effects (for example the borehole and formation salinity variation) on the accuracy of the SNGL method can be done by the use of the MCNP code. The preliminary MCNP results showing the effect of the borehole and formation fluids salinity variations on the accuracy of silicon (Si), calcium (Ca) and iron (Fe) content determination are presented in the paper. The main effort has been focused on a modelling of the complex SNGL probe situated in a fluid filled borehole, surrounded by a geological formation. Track length estimate of the photon flux from the (n,gamma) interactions as a function of gamma-rays energy was used. Calculations were run on the PC computer with AMD Athlon 1.33 GHz processor. Neutron and photon cross-sections libraries were taken from the MCNP4c package and based mainly on the ENDF/B-6, ENDF/B-5 and MCPLIB02 data. The results of simulated experiment are in conformity with results of the real experiment performed with the use of the main lithology models (sandstones, limestones and dolomite). (authors)
International Nuclear Information System (INIS)
Zhang, Shuai; Morita, Koji; Shirakawa, Noriyuki; Yamamoto, Yuichi
2010-01-01
The COMPASS code is designed based on the moving particle semi-implicit method to simulate various complex mesoscale phenomena relevant to core disruptive accidents of sodium-cooled fast reactors. In this study, a computational framework for fluid-solid mixture flow simulations was developed for the COMPASS code. The passively moving solid model was used to simulate hydrodynamic interactions between fluid and solids. Mechanical interactions between solids were modeled by the distinct element method. A multi-time-step algorithm was introduced to couple these two calculations. The proposed computational framework for fluid-solid mixture flow simulations was verified by the comparison between experimental and numerical studies on the water-dam break with multiple solid rods. (author)
RADTRAN 5 - A computer code for transportation risk analysis
International Nuclear Information System (INIS)
Neuhauser, K.S.; Kanipe, F.L.
1993-01-01
The RADTRAN 5 computer code has been developed to estimate radiological and nonradiological risks of radioactive materials transportation. RADTRAN 5 is written in ANSI standard FORTRAN 77; the code contains significant advances in the methodology first pioneered with the LINK option of RADTRAN 4. A major application of the LINK methodology is route-specific analysis. Another application is comparisons of attributes along the same route segments. Nonradiological risk factors have been incorporated to allow users to estimate nonradiological fatalities and injuries that might occur during the transportation event(s) being analyzed. These fatalities include prompt accidental fatalities from mechanical causes. Values of these risk factors for the United States have been made available in the code as optional defaults. Several new health effects models have been published in the wake of the Hiroshima-Nagasaki dosimetry reassessment, and this has emphasized the need for flexibility in the RADTRAN approach to health-effects calculations. Therefore, the basic set of health-effects conversion equations in RADTRAN have been made user-definable. All parameter values can be changed by the user, but a complete set of default values are available for both the new International Commission on Radiation Protection model (ICRP Publication 60) and the recent model of the U.S. National Research Council's Committee on the Biological Effects of Radiation (BEIR V). The meteorological input data tables have been modified to permit optional entry of maximum downwind distances for each dose isopleth. The expected dose to an individual in each isodose area is also calculated and printed automatically. Examples are given that illustrate the power and flexibility of the RADTRAN 5 computer code. (J.P.N.)
CARP: a computer code and albedo data library for use by BREESE, the MORSE albedo package
International Nuclear Information System (INIS)
Emmett, M.B.; Rhoades, W.A.
1978-10-01
The CARP computer code was written to allow processing of DOT angular flux tapes to produce albedo data for use in the MORSE computer code. An albedo data library was produced containing several materials. 3 tables
Compilation of the abstracts of nuclear computer codes available at CPD/IPEN
International Nuclear Information System (INIS)
Granzotto, A.; Gouveia, A.S. de; Lourencao, E.M.
1981-06-01
A compilation of all computer codes available at IPEN in S.Paulo are presented. These computer codes are classified according to Argonne National Laboratory - and Energy Nuclear Agency schedule. (E.G.) [pt
International Nuclear Information System (INIS)
Pot, G.; Laurence, D.; Rharif, N.E.; Leal de Sousa, L.; Compe, C.
1995-12-01
This paper deals with the introduction of a second moment closure turbulence model (Reynolds Stress Model) in an industrial finite element code, N3S, developed at Electricite de France.The numerical implementation of the model in N3S will be detailed in 2D and 3D. Some details are given concerning finite element computations and solvers. Then, some results will be given, including a comparison between standard k-ε model, R.S.M. model and experimental data for some test case. (authors). 22 refs., 3 figs
01010000 01001100 01000001 01011001: Play Elements in Computer Programming
Breslin, Samantha
2013-01-01
This article explores the role of play in human interaction with computers in the context of computer programming. The author considers many facets of programming including the literary practice of coding, the abstract design of programs, and more mundane activities such as testing, debugging, and hacking. She discusses how these incorporate the…
Multicode comparison of selected source-term computer codes
Energy Technology Data Exchange (ETDEWEB)
Hermann, O.W.; Parks, C.V.; Renier, J.P.; Roddy, J.W.; Ashline, R.C.; Wilson, W.B.; LaBauve, R.J.
1989-04-01
This report summarizes the results of a study to assess the predictive capabilities of three radionuclide inventory/depletion computer codes, ORIGEN2, ORIGEN-S, and CINDER-2. The task was accomplished through a series of comparisons of their output for several light-water reactor (LWR) models (i.e., verification). Of the five cases chosen, two modeled typical boiling-water reactors (BWR) at burnups of 27.5 and 40 GWd/MTU and two represented typical pressurized-water reactors (PWR) at burnups of 33 and 50 GWd/MTU. In the fifth case, identical input data were used for each of the codes to examine the results of decay only and to show differences in nuclear decay constants and decay heat rates. Comparisons were made for several different characteristics (mass, radioactivity, and decay heat rate) for 52 radionuclides and for nine decay periods ranging from 30 d to 10,000 years. Only fission products and actinides were considered. The results are presented in comparative-ratio tables for each of the characteristics, decay periods, and cases. A brief summary description of each of the codes has been included. Of the more than 21,000 individual comparisons made for the three codes (taken two at a time), nearly half (45%) agreed to within 1%, and an additional 17% fell within the range of 1 to 5%. Approximately 8% of the comparison results disagreed by more than 30%. However, relatively good agreement was obtained for most of the radionuclides that are expected to contribute the greatest impact to waste disposal. Even though some defects have been noted, each of the codes in the comparison appears to produce respectable results. 12 figs., 12 tabs.
Code Verification of the HIGRAD Computational Fluid Dynamics Solver
Energy Technology Data Exchange (ETDEWEB)
Van Buren, Kendra L. [Los Alamos National Laboratory; Canfield, Jesse M. [Los Alamos National Laboratory; Hemez, Francois M. [Los Alamos National Laboratory; Sauer, Jeremy A. [Los Alamos National Laboratory
2012-05-04
The purpose of this report is to outline code and solution verification activities applied to HIGRAD, a Computational Fluid Dynamics (CFD) solver of the compressible Navier-Stokes equations developed at the Los Alamos National Laboratory, and used to simulate various phenomena such as the propagation of wildfires and atmospheric hydrodynamics. Code verification efforts, as described in this report, are an important first step to establish the credibility of numerical simulations. They provide evidence that the mathematical formulation is properly implemented without significant mistakes that would adversely impact the application of interest. Highly accurate analytical solutions are derived for four code verification test problems that exercise different aspects of the code. These test problems are referred to as: (i) the quiet start, (ii) the passive advection, (iii) the passive diffusion, and (iv) the piston-like problem. These problems are simulated using HIGRAD with different levels of mesh discretization and the numerical solutions are compared to their analytical counterparts. In addition, the rates of convergence are estimated to verify the numerical performance of the solver. The first three test problems produce numerical approximations as expected. The fourth test problem (piston-like) indicates the extent to which the code is able to simulate a 'mild' discontinuity, which is a condition that would typically be better handled by a Lagrangian formulation. The current investigation concludes that the numerical implementation of the solver performs as expected. The quality of solutions is sufficient to provide credible simulations of fluid flows around wind turbines. The main caveat associated to these findings is the low coverage provided by these four problems, and somewhat limited verification activities. A more comprehensive evaluation of HIGRAD may be beneficial for future studies.
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.)
An Interactive Preprocessor Program with Graphics for a Three-Dimensional Finite Element Code.
Hamilton, Claude Hayden, III
The development and capabilities of an interactive preprocessor program with graphics for an existing three-dimensional finite element code is presented. This preprocessor program, EDGAP3D, is designed to be used in conjunction with the Texas Three Dimensional Grain Analysis Program (TXCAP3D). The code presented in this research is capable of the…
Computer codes in nuclear safety, radiation transport and dosimetry
International Nuclear Information System (INIS)
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.
2006-01-01
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
Utilization of Relap 5 computer code for analyzing thermohydraulic projects
International Nuclear Information System (INIS)
Silva Filho, E.
1987-01-01
This work deals with the design of a scaled test facility of a typical pressurized water reactor plant of the 1300 MW (electric) class. A station blackout has been choosen to investigate the thermohydraulic behaviour of the the test facility in comparison to the reactor plant. The computer code RELAPS/MOD1 has been utilized to simulate the blackout and to compare the test facility behaviour with the reactor plant one. The results demonstrate similar thermohydraulic behaviours of the two systems. (author) [pt
Knowlton, Marie; Wetzel, Robin
2006-01-01
This study compared the length of text in English Braille American Edition, the Nemeth code, and the computer braille code with the Unified English Braille Code (UEBC)--also known as Unified English Braille (UEB). The findings indicate that differences in the length of text are dependent on the type of material that is transcribed and the grade…
Block-iterative finite element computations for incompressible flow problems
International Nuclear Information System (INIS)
Tezduyar, T.E.; Liou, J.; Glowinski, R.; Nguyen, T.; Poole, S.
1988-01-01
A block-iterative finite element procedure is presented for two-dimensional fluid dynamics computations on multiply-connected domains based on the vorticity-stream function formulation of the incompressible Navier-Stokes equations. The difficulty associated with the convection term in the vorticity transport equation is addressed by using a streamline-upwind/Petrov-Galerkin scheme. Element-by-element preconditioned iteration techniques with high degree of vectorization and high computational speed are employed to solve the linear equation system for each block. The authors conclude that performance evaluations show the potential of these techniques to be used for large-scale computations
International Nuclear Information System (INIS)
Anderson, D.V.; Breazeal, J.; Finan, C.H.; Johnston, B.M.
1976-01-01
ABCXYZ is a computer code for obtaining the Cartesian components of the vector potential and the magnetic field on an observed grid from an arrangement of current-carrying wires. Arbitrary combinations of straight line segments, arcs, and loops are allowed in the specification of the currents. Arbitrary positions and orientations of the current-carrying elements are also allowed. Specification of the wire diameter permits the computation of well-defined fields, even in the interiors of the conductors. An optical feature generates magnetic field lines. Extensive graphical and printed output is available to the user including contour, grid-line, and field-line plots. 12 figures, 1 table
Coordinate-free sensorimotor processing: computing with population codes.
Morasso, Pietro G.; Sanguineti, Vittorio; Frisone, Francesco; Perico, Luca
1998-10-01
The purpose of the study is to outline a computational architecture for the intelligent processing of sensorimotor patterns. The focus is on the nature of the internal representations of the outside world which are necessary for planning and other goal-oriented functions. A model of cortical map dynamics and self-organization is proposed that integrates a number of concepts and methods partly explored in the field. The novelty and the biological plausibility is related to the global architecture which allows one to deal with sensorimotor patterns in a coordinate-free way, using population codes as distributed internal representations of external variables and the coupled dynamics of cortical maps as a general tool of trajectory formation. The basic computational features of the model are demonstrated in the case of articulatory speech synthesis and some of the metric properties are evaluated by means of simple simulation studies.
Measurements by activation foils and comparative computations by MCNP code
International Nuclear Information System (INIS)
Kyncl, J.
2008-01-01
Systematic study of the radioactive waste minimisation problem is subject of the SPHINX project. Its idea is that burning or transmutation of the waste inventory problematic part will be realized in a nuclear reactor the fuel of which is in the form of liquid fluorides. In frame of the project, several experiments have been performed with so-called inserted experimental channel. The channel was filled up by the fluorides mixture, surrounded by six fuel assemblies with moderator and placed into LR-0 reactor vessel. This formation was brought to critical state and measurement with activation foil detectors were carried out at selected positions of the inserted channel. Main aim of the measurements was to determine reaction rates for the detectors mentioned. For experiment evaluation, comparative computations were accomplished by code MCNP4a. The results obtained show that very often, computed values of reaction rates differ substantially from the values that were obtained from the experiment. This contribution deals with analysis of the reasons of these differences from the point of view of computations by Monte Carlo method. The analysis of concrete cases shows that the inaccuracy of reaction rate computed is caused mostly by three circumstances:-space region that is occupied by detector is relatively very small;- microscopic effective cross-section R(E) of the reaction changes strongly with energy just in the energy interval that gives the greatest contribution to the reaction; - in the energy interval that gives the greatest contribution to reaction rate, the error of the computed neutron flux is great. These circumstances evoke that the computation of reaction rate with casual accuracy submits extreme demands on computing time. (Author)
International Nuclear Information System (INIS)
Tso, C.F.; Hueggenberg, R.
2004-01-01
Drop testing and analysis are the two methods for demonstrating the performance of packages in hypothetical drop accident scenarios. The exact purpose of the tests and the analyses, and the relative prominence of the two in the license application, may depend on the Competent Authority and will vary between countries. The Finite Element Method (FEM) is a powerful analysis tool. A reliable finite element (FE) code when used correctly and appropriately, will allow a package's behaviour to be simulated reliably. With improvements in computing power, and in sophistication and reliability of FE codes, it is likely that FEM calculations will increasingly be used as evidence of drop test performance when seeking Competent Authority approval. What is lacking at the moment, however, is a standardised method of assessing a FE code in order to determine whether it is sufficiently reliable or pessimistic. To this end, the project Evaluation of Codes for Analysing the Drop Test Performance of Radioactive Material Transport Containers, funded by the European Commission Directorate-General XVII (now Directorate-General for Energy and Transport) and jointly performed by Arup and Gesellschaft fuer Nuklear-Behaelter mbH, was carried out in 1998. The work consisted of three components: Survey of existing finite element software, with a view to finding codes that may be capable of analysing drop test performance of radioactive material packages, and to produce an inventory of them. Develop a set of benchmark problems to evaluate software used for analysing the drop test performance of packages. Evaluate the finite element codes by testing them against the benchmarks This paper presents a summary of this work
Non-coding sequence retrieval system for comparative genomic analysis of gene regulatory elements
Directory of Open Access Journals (Sweden)
Temple Matthew H
2007-03-01
Full Text Available Abstract Background Completion of the human genome sequence along with other species allows for greater understanding of the biochemical mechanisms and processes that govern healthy as well as diseased states. The large size of the genome sequences has made them difficult to study using traditional methods. There are many studies focusing on the protein coding sequences, however, not much is known about the function of non-coding regions of the genome. It has been demonstrated that parts of the non-coding region play a critical role as gene regulatory elements. Enhancers that regulate transcription processes have been found in intergenic regions. Furthermore, it is observed that regulatory elements found in non-coding regions are highly conserved across different species. However, the analysis of these regulatory elements is not as straightforward as it may first seem. The development of a centralized resource that allows for the quick and easy retrieval of non-coding sequences from multiple species and is capable of handing multi-gene queries is critical for the analysis of non-coding sequences. Here we describe the development of a web-based non-coding sequence retrieval system. Results This paper presents a Non-Coding Sequences Retrieval System (NCSRS. The NCSRS is a web-based bioinformatics tool that performs fast and convenient retrieval of non-coding and coding sequences from multiple species related to a specific gene or set of genes. This tool has compiled resources from multiple sources into one easy to use and convenient web based interface. With no software installation necessary, the user needs only internet access to use this tool. Conclusion The unique features of this tool will be very helpful for those studying gene regulatory elements that exist in non-coding regions. The web based application can be accessed on the internet at: http://cell.rutgers.edu/ncsrs/.
Smith, J. A.; Peter, D. B.; Tromp, J.; Komatitsch, D.; Lefebvre, M. P.
2015-12-01
We present both SPECFEM3D_Cartesian and SPECFEM3D_GLOBE open-source codes, representing high-performance numerical wave solvers simulating seismic wave propagation for local-, regional-, and global-scale application. These codes are suitable for both forward propagation in complex media and tomographic imaging. Both solvers compute highly accurate seismic wave fields using the continuous Galerkin spectral-element method on unstructured meshes. Lateral variations in compressional- and shear-wave speeds, density, as well as 3D attenuation Q models, topography and fluid-solid coupling are all readily included in both codes. For global simulations, effects due to rotation, ellipticity, the oceans, 3D crustal models, and self-gravitation are additionally included. Both packages provide forward and adjoint functionality suitable for adjoint tomography on high-performance computing architectures. We highlight the most recent release of the global version which includes improved performance, simultaneous MPI runs, OpenCL and CUDA support via an automatic source-to-source transformation library (BOAST), parallel I/O readers and writers for databases using ADIOS and seismograms using the recently developed Adaptable Seismic Data Format (ASDF) with built-in provenance. This makes our spectral-element solvers current state-of-the-art, open-source community codes for high-performance seismic wave propagation on arbitrarily complex 3D models. Together with these solvers, we provide full-waveform inversion tools to image the Earth's interior at unprecedented resolution.
Elements of matrix modeling and computing with Matlab
White, Robert E
2006-01-01
As discrete models and computing have become more common, there is a need to study matrix computation and numerical linear algebra. Encompassing a diverse mathematical core, Elements of Matrix Modeling and Computing with MATLAB examines a variety of applications and their modeling processes, showing you how to develop matrix models and solve algebraic systems. Emphasizing practical skills, it creates a bridge from problems with two and three variables to more realistic problems that have additional variables. Elements of Matrix Modeling and Computing with MATLAB focuses on seven basic applicat
Moon, Hongsik
What is the impact of multicore and associated advanced technologies on computational software for science? Most researchers and students have multicore laptops or desktops for their research and they need computing power to run computational software packages. Computing power was initially derived from Central Processing Unit (CPU) clock speed. That changed when increases in clock speed became constrained by power requirements. Chip manufacturers turned to multicore CPU architectures and associated technological advancements to create the CPUs for the future. Most software applications benefited by the increased computing power the same way that increases in clock speed helped applications run faster. However, for Computational ElectroMagnetics (CEM) software developers, this change was not an obvious benefit - it appeared to be a detriment. Developers were challenged to find a way to correctly utilize the advancements in hardware so that their codes could benefit. The solution was parallelization and this dissertation details the investigation to address these challenges. Prior to multicore CPUs, advanced computer technologies were compared with the performance using benchmark software and the metric was FLoting-point Operations Per Seconds (FLOPS) which indicates system performance for scientific applications that make heavy use of floating-point calculations. Is FLOPS an effective metric for parallelized CEM simulation tools on new multicore system? Parallel CEM software needs to be benchmarked not only by FLOPS but also by the performance of other parameters related to type and utilization of the hardware, such as CPU, Random Access Memory (RAM), hard disk, network, etc. The codes need to be optimized for more than just FLOPs and new parameters must be included in benchmarking. In this dissertation, the parallel CEM software named High Order Basis Based Integral Equation Solver (HOBBIES) is introduced. This code was developed to address the needs of the
Delivering LHC software to HPC compute elements
Blomer, Jakob; Hardi, Nikola; Popescu, Radu
2017-01-01
In recent years, there was a growing interest in improving the utilization of supercomputers by running applications of experiments at the Large Hadron Collider (LHC) at CERN when idle cores cannot be assigned to traditional HPC jobs. At the same time, the upcoming LHC machine and detector upgrades will produce some 60 times higher data rates and challenge LHC experiments to use so far untapped compute resources. LHC experiment applications are tailored to run on high-throughput computing resources and they have a different anatomy than HPC applications. LHC applications comprise a core framework that allows hundreds of researchers to plug in their specific algorithms. The software stacks easily accumulate to many gigabytes for a single release. New releases are often produced on a daily basis. To facilitate the distribution of these software stacks to world-wide distributed computing resources, LHC experiments use a purpose-built, global, POSIX file system, the CernVM File System. CernVM-FS pre-processes dat...
Promiscuity of enhancer, coding and non-coding transcription functions in ultraconserved elements
Directory of Open Access Journals (Sweden)
Sanges Remo
2010-03-01
Full Text Available Abstract Background Ultraconserved elements (UCEs are highly constrained elements of mammalian genomes, whose functional role has not been completely elucidated yet. Previous studies have shown that some of them act as enhancers in mouse, while some others are expressed in both normal and cancer-derived human tissues. Only one UCE element so far was shown to present these two functions concomitantly, as had been observed in other isolated instances of single, non ultraconserved enhancer elements. Results We used a custom microarray to assess the levels of UCE transcription during mouse development and integrated these data with published microarray and next-generation sequencing datasets as well as with newly produced PCR validation experiments. We show that a large fraction of non-exonic UCEs is transcribed across all developmental stages examined from only one DNA strand. Although the nature of these transcripts remains a mistery, our meta-analysis of RNA-Seq datasets indicates that they are unlikely to be short RNAs and that some of them might encode nuclear transcripts. In the majority of cases this function overlaps with the already established enhancer function of these elements during mouse development. Utilizing several next-generation sequencing datasets, we were further able to show that the level of expression observed in non-exonic UCEs is significantly higher than in random regions of the genome and that this is also seen in other regions which act as enhancers. Conclusion Our data shows that the concurrent presence of enhancer and transcript function in non-exonic UCE elements is more widespread than previously shown. Moreover through our own experiments as well as the use of next-generation sequencing datasets, we were able to show that the RNAs encoded by non-exonic UCEs are likely to be long RNAs transcribed from only one DNA strand.
Good, Jonathon; Keenan, Sarah; Mishra, Punya
2016-01-01
The popular press is rife with examples of how students in the United States and around the globe are learning to program, make, and tinker. The Hour of Code, maker-education, and similar efforts are advocating that more students be exposed to principles found within computer science. We propose an expansion beyond simply teaching computational…
FLICA-4 (version 1) a computer code for three dimensional thermal analysis of nuclear reactor cores
International Nuclear Information System (INIS)
Raymond, P.; Allaire, G.; Boudsocq, G.
1995-01-01
FLICA-4 is a thermal-hydraulic computer code developed at the French Energy Atomic Commission (CEA) for three dimensional steady state or transient two phase flow for design and safety thermal analysis of nuclear reactor cores. The two phase flow model of FLICA-4 is based on four balance equations for the fluid which includes: three balance equations for the mixture and a mass balance equation for the less concentrated phase which permits the calculation of non-equilibrium flows as sub cooled boiling and superheated steam. A drift velocity model takes into account the velocity disequilibrium between phases. The thermal behaviour of fuel elements can be computed by a one dimensional heat conduction equation in plane, cylindrical or spherical geometries and coupled to the fluid flow calculation. Convection and diffusion of solution products which are transported either by the liquid or by the gas, can be evaluated by solving specific mass conservation equations. A one dimensional two phase flow model can also be used to compute 1-D flow in pipes, guide tubes, BWR assemblies or RBMK channels. The FLICA-4 computer code uses fast running time steam-water functions. Phasic and saturation physical properties are computed by using bi-cubic spline functions. Polynomial coefficients are tabulated from 0.1 to 22 MPa and 0 to 800 degrees C. Specific modules can be utilised in order to generate the spline coefficients for any other fluid properties
International Nuclear Information System (INIS)
Yamada, Hiroyuki; Tsutsumi, Hideaki; Ebisawa, Katsumi; Suzuki, Masahide
2002-03-01
The SHEAT code developed at Japan Atomic Energy Research Institute is for probabilistic seismic hazard analysis which is one of the tasks needed for seismic Probabilistic Safety Assessment (PSA) of a nuclear power plant. At first, SHEAT was developed as the large sized computer version. In addition, a personal computer version was provided to improve operation efficiency and generality of this code in 2001. It is possible to perform the earthquake hazard analysis, display and the print functions with the Graphical User Interface. With the SHEAT for PC code, seismic hazard which is defined as an annual exceedance frequency of occurrence of earthquake ground motions at various levels of intensity at a given site is calculated by the following two steps as is done with the large sized computer. One is the modeling of earthquake generation around a site. Future earthquake generation (locations, magnitudes and frequencies of postulated earthquake) is modeled based on the historical earthquake records, active fault data and expert judgment. Another is the calculation of probabilistic seismic hazard at the site. An earthquake ground motion is calculated for each postulated earthquake using an attenuation model taking into account its standard deviation. Then the seismic hazard at the site is calculated by summing the frequencies of ground motions by all the earthquakes. This document is the user's manual of the SHEAT for PC code. It includes: (1) Outline of the code, which include overall concept, logical process, code structure, data file used and special characteristics of code, (2) Functions of subprogram and analytical models in them, (3) Guidance of input and output data, (4) Sample run result, and (5) Operational manual. (author)
A computer code to simulate X-ray imaging techniques
Energy Technology Data Exchange (ETDEWEB)
Duvauchelle, Philippe E-mail: philippe.duvauchelle@insa-lyon.fr; Freud, Nicolas; Kaftandjian, Valerie; Babot, Daniel
2000-09-01
A computer code was developed to simulate the operation of radiographic, radioscopic or tomographic devices. The simulation is based on ray-tracing techniques and on the X-ray attenuation law. The use of computer-aided drawing (CAD) models enables simulations to be carried out with complex three-dimensional (3D) objects and the geometry of every component of the imaging chain, from the source to the detector, can be defined. Geometric unsharpness, for example, can be easily taken into account, even in complex configurations. Automatic translations or rotations of the object can be performed to simulate radioscopic or tomographic image acquisition. Simulations can be carried out with monochromatic or polychromatic beam spectra. This feature enables, for example, the beam hardening phenomenon to be dealt with or dual energy imaging techniques to be studied. The simulation principle is completely deterministic and consequently the computed images present no photon noise. Nevertheless, the variance of the signal associated with each pixel of the detector can be determined, which enables contrast-to-noise ratio (CNR) maps to be computed, in order to predict quantitatively the detectability of defects in the inspected object. The CNR is a relevant indicator for optimizing the experimental parameters. This paper provides several examples of simulated images that illustrate some of the rich possibilities offered by our software. Depending on the simulation type, the computation time order of magnitude can vary from 0.1 s (simple radiographic projection) up to several hours (3D tomography) on a PC, with a 400 MHz microprocessor. Our simulation tool proves to be useful in developing new specific applications, in choosing the most suitable components when designing a new testing chain, and in saving time by reducing the number of experimental tests.
A computer code to simulate X-ray imaging techniques
International Nuclear Information System (INIS)
Duvauchelle, Philippe; Freud, Nicolas; Kaftandjian, Valerie; Babot, Daniel
2000-01-01
A computer code was developed to simulate the operation of radiographic, radioscopic or tomographic devices. The simulation is based on ray-tracing techniques and on the X-ray attenuation law. The use of computer-aided drawing (CAD) models enables simulations to be carried out with complex three-dimensional (3D) objects and the geometry of every component of the imaging chain, from the source to the detector, can be defined. Geometric unsharpness, for example, can be easily taken into account, even in complex configurations. Automatic translations or rotations of the object can be performed to simulate radioscopic or tomographic image acquisition. Simulations can be carried out with monochromatic or polychromatic beam spectra. This feature enables, for example, the beam hardening phenomenon to be dealt with or dual energy imaging techniques to be studied. The simulation principle is completely deterministic and consequently the computed images present no photon noise. Nevertheless, the variance of the signal associated with each pixel of the detector can be determined, which enables contrast-to-noise ratio (CNR) maps to be computed, in order to predict quantitatively the detectability of defects in the inspected object. The CNR is a relevant indicator for optimizing the experimental parameters. This paper provides several examples of simulated images that illustrate some of the rich possibilities offered by our software. Depending on the simulation type, the computation time order of magnitude can vary from 0.1 s (simple radiographic projection) up to several hours (3D tomography) on a PC, with a 400 MHz microprocessor. Our simulation tool proves to be useful in developing new specific applications, in choosing the most suitable components when designing a new testing chain, and in saving time by reducing the number of experimental tests
PEAKS: Computer code for solving partly overlapped photopeak in gamma spectrometry
International Nuclear Information System (INIS)
Jerez Vergueria, Sergio; Jerez Vergueria, Pablo
1996-01-01
The paper describes the main elements of the code according to purposes and contents. The PEAKS code is a useful tool of comfortable and easy handling for solving, partly overlapped photopeak in gamma spectrometry with NaI(Ti) detector
International Nuclear Information System (INIS)
Viswanathan, H.S.
1996-08-01
The finite element code FEHMN, developed by scientists at Los Alamos National Laboratory (LANL), is a three-dimensional finite element heat and mass transport simulator that can handle complex stratigraphy and nonlinear processes such as vadose zone flow, heat flow and solute transport. Scientists at LANL have been developing hydrologic flow and transport models of the Yucca Mountain site using FEHMN. Previous FEHMN simulations have used an equivalent Kd model to model solute transport. In this thesis, FEHMN is modified making it possible to simulate the transport of a species with a rigorous chemical model. Including the rigorous chemical equations into FEHMN simulations should provide for more representative transport models for highly reactive chemical species. A fully kinetic formulation is chosen for the FEHMN reactive transport model. Several methods are available to computationally implement a fully kinetic formulation. Different numerical algorithms are investigated in order to optimize computational efficiency and memory requirements of the reactive transport model. The best algorithm of those investigated is then incorporated into FEHMN. The algorithm chosen requires for the user to place strongly coupled species into groups which are then solved for simultaneously using FEHMN. The complete reactive transport model is verified over a wide variety of problems and is shown to be working properly. The new chemical capabilities of FEHMN are illustrated by using Los Alamos National Laboratory's site scale model of Yucca Mountain to model two-dimensional, vadose zone 14 C transport. The simulations demonstrate that gas flow and carbonate chemistry can significantly affect 14 C transport at Yucca Mountain. The simulations also prove that the new capabilities of FEHMN can be used to refine and buttress already existing Yucca Mountain radionuclide transport studies
International Nuclear Information System (INIS)
Couto, R.T.
1987-01-01
The implementation of the CP1 computer code in the Honeywell Bull computer in Brazilian Nuclear Energy Comission is presented. CP1 is a computer code used to solve the equations of punctual kinetic with Doppler feed back from the system temperature variation based on the Newton refrigeration equation (E.G.) [pt
A computer code for analysis of severe accidents in LWRs
International Nuclear Information System (INIS)
2001-01-01
The ICARE2 computer code, developed and validated since 1988 at IPSN (nuclear safety and protection institute), calculates in a mechanistic way the physical and chemical phenomena involved in the core degradation process during possible severe accidents in LWR's. The coupling between ICARE2 and the best-estimate thermal-hydraulics code CATHARE2 was completed at IPSN and led to the release of a first ICARE/CATHARE V1 version in 1999, followed by 2 successive revisions in 2000 and 2001. This documents gathers all the contributions presented at the first international ICARE/CATHARE users'club seminar that took place in November 2001. This seminar was characterized by a high quality and variety of the presentations, showing an increase of reactor applications and user needs in this area (2D/3D aspects, reflooding, corium slumping into the lower head,...). 2 sessions were organized. The first one was dedicated to the applications of ICARE2 V3mod1 against small-scale experiments such as PHEBUS FPT2 and FPT3 tests, PHEBUS AIC, QUENCH experiments, NRU-FLHT-5 test, ACRR-MP1 and DC1 experiments, CORA-PWR tests, and PBF-SFD1.4 test. The second session involved ICARE/CATHARE V1mod1 reactor applications and users'guidelines. Among reactor applications we found: code applicability to high burn-up fuel rods, simulation of the TMI-2 transient, simulation of a PWR-900 high pressure severe accident sequence, and the simulation of a VVER-1000 large break LOCA scenario. (A.C.)
A computer code for analysis of severe accidents in LWRs
Energy Technology Data Exchange (ETDEWEB)
NONE
2001-07-01
The ICARE2 computer code, developed and validated since 1988 at IPSN (nuclear safety and protection institute), calculates in a mechanistic way the physical and chemical phenomena involved in the core degradation process during possible severe accidents in LWR's. The coupling between ICARE2 and the best-estimate thermal-hydraulics code CATHARE2 was completed at IPSN and led to the release of a first ICARE/CATHARE V1 version in 1999, followed by 2 successive revisions in 2000 and 2001. This documents gathers all the contributions presented at the first international ICARE/CATHARE users'club seminar that took place in November 2001. This seminar was characterized by a high quality and variety of the presentations, showing an increase of reactor applications and user needs in this area (2D/3D aspects, reflooding, corium slumping into the lower head,...). 2 sessions were organized. The first one was dedicated to the applications of ICARE2 V3mod1 against small-scale experiments such as PHEBUS FPT2 and FPT3 tests, PHEBUS AIC, QUENCH experiments, NRU-FLHT-5 test, ACRR-MP1 and DC1 experiments, CORA-PWR tests, and PBF-SFD1.4 test. The second session involved ICARE/CATHARE V1mod1 reactor applications and users'guidelines. Among reactor applications we found: code applicability to high burn-up fuel rods, simulation of the TMI-2 transient, simulation of a PWR-900 high pressure severe accident sequence, and the simulation of a VVER-1000 large break LOCA scenario. (A.C.)
Computer code for simulating pressurized water reactor core
International Nuclear Information System (INIS)
Serrano, A.M.B.
1978-01-01
A computer code was developed for the simulation of the steady-state and transient behaviour of the average channel of a Pressurizer Water Reactor core. Point kinetics equations were used with the reactivity calculated for average temperatures in the channel with the fuel and moderator temperature feedbacks. The radial heat conduction equation in the fuel was solved numerically. For calculating the thermodynamic properties of the coolant, the fundamental equations of conservation (mass, energy and momentum) were solved. The gap and clad were treated as a resistance added to the film coefficient. The fuel system equations were decoupled from the coolant equations. The program permitted the changes in the heat transfer correlations and the flow patterns along the coolant channel. Various test were performed to determine the steady-state and transient response employing the PWR core simulator developed, obtaining results with adequate precision. (author)
A computer program for transient incompressible fluid flow based on quadratic finite elements
International Nuclear Information System (INIS)
Laval, H.
1981-01-01
The computer code CONVEC is a general computer program designed for the solution of transient two-dimensional incompressible fluid flow problems. The solution procedure is based on the finite element method. The class of problems treated by the present version of CONVEC are those described by the time-dependent, two-dimensional (plane or axisymmetric) form of the Navier-Stokes equations. The flow field is assumed to be laminar and the fluid Newtonian and incompressible within the Boussinesq approximation. The present report is intended to provide a description of the input data necessary to access and execute the code. In the following sections, the basic dimensions and input data of CONVEC are described. Then a test problem is solved in order to illustrate the input and output of the code
Application of the RESRAD computer code to VAMP scenario S
International Nuclear Information System (INIS)
Gnanapragasam, E.K.; Yu, C.
1997-03-01
The RESRAD computer code developed at Argonne National Laboratory was among 11 models from 11 countries participating in the international Scenario S validation of radiological assessment models with Chernobyl fallout data from southern Finland. The validation test was conducted by the Multiple Pathways Assessment Working Group of the Validation of Environmental Model Predictions (VAMP) program coordinated by the International Atomic Energy Agency. RESRAD was enhanced to provide an output of contaminant concentrations in environmental media and in food products to compare with measured data from southern Finland. Probability distributions for inputs that were judged to be most uncertain were obtained from the literature and from information provided in the scenario description prepared by the Finnish Centre for Radiation and Nuclear Safety. The deterministic version of RESRAD was run repeatedly to generate probability distributions for the required predictions. These predictions were used later to verify the probabilistic RESRAD code. The RESRAD predictions of radionuclide concentrations are compared with measured concentrations in selected food products. The radiological doses predicted by RESRAD are also compared with those estimated by the Finnish Centre for Radiation and Nuclear Safety
Application of the RESRAD computer code to VAMP scenario S
Energy Technology Data Exchange (ETDEWEB)
Gnanapragasam, E.K.; Yu, C.
1997-03-01
The RESRAD computer code developed at Argonne National Laboratory was among 11 models from 11 countries participating in the international Scenario S validation of radiological assessment models with Chernobyl fallout data from southern Finland. The validation test was conducted by the Multiple Pathways Assessment Working Group of the Validation of Environmental Model Predictions (VAMP) program coordinated by the International Atomic Energy Agency. RESRAD was enhanced to provide an output of contaminant concentrations in environmental media and in food products to compare with measured data from southern Finland. Probability distributions for inputs that were judged to be most uncertain were obtained from the literature and from information provided in the scenario description prepared by the Finnish Centre for Radiation and Nuclear Safety. The deterministic version of RESRAD was run repeatedly to generate probability distributions for the required predictions. These predictions were used later to verify the probabilistic RESRAD code. The RESRAD predictions of radionuclide concentrations are compared with measured concentrations in selected food products. The radiological doses predicted by RESRAD are also compared with those estimated by the Finnish Centre for Radiation and Nuclear Safety.
Computation of coupling elements between waveguides and modes for large launching structures
International Nuclear Information System (INIS)
Hurtak, O.; Preinhaelter, J.
1989-05-01
Optimization of large launching structures - grills and multijunction grills which are intended for lower hybrid current drive in big tokamaks - calls for quick and sufficiently precise computation of the coupling elements between waveguides and modes. The original method of computation is described which is efficient for small structures only and is based on integration along the real axis. For computation of coupling elements between more distant waveguides another method of integration is developed which uses a shift of the integration path. For this purpose, the analytical continuation of the plasma surface impedance into the complex N z -plane is calculated and a new numerical code developed. Some numerical results are given to compare both methods. A combination of both methods speeds up the computation needed for the grill design. (author). 2 figs., 2 tabs., 2 refs
Computer code for the atomistic simulation of lattice defects and dynamics. [COMENT code
Energy Technology Data Exchange (ETDEWEB)
Schiffgens, J.O.; Graves, N.J.; Oster, C.A.
1980-04-01
This document has been prepared to satisfy the need for a detailed, up-to-date description of a computer code that can be used to simulate phenomena on an atomistic level. COMENT was written in FORTRAN IV and COMPASS (CDC assembly language) to solve the classical equations of motion for a large number of atoms interacting according to a given force law, and to perform the desired ancillary analysis of the resulting data. COMENT is a dual-purpose intended to describe static defect configurations as well as the detailed motion of atoms in a crystal lattice. It can be used to simulate the effect of temperature, impurities, and pre-existing defects on radiation-induced defect production mechanisms, defect migration, and defect stability.
Users guide for NRC145-2 accident assessment computer code
International Nuclear Information System (INIS)
Pendergast, M.M.
1982-08-01
An accident assessment computer code has been developed for use at the Savannah River Plant. This computer code is based upon NRC Regulatory Guide 1.145 which provides guidence for accident assessements for power reactors. The code contains many options so that the user may utilize the code for many different assessments. For example the code can be used for non-nuclear assessments such as Sulpher Dioxide which may be required by the EPA. A discription of the code is contained in DP-1646. This document is a compilation of step-by-step instructions on how to use the code on the SRP IBM 3308 computer. This document consists of a number of tables which contain copies of computer listings. Some of the computer listings are copies of input; other listings give examples of computer output
Computation of Thermodynamic Equilibria Pertinent to Nuclear Materials in Multi-Physics Codes
Piro, Markus Hans Alexander
components at each iterative step, and the objective is to minimize the residuals of the mass balance equations. Several numerical advantages are achieved through this simplification. In particular, computational expense is reduced and the rate of convergence is enhanced. Furthermore, the software has demonstrated the ability to solve systems involving as many as 118 component elements. An early version of the code has already been integrated into the Advanced Multi-Physics (AMP) code under development by the Oak Ridge National Laboratory, Los Alamos National Laboratory, Idaho National Laboratory and Argonne National Laboratory. Keywords: Engineering, Nuclear -- 0552, Engineering, Material Science -- 0794, Chemistry, Mathematics -- 0405, Computer Science -- 0984
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.
Energy Technology Data Exchange (ETDEWEB)
Grebennikov, A.N.; Zhitnik, A.K.; Zvenigorodskaya, O.A. [and others
1995-12-31
In conformity with the protocol of the Workshop under Contract {open_quotes}Assessment of RBMK reactor safety using modern Western Codes{close_quotes} VNIIEF performed a neutronics computation series to compare western and VNIIEF codes and assess whether VNIIEF codes are suitable for RBMK type reactor safety assessment computation. The work was carried out in close collaboration with M.I. Rozhdestvensky and L.M. Podlazov, NIKIET employees. The effort involved: (1) cell computations with the WIMS, EKRAN codes (improved modification of the LOMA code) and the S-90 code (VNIIEF Monte Carlo). Cell, polycell, burnup computation; (2) 3D computation of static states with the KORAT-3D and NEU codes and comparison with results of computation with the NESTLE code (USA). The computations were performed in the geometry and using the neutron constants presented by the American party; (3) 3D computation of neutron kinetics with the KORAT-3D and NEU codes. These computations were performed in two formulations, both being developed in collaboration with NIKIET. Formulation of the first problem maximally possibly agrees with one of NESTLE problems and imitates gas bubble travel through a core. The second problem is a model of the RBMK as a whole with imitation of control and protection system controls (CPS) movement in a core.
International Nuclear Information System (INIS)
Dash, Z.V.; Robinson, B.A.; Zyvoloski, G.A.
1997-07-01
The requirements, design, and verification and validation of the software used in the FEHM application, a finite-element heat- and mass-transfer computer code that can simulate nonisothermal multiphase multicomponent flow in porous media, are described. The test of the DOE Code Comparison Project, Problem Five, Case A, which verifies that FEHM has correctly implemented heat and mass transfer and phase partitioning, is also covered
International Nuclear Information System (INIS)
Chan, M.K.; Ballinger, M.Y.; Owczarski, P.C.
1989-02-01
This manual describes the technical bases and use of the computer code FIRIN. This code was developed to estimate the source term release of smoke and radioactive particles from potential fires in nuclear fuel cycle facilities. FIRIN is a product of a broader study, Fuel Cycle Accident Analysis, which Pacific Northwest Laboratory conducted for the US Nuclear Regulatory Commission. The technical bases of FIRIN consist of a nonradioactive fire source term model, compartment effects modeling, and radioactive source term models. These three elements interact with each other in the code affecting the course of the fire. This report also serves as a complete FIRIN user's manual. Included are the FIRIN code description with methods/algorithms of calculation and subroutines, code operating instructions with input requirements, and output descriptions. 40 refs., 5 figs., 31 tabs
Padovan, J.; Adams, M.; Lam, P.; Fertis, D.; Zeid, I.
1982-01-01
Second-year efforts within a three-year study to develop and extend finite element (FE) methodology to efficiently handle the transient/steady state response of rotor-bearing-stator structure associated with gas turbine engines are outlined. The two main areas aim at (1) implanting the squeeze film damper element into a general purpose FE code for testing and evaluation; and (2) determining the numerical characteristics of the FE-generated rotor-bearing-stator simulation scheme. The governing FE field equations are set out and the solution methodology is presented. The choice of ADINA as the general-purpose FE code is explained, and the numerical operational characteristics of the direct integration approach of FE-generated rotor-bearing-stator simulations is determined, including benchmarking, comparison of explicit vs. implicit methodologies of direct integration, and demonstration problems.
RADTRAN II: revised computer code to analyze transportation of radioactive material
International Nuclear Information System (INIS)
Taylor, J.M.; Daniel, S.L.
1982-10-01
A revised and updated version of the RADTRAN computer code is presented. This code has the capability to predict the radiological impacts associated with specific schemes of radioactive material shipments and mode specific transport variables
Evaluation of QUENCH 12 test with ASTEC1.3R2 computer code
International Nuclear Information System (INIS)
Stefanova, A.; Groudev, P.
2009-01-01
This paper presents an assessment of QUENCH 12 test with ASTEC1.3R2 computer code. The test was performed to investigate the behavior of VVER fuel assemblies with a hexagonal lattice and containing fuel rod claddings made of Zrl%Nb(E 110) as used in eastern VVER reactors. The fuel is represented by ZrO 2 pellets. This investigation is a part of the 6-th framework program of the European Community supported ISTC program. The test facility is located at Forschungszentrum in Karlsruhe. The structure of the test facility allows experimental studies under transient and accident conditions. The purpose of QUENCH experiments is to investigate the hydrogen source term resulting from the water or steam injection into an uncovered degraded core of a Light-Water Reactor (LWR), to examine the behavior of overheated fuel elements under different flooding conditions, and to create a data base for model development and improvement of Severe Fuel Damage code packages. The ASTEC1.3R2 computer codes have been used to simulate the investigated transient. The base line input model for ASTEC was provided from Forschungszentrum, Karlsruhe. During the preparation of QUENCH 12 test, the input deck was adapted to new initial and boundary conditions. The new geometry of VVER bundle, grid spacer and corner rods was corrected. The comparison show very good agreement between measured data and ASTEC results. (authors)
Validation of a tetrahedral spectral element code for solving the Navier Stokes equation
International Nuclear Information System (INIS)
Niewiadomski, C.; Paraschivoiu, M.
2004-01-01
The tetrahedral spectral element method is considered to solve the incompressible Navier-Stokes equations because it is capable to capture complex geometries and obtain highly accurate solutions. This method allows accuracy improvements both by decreasing the spatial discretization as well as increasing the expansion order. The method is presented here-in as a modification of an standard finite element code. Some recent improvement to the baseline spectral element method for the tetrahedron described in References 3 and 2 are presented. These improvements include: the continuity enforcement procedure avoiding the need to change the global assembly operation and the removal of the reference coordinate system from the elemental evaluations thus simplifying greatly the method. A study is performed on the Stokes and Navier-Stokes equations to validate the method and the resulting code. (author)
International Nuclear Information System (INIS)
Tayal, M.
1987-01-01
Structures often operate at elevated temperatures. Temperature calculations are needed so that the design can accommodate thermally induced stresses and material changes. A finite element computer called FEAT has been developed to calculate temperatures in solids of arbitrary shapes. FEAT solves the classical equation for steady state conduction of heat. The solution is obtained for two-dimensional (plane or axisymmetric) or for three-dimensional problems. Gap elements are use to simulate interfaces between neighbouring surfaces. The code can model: conduction; internal generation of heat; prescribed convection to a heat sink; prescribed temperatures at boundaries; prescribed heat fluxes on some surfaces; and temperature-dependence of material properties like thermal conductivity. The user has a option of specifying the detailed variation of thermal conductivity with temperature. For convenience to the nuclear fuel industry, the user can also opt for pre-coded values of thermal conductivity, which are obtained from the MATPRO data base (sponsored by the U.S. Nuclear Regulatory Commission). The finite element method makes FEAT versatile, and enables it to accurately accommodate complex geometries. The optional link to MATPRO makes it convenient for the nuclear fuel industry to use FEAT, without loss of generality. Special numerical techniques make the code inexpensive to run, for the type of material non-linearities often encounter in the analysis of nuclear fuel. The code, however, is general, and can be used for other components of the reactor, or even for non-nuclear systems. The predictions of FEAT have been compared against several analytical solutions. The agreement is usually better than 5%. Thermocouple measurements show that the FEAT predictions are consistent with measured changes in temperatures in simulated pressure tubes. FEAT was also found to predict well, the axial variations in temperatures in the end-pellets(UO 2 ) of two fuel elements irradiated
CASKETSS: a computer code system for thermal and structural analysis of nuclear fuel shipping casks
International Nuclear Information System (INIS)
Ikushima, Takeshi
1989-02-01
A computer program CASKETSS has been developed for the purpose of thermal and structural analysis of nuclear fuel shipping casks. CASKETSS measn a modular code system for CASK Evaluation code system Thermal and Structural Safety. Main features of CASKETSS are as follow; (1) Thermal and structural analysis computer programs for one-, two-, three-dimensional geometries are contained in the code system. (2) Some of the computer programs in the code system has been programmed to provide near optimal speed on vector processing computers. (3) Data libralies fro thermal and structural analysis are provided in the code system. (4) Input data generator is provided in the code system. (5) Graphic computer program is provided in the code system. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)
International Nuclear Information System (INIS)
Russell, S.B.; Kempe, T.F.; Donnelly, K.J.
1985-05-01
Computer codes for modelling the dispersion and transfer of tritium released to the atmosphere were compared. The codes originated from Canada, the United States, Sweden and Japan. The comparisons include acute and chronic emissions of tritiated water vapour or elemental tritium from a hypothetical nuclear facility. Individual and collective doses to the population within 100 km of the site were calculated. The discrepancies among the code predictions were about one order of magnitude for the HTO emissions but were significantly more varied for the HT emissions. Codes that did not account for HT to HTO conversion and cycling of tritium in the environment predicted doses that were several orders of magnitude less than codes that incorporate this feature into the model. A field experiment consisting of the release of tritium gas and subsequent measurements of tritium concentrations in the environment has been recommended to validate the tritium transport models and code predictions and to refine the values of key parameters in the models. It is further recommended that the results from a field experiment be used in a follow-up code validation study to confirm which conceptual model and derived computer code provides the best representation of tritium transport in the real system
Computational and Experimental Methods to Decipher the Epigenetic Code
Directory of Open Access Journals (Sweden)
Stefano ede Pretis
2014-09-01
Full Text Available A multi-layered set of epigenetic marks, including post-translational modifications of histones and methylation of DNA, is finely tuned to define the epigenetic state of chromatin in any given cell type under specific conditions. Recently, the knowledge about the combinations of epigenetic marks occurring in the genome of different cell types under various conditions is rapidly increasing. Computational methods were developed for the identification of these states, unraveling the combinatorial nature of epigenetic marks and their association to genomic functional elements and transcriptional states. Nevertheless, the precise rules defining the interplay between all these marks remain poorly characterized. In this perspective we review the current state of this research field, illustrating the power and the limitations of current approaches. Finally, we sketch future avenues of research illustrating how the adoption of specific experimental designs coupled with available experimental approaches could be critical for a significant progress in this area.
User's Manual for the FEHM Application-A Finite-Element Heat- and Mass-Transfer Code
Energy Technology Data Exchange (ETDEWEB)
George A. Zyvoloski; Bruce A. Robinson; Zora V. Dash; Lynn L. Trease
1997-07-07
This document is a manual for the use of the FEHM application, a finite-element heat- and mass-transfer computer code that can simulate nonisothermal multiphase multicomponent flow in porous media. The use of this code is applicable to natural-state studies of geothermal systems and groundwater flow. A primary use of the FEHM application will be to assist in the understanding of flow fields and mass transport in the saturated and unsaturated zones below the proposed Yucca Mountain nuclear waste repository in Nevada. The equations of heat and mass transfer for multiphase flow in porous and permeable media are solved in the FEHM application by using the finite-element method. The permeability and porosity of the medium are allowed to depend on pressure and temperature. The code also has provisions for movable air and water phases and noncoupled tracers; that is, tracer solutions that do not affect the heat- and mass-transfer solutions. The tracers can be passive or reactive. The code can simulate two-dimensional, two-dimensional radial, or three-dimensional geometries. In fact, FEHM is capable of describing flow that is dominated in many areas by fracture and fault flow, including the inherently three-dimensional flow that results from permeation to and from faults and fractures. The code can handle coupled heat and mass-transfer effects, such as boiling, dryout, and condensation that can occur in the near-field region surrounding the potential repository and the natural convection that occurs through Yucca Mountain due to seasonal temperature changes. The code is also capable of incorporating the various adsorption mechanisms, ranging from simple linear relations to nonlinear isotherms, needed to describe the very complex transport processes at Yucca Mountain. This report outlines the uses and capabilities of the FEHM application, initialization of code variables, restart procedures, and error processing. The report describes all the data files, the input data
Lopez-Piñeiro, A.; Sanchez, M. L.; Moreno, B.
1992-06-01
The computer program MORSMATEL has been developed to calculate vibrational-rotational matrix elements of several r-dependent operators of two Morse oscillators. This code is based on a set of recurrence relations which are valid for any value of the power and of the quantum numbers v and J of each oscillator.
International Nuclear Information System (INIS)
Grebennikov, A.N.; Zhitnik, A.K.; Zvenigorodskaya, O.A.
1995-01-01
In conformity with the protocol of the Workshop under Contract open-quotes Assessment of RBMK reactor safety using modern Western Codesclose quotes VNIIEF performed a neutronics computation series to compare western and VNIIEF codes and assess whether VNIIEF codes are suitable for RBMK type reactor safety assessment computation. The work was carried out in close collaboration with M.I. Rozhdestvensky and L.M. Podlazov, NIKIET employees. The effort involved: (1) cell computations with the WIMS, EKRAN codes (improved modification of the LOMA code) and the S-90 code (VNIIEF Monte Carlo). Cell, polycell, burnup computation; (2) 3D computation of static states with the KORAT-3D and NEU codes and comparison with results of computation with the NESTLE code (USA). The computations were performed in the geometry and using the neutron constants presented by the American party; (3) 3D computation of neutron kinetics with the KORAT-3D and NEU codes. These computations were performed in two formulations, both being developed in collaboration with NIKIET. Formulation of the first problem maximally possibly agrees with one of NESTLE problems and imitates gas bubble travel through a core. The second problem is a model of the RBMK as a whole with imitation of control and protection system controls (CPS) movement in a core
Verification study of the FORE-2M nuclear/thermal-hydraulilc analysis computer code
International Nuclear Information System (INIS)
Coffield, R.D.; Tang, Y.S.; Markley, R.A.
1982-01-01
The verification of the LMFBR core transient performance code, FORE-2M, was performed in two steps. Different components of the computation (individual models) were verified by comparing with analytical solutions and with results obtained from other conventionally accepted computer codes (e.g., TRUMP, LIFE, etc.). For verification of the integral computation method of the code, experimental data in TREAT, SEFOR and natural circulation experiments in EBR-II were compared with the code calculations. Good agreement was obtained for both of these steps. Confirmation of the code verification for undercooling transients is provided by comparisons with the recent FFTF natural circulation experiments. (orig.)
Finite Macro-Element Mesh Deformation in a Structured Multi-Block Navier-Stokes Code
Bartels, Robert E.
2005-01-01
A mesh deformation scheme is developed for a structured multi-block Navier-Stokes code consisting of two steps. The first step is a finite element solution of either user defined or automatically generated macro-elements. Macro-elements are hexagonal finite elements created from a subset of points from the full mesh. When assembled, the finite element system spans the complete flow domain. Macro-element moduli vary according to the distance to the nearest surface, resulting in extremely stiff elements near a moving surface and very pliable elements away from boundaries. Solution of the finite element system for the imposed boundary deflections generally produces smoothly varying nodal deflections. The manner in which distance to the nearest surface has been found to critically influence the quality of the element deformation. The second step is a transfinite interpolation which distributes the macro-element nodal deflections to the remaining fluid mesh points. The scheme is demonstrated for several two-dimensional applications.
International Nuclear Information System (INIS)
Li Hai; Huang Chen; Du Aibing; Xu Baoyu
2014-01-01
The thermal conductivity is one of the most important parameters in the computer code for performance prediction for fuel rods. Several fuel thermal conductivity models used in foreign computer code, including thermal conductivity models for MOX fuel and UO 2 fuel were introduced in this paper. Thermal conductivities were calculated by using these models, and the results were compared and analyzed. Finally, the thermal conductivity model for the native computer code for performance prediction for fuel rods in fast reactor was recommended. (authors)
Axisym finite element code: modifications for pellet-cladding mechanical interaction analysis
International Nuclear Information System (INIS)
Engelman, G.P.
1978-10-01
Local strain concentrations in nuclear fuel rods are known to be potential sites for failure initiation. Assessment of such strain concentrations requires a two-dimensional analysis of stress and strain in both the fuel and the cladding during pellet-cladding mechanical interaction. To provide such a capability in the FRAP (Fuel Rod Analysis Program) codes, the AXISYM code (a small finite element program developed at the Idaho National Engineering Laboratory) was modified to perform a detailed fuel rod deformation analysis. This report describes the modifications which were made to the AXISYM code to adapt it for fuel rod analysis and presents comparisons made between the two-dimensional AXISYM code and the FRACAS-II code. FRACAS-II is the one-dimensional (generalized plane strain) fuel rod mechanical deformation subcode used in the FRAP codes. Predictions of these two codes should be comparable away from the fuel pellet free ends if the state of deformation at the pellet midplane is near that of generalized plane strain. The excellent agreement obtained in these comparisons checks both the correctness of the AXISYM code modifications as well as the validity of the assumption of generalized plane strain upon which the FRACAS-II subcode is based
Internal radiation dose calculations with the INREM II computer code
International Nuclear Information System (INIS)
Dunning, D.E. Jr.; Killough, G.G.
1978-01-01
A computer code, INREM II, was developed to calculate the internal radiation dose equivalent to organs of man which results from the intake of a radionuclide by inhalation or ingestion. Deposition and removal of radioactivity from the respiratory tract is represented by the Internal Commission on Radiological Protection Task Group Lung Model. A four-segment catenary model of the gastrointestinal tract is used to estimate movement of radioactive material that is ingested, or swallowed after being cleared from the respiratory tract. Retention of radioactivity in other organs is specified by linear combinations of decaying exponential functions. The formation and decay of radioactive daughters is treated explicitly, with each radionuclide in the decay chain having its own uptake and retention parameters, as supplied by the user. The dose equivalent to a target organ is computed as the sum of contributions from each source organ in which radioactivity is assumed to be situated. This calculation utilizes a matrix of dosimetric S-factors (rem/μCi-day) supplied by the user for the particular choice of source and target organs. Output permits the evaluation of components of dose from cross-irradiations when penetrating radiations are present. INREM II has been utilized with current radioactive decay data and metabolic models to produce extensive tabulations of dose conversion factors for a reference adult for approximately 150 radionuclides of interest in environmental assessments of light-water-reactor fuel cycles. These dose conversion factors represent the 50-year dose commitment per microcurie intake of a given radionuclide for 22target organs including contributions from specified source organs and surplus activity in the rest of the body. These tabulations are particularly significant in their consistent use of contemporary models and data and in the detail of documentation
SQA of finite element method (FEM) codes used for analyses of pit storage/transport packages
Energy Technology Data Exchange (ETDEWEB)
Russel, E. [Lawrence Livermore National Lab., CA (United States)
1997-11-01
This report contains viewgraphs on the software quality assurance of finite element method codes used for analyses of pit storage and transport projects. This methodology utilizes the ISO 9000-3: Guideline for application of 9001 to the development, supply, and maintenance of software, for establishing well-defined software engineering processes to consistently maintain high quality management approaches.
A sliding point contact model for the finite element structures code EURDYN
International Nuclear Information System (INIS)
Smith, B.L.
1986-01-01
A method is developed by which sliding point contact between two moving deformable structures may be incorporated within a lumped mass finite element formulation based on displacements. The method relies on a simple mechanical interpretation of the contact constraint in terms of equivalent nodal forces and avoids the use of nodal connectivity via a master slave arrangement or pseudo contact element. The methodology has been iplemented into the EURDYN finite element program for the (2D axisymmetric) version coupled to the hydro code SEURBNUK. Sample calculations are presented illustrating the use of the model in various contact situations. Effects due to separation and impact of structures are also included. (author)
International Nuclear Information System (INIS)
Candel, A.; Kabel, A.; Ko, K.; Lee, L.; Li, Z.; Limborg, C.; Ng, C.; Prudencio, E.; Schussman, G.; Uplenchwar, R.
2007-01-01
Over the past years, SLAC's Advanced Computations Department (ACD) has developed the parallel finite element (FE) particle-in-cell code Pic3P (Pic2P) for simulations of beam-cavity interactions dominated by space-charge effects. As opposed to standard space-charge dominated beam transport codes, which are based on the electrostatic approximation, Pic3P (Pic2P) includes space-charge, retardation and boundary effects as it self-consistently solves the complete set of Maxwell-Lorentz equations using higher-order FE methods on conformal meshes. Use of efficient, large-scale parallel processing allows for the modeling of photoinjectors with unprecedented accuracy, aiding the design and operation of the next-generation of accelerator facilities. Applications to the Linac Coherent Light Source (LCLS) RF gun are presented
The TESS [Tandem Experiment Simulation Studies] computer code user's manual
International Nuclear Information System (INIS)
Procassini, R.J.
1990-01-01
TESS (Tandem Experiment Simulation Studies) is a one-dimensional, bounded particle-in-cell (PIC) simulation code designed to investigate the confinement and transport of plasma in a magnetic mirror device, including tandem mirror configurations. Mirror plasmas may be modeled in a system which includes an applied magnetic field and/or a self-consistent or applied electrostatic potential. The PIC code TESS is similar to the PIC code DIPSI (Direct Implicit Plasma Surface Interactions) which is designed to study plasma transport to and interaction with a solid surface. The codes TESS and DIPSI are direct descendants of the PIC code ES1 that was created by A. B. Langdon. This document provides the user with a brief description of the methods used in the code and a tutorial on the use of the code. 10 refs., 2 tabs
A study on the nuclear computer codes installation and management system
International Nuclear Information System (INIS)
Kim, Yeon Seung; Huh, Young Hwan; Kim, Hee Kyung; Kang, Byung Heon; Kim, Ko Ryeo; Suh, Soong Hyok; Choi, Young Gil; Lee, Jong Bok
1990-12-01
From 1987 a number of technical transfer related to nuclear power plant had been performed from C-E for YGN 3 and 4 construction. Among them, installation and management of the computer codes for YGN 3 and 4 fuel and nuclear steam supply system was one of the most important project. Main objectives of this project are to establish the nuclear computer code management system, to develop QA procedure for nuclear codes, to secure the nuclear code reliability and to extend techanical applicabilities including the user-oriented utility programs for nuclear codes. Contents of performing the project in this year was to produce 215 transmittal packages of nuclear codes installation including making backup magnetic tape and microfiche for software quality assurance. Lastly, for easy reference about the nuclear codes information we presented list of code names and information on the codes which were introduced from C-E. (Author)
SPLAI: Computational Finite Element Model for Sensor Networks
Directory of Open Access Journals (Sweden)
Ruzana Ishak
2006-01-01
Full Text Available Wireless sensor network refers to a group of sensors, linked by a wireless medium to perform distributed sensing task. The primary interest is their capability in monitoring the physical environment through the deployment of numerous tiny, intelligent, wireless networked sensor nodes. Our interest consists of a sensor network, which includes a few specialized nodes called processing elements that can perform some limited computational capabilities. In this paper, we propose a model called SPLAI that allows the network to compute a finite element problem where the processing elements are modeled as the nodes in the linear triangular approximation problem. Our model also considers the case of some failures of the sensors. A simulation model to visualize this network has been developed using C++ on the Windows environment.
Assessment of the computer code COBRA/CFTL
International Nuclear Information System (INIS)
Baxi, C.B.; Burhop, C.J.
1981-07-01
The COBRA/CFTL code has been developed by Oak Ridge National Laboratory (ORNL) for thermal-hydraulic analysis of simulated gas-cooled fast breeder reactor (GCFR) core assemblies to be tested in the core flow test loop (CFTL). The COBRA/CFTL code was obtained by modifying the General Atomic code COBRA*GCFR. This report discusses these modifications, compares the two code results for three cases which represent conditions from fully rough turbulent flow to laminar flow. Case 1 represented fully rough turbulent flow in the bundle. Cases 2 and 3 represented laminar and transition flow regimes. The required input for the COBRA/CFTL code, a sample problem input/output and the code listing are included in the Appendices
Assessment of the computer code COBRA/CFTL
Energy Technology Data Exchange (ETDEWEB)
Baxi, C. B.; Burhop, C. J.
1981-07-01
The COBRA/CFTL code has been developed by Oak Ridge National Laboratory (ORNL) for thermal-hydraulic analysis of simulated gas-cooled fast breeder reactor (GCFR) core assemblies to be tested in the core flow test loop (CFTL). The COBRA/CFTL code was obtained by modifying the General Atomic code COBRA*GCFR. This report discusses these modifications, compares the two code results for three cases which represent conditions from fully rough turbulent flow to laminar flow. Case 1 represented fully rough turbulent flow in the bundle. Cases 2 and 3 represented laminar and transition flow regimes. The required input for the COBRA/CFTL code, a sample problem input/output and the code listing are included in the Appendices.
Status of the CONTAIN computer code for LWR containment analysis
International Nuclear Information System (INIS)
Bergeron, K.D.; Murata, K.K.; Rexroth, P.E.; Clauser, M.J.; Senglaub, M.E.; Sciacca, F.W.; Trebilcock, W.
1982-01-01
The current status of the CONTAIN code for LWR safety analysis is reviewed. Three example calculations are discussed as illustrations of the code's capabilities: (1) a demonstration of the spray model in a realistic PWR problem, and a comparison with CONTEMPT results; (2) a comparison of CONTAIN results for a major aerosol experiment against experimental results and predictions of the HAARM aerosol code; and (3) an LWR sample problem, involving a TMLB' sequence for the Zion reactor containment
Status of the CONTAIN computer code for LWR containment analysis
International Nuclear Information System (INIS)
Bergeron, K.D.; Murata, K.K.; Rexroth, P.E.; Clauser, M.J.; Senglaub, M.E.; Sciacca, F.W.; Trebilcock, W.
1983-01-01
The current status of the CONTAIN code for LWR safety analysis is reviewed. Three example calculations are discussed as illustrations of the code's capabilities: (1) a demonstration of the spray model in a realistic PWR problem, and a comparison with CONTEMPT results; (2) a comparison of CONTAIN results for a major aerosol experiment against experimental results and predictions of the HAARM aerosol code; and (3) an LWR sample problem, involving a TMLB' sequence for the Zion reactor containment
Energy Technology Data Exchange (ETDEWEB)
Bjerke, M.A.
1983-02-01
A package of computer codes has been developed to perform a nonlinear uncertainty analysis on transient thermal-hydraulic systems which are modeled with the RELAP computer code. Using an uncertainty around the analyses of experiments in the PWR-BDHT Separate Effects Program at Oak Ridge National Laboratory. The use of FORTRAN programs running interactively on the PDP-10 computer has made the system very easy to use and provided great flexibility in the choice of processing paths. Several experiments simulating a loss-of-coolant accident in a nuclear reactor have been successfully analyzed. It has been shown that the system can be automated easily to further simplify its use and that the conversion of the entire system to a base code other than RELAP is possible.
Finite element electromagnetic field computation on the Sequent Symmetry 81 parallel computer
International Nuclear Information System (INIS)
Ratnajeevan, S.; Hoole, H.
1990-01-01
Finite element field analysis algorithms lend themselves to parallelization and this fact is exploited in this paper to implement a finite element analysis program for electromagnetic field computation on the Sequent Symmetry 81 parallel computer with three processors. In terms of waiting time, the maximum gains are to be made in matrix solution and therefore this paper concentrates on the gains in parallelizing the solution part of finite element analysis. An outline of how parallelization could be exploited in most finite element operations is given in this paper although the actual implemention of parallelism on the Sequent Symmetry 81 parallel computer was in sparsity computation, matrix assembly and the matrix solution areas. In all cases, the algorithms were modified suit the parallel programming application rather than allowing the compiler to parallelize on existing algorithms
International Nuclear Information System (INIS)
Ohnishi, Y.; Shibata, H.; Kobayashi, A.
1985-01-01
A model is presented which describes fully coupled thermo-hydro-mechanical behavior of porous geologic medium. The mathematical formulation for the model utilizes the Biot theory for the consolidation and the energy balance equation. The medium is in the condition of saturated-unsaturated flow, then the free surfaces are taken into consideration in the model. The model, incorporated in a finite element numerical procedure, was implemented in a two-dimensional computer code. The code was developed under the assumptions that the medium is poro-elastic and in plane strain condition; water in the ground does not change its phase; heat is transferred by conductive and convective flow. Analytical solutions pertaining to consolidation theory for soils and rocks, thermoelasticity for solids and hydrothermal convection theory provided verification of stress and fluid flow couplings, respectively in the coupled model. Several types of problems are analyzed. The one is a study of some of the effects of completely coupled thermo-hydro-mechanical behavior on the response of a saturated-unsaturated porous rock containing a buried heat source. Excavation of an underground opening which has radioactive wastes at elevated temperatures is modeled and analyzed. The results shows that the coupling phenomena can be estimated at some degree by the numerical procedure. The computer code has a powerful ability to analyze of the repository the complex nature of the repository
International Nuclear Information System (INIS)
Silva, R. da.
1982-10-01
The computer code SUPERFISH has been implemented in CYBER - IEAv computer system. This code locates eletromagnetic modes in rf ressonant cavities. The manipulation of the boundary conditions and of the driving point was optimized. A computer program (ARRUELA) was developed in order to make easier SUPERFISH analysis of the rf properties of disc-and-washer cavities. This version of SUPERFISH showed satisfactory performance under tests. (Author) [pt
International Nuclear Information System (INIS)
Hanusik, V.; Kopcani, I.; Gedeon, M.
2000-01-01
This paper describes selection and adaptation of computer codes required to assess the effects of radionuclide release from Mochovce Radioactive Waste Disposal Facility. The paper also demonstrates how these codes can be integrated into performance assessment methodology. The considered codes include DUST-MS for source term release, MODFLOW for ground-water flow and BS for transport through biosphere and dose assessment. (author)
User manual for PACTOLUS: a code for computing power costs
International Nuclear Information System (INIS)
Huber, H.D.; Bloomster, C.H.
1979-02-01
PACTOLUS is a computer code for calculating the cost of generating electricity. Through appropriate definition of the input data, PACTOLUS can calculate the cost of generating electricity from a wide variety of power plants, including nuclear, fossil, geothermal, solar, and other types of advanced energy systems. The purpose of PACTOLUS is to develop cash flows and calculate the unit busbar power cost (mills/kWh) over the entire life of a power plant. The cash flow information is calculated by two principal models: the Fuel Model and the Discounted Cash Flow Model. The Fuel Model is an engineering cost model which calculates the cash flow for the fuel cycle costs over the project lifetime based on input data defining the fuel material requirements, the unit costs of fuel materials and processes, the process lead and lag times, and the schedule of the capacity factor for the plant. For nuclear plants, the Fuel Model calculates the cash flow for the entire nuclear fuel cycle. For fossil plants, the Fuel Model calculates the cash flow for the fossil fuel purchases. The Discounted Cash Flow Model combines the fuel costs generated by the Fuel Model with input data on the capital costs, capital structure, licensing time, construction time, rates of return on capital, tax rates, operating costs, and depreciation method of the plant to calculate the cash flow for the entire lifetime of the project. The financial and tax structure for both investor-owned utilities and municipal utilities can be simulated through varying the rates of return on equity and debt, the debt-equity ratios, and tax rates. The Discounted Cash Flow Model uses the principal that the present worth of the revenues will be equal to the present worth of the expenses including the return on investment over the economic life of the project. This manual explains how to prepare the input data, execute cases, and interpret the output results with the updated version of PACTOLUS. 11 figures, 2 tables
User manual for PACTOLUS: a code for computing power costs.
Energy Technology Data Exchange (ETDEWEB)
Huber, H.D.; Bloomster, C.H.
1979-02-01
PACTOLUS is a computer code for calculating the cost of generating electricity. Through appropriate definition of the input data, PACTOLUS can calculate the cost of generating electricity from a wide variety of power plants, including nuclear, fossil, geothermal, solar, and other types of advanced energy systems. The purpose of PACTOLUS is to develop cash flows and calculate the unit busbar power cost (mills/kWh) over the entire life of a power plant. The cash flow information is calculated by two principal models: the Fuel Model and the Discounted Cash Flow Model. The Fuel Model is an engineering cost model which calculates the cash flow for the fuel cycle costs over the project lifetime based on input data defining the fuel material requirements, the unit costs of fuel materials and processes, the process lead and lag times, and the schedule of the capacity factor for the plant. For nuclear plants, the Fuel Model calculates the cash flow for the entire nuclear fuel cycle. For fossil plants, the Fuel Model calculates the cash flow for the fossil fuel purchases. The Discounted Cash Flow Model combines the fuel costs generated by the Fuel Model with input data on the capital costs, capital structure, licensing time, construction time, rates of return on capital, tax rates, operating costs, and depreciation method of the plant to calculate the cash flow for the entire lifetime of the project. The financial and tax structure for both investor-owned utilities and municipal utilities can be simulated through varying the rates of return on equity and debt, the debt-equity ratios, and tax rates. The Discounted Cash Flow Model uses the principal that the present worth of the revenues will be equal to the present worth of the expenses including the return on investment over the economic life of the project. This manual explains how to prepare the input data, execute cases, and interpret the output results. (RWR)
MMA, A Computer Code for Multi-Model Analysis
Energy Technology Data Exchange (ETDEWEB)
Eileen P. Poeter and Mary C. Hill
2007-08-20
This report documents the Multi-Model Analysis (MMA) computer code. MMA can be used to evaluate results from alternative models of a single system using the same set of observations for all models. As long as the observations, the observation weighting, and system being represented are the same, the models can differ in nearly any way imaginable. For example, they may include different processes, different simulation software, different temporal definitions (for example, steady-state and transient models could be considered), and so on. The multiple models need to be calibrated by nonlinear regression. Calibration of the individual models needs to be completed before application of MMA. MMA can be used to rank models and calculate posterior model probabilities. These can be used to (1) determine the relative importance of the characteristics embodied in the alternative models, (2) calculate model-averaged parameter estimates and predictions, and (3) quantify the uncertainty of parameter estimates and predictions in a way that integrates the variations represented by the alternative models. There is a lack of consensus on what model analysis methods are best, so MMA provides four default methods. Two are based on Kullback-Leibler information, and use the AIC (Akaike Information Criterion) or AICc (second-order-bias-corrected AIC) model discrimination criteria. The other two default methods are the BIC (Bayesian Information Criterion) and the KIC (Kashyap Information Criterion) model discrimination criteria. Use of the KIC criterion is equivalent to using the maximum-likelihood Bayesian model averaging (MLBMA) method. AIC, AICc, and BIC can be derived from Frequentist or Bayesian arguments. The default methods based on Kullback-Leibler information have a number of theoretical advantages, including that they tend to favor more complicated models as more data become available than do the other methods, which makes sense in many situations.
Utilization of KENO-IV computer code with HANSEN-ROACH library
International Nuclear Information System (INIS)
Lima Barros, M. de; Vellozo, S.O.
1982-01-01
Several analysis with KENO-IV computer code, which is based in the Monte Carlo method, and the cross section library HANSEN-ROACH, were done, aiming to present the more convenient form to execute criticality calculations with this computer code and this cross sections. (E.G.) [pt
Computation of asteroid proper elements on the Grid
Directory of Open Access Journals (Sweden)
Novaković B.
2009-01-01
Full Text Available A procedure of gridification of the computation of asteroid proper orbital elements is described. The need to speed up the time consuming computations and make them more efficient is justified by the large increase of observational data expected from the next generation all sky surveys. We give the basic notion of proper elements and of the contemporary theories and methods used to compute them for different populations of objects. Proper elements for nearly 70,000 asteroids are derived since the beginning of use of the Grid infrastructure for the purpose. The average time for the catalogs update is significantly shortened with respect to the time needed with stand-alone workstations. We also present basics of the Grid computing, the concepts of Grid middleware and its Workload management system. The practical steps we undertook to efficiently gridify our application are described in full detail. We present the results of a comprehensive testing of the performance of different Grid sites, and offer some practical conclusions based on the benchmark results and on our experience. Finally, we propose some possibilities for the future work.
Computation of Asteroid Proper Elements on the Grid
Directory of Open Access Journals (Sweden)
Novaković, B.
2009-12-01
Full Text Available A procedure of gridification of the computation of asteroid proper orbital elements is described. The need to speed up the time consuming computations and make them more efficient is justified by the large increase of observational data expected from the next generation all sky surveys. We give the basic notion of proper elements and of the contemporary theories and methods used to compute them for different populations of objects. Proper elements for nearly 70,000 asteroids are derived since the beginning of use of the Grid infrastructure for the purpose. The average time for the catalogs update is significantly shortened with respect to the time needed with stand-alone workstations. We also present basics of the Grid computing, the concepts of Grid middleware and its Workload management system. The practical steps we undertook to efficiently gridify our application are described in full detail. We present the results of a comprehensive testing of the performance of different Grid sites, and offer some practical conclusions based on the benchmark results and on our experience. Finally, we propose some possibilities for the future work.
Energy Technology Data Exchange (ETDEWEB)
Glass, Micheal W.; Hogan, Roy E., Jr.; Gartling, David K.
2010-03-01
The need for the engineering analysis of systems in which the transport of thermal energy occurs primarily through a conduction process is a common situation. For all but the simplest geometries and boundary conditions, analytic solutions to heat conduction problems are unavailable, thus forcing the analyst to call upon some type of approximate numerical procedure. A wide variety of numerical packages currently exist for such applications, ranging in sophistication from the large, general purpose, commercial codes, such as COMSOL, COSMOSWorks, ABAQUS and TSS to codes written by individuals for specific problem applications. The original purpose for developing the finite element code described here, COYOTE, was to bridge the gap between the complex commercial codes and the more simplistic, individual application programs. COYOTE was designed to treat most of the standard conduction problems of interest with a user-oriented input structure and format that was easily learned and remembered. Because of its architecture, the code has also proved useful for research in numerical algorithms and development of thermal analysis capabilities. This general philosophy has been retained in the current version of the program, COYOTE, Version 5.0, though the capabilities of the code have been significantly expanded. A major change in the code is its availability on parallel computer architectures and the increase in problem complexity and size that this implies. The present document describes the theoretical and numerical background for the COYOTE program. This volume is intended as a background document for the user's manual. Potential users of COYOTE are encouraged to become familiar with the present report and the simple example analyses reported in before using the program. The theoretical and numerical background for the finite element computer program, COYOTE, is presented in detail. COYOTE is designed for the multi-dimensional analysis of nonlinear heat conduction
International Nuclear Information System (INIS)
Glass, Micheal W.; Hogan, Roy E. Jr.; Gartling, David K.
2010-01-01
The need for the engineering analysis of systems in which the transport of thermal energy occurs primarily through a conduction process is a common situation. For all but the simplest geometries and boundary conditions, analytic solutions to heat conduction problems are unavailable, thus forcing the analyst to call upon some type of approximate numerical procedure. A wide variety of numerical packages currently exist for such applications, ranging in sophistication from the large, general purpose, commercial codes, such as COMSOL, COSMOSWorks, ABAQUS and TSS to codes written by individuals for specific problem applications. The original purpose for developing the finite element code described here, COYOTE, was to bridge the gap between the complex commercial codes and the more simplistic, individual application programs. COYOTE was designed to treat most of the standard conduction problems of interest with a user-oriented input structure and format that was easily learned and remembered. Because of its architecture, the code has also proved useful for research in numerical algorithms and development of thermal analysis capabilities. This general philosophy has been retained in the current version of the program, COYOTE, Version 5.0, though the capabilities of the code have been significantly expanded. A major change in the code is its availability on parallel computer architectures and the increase in problem complexity and size that this implies. The present document describes the theoretical and numerical background for the COYOTE program. This volume is intended as a background document for the user's manual. Potential users of COYOTE are encouraged to become familiar with the present report and the simple example analyses reported in before using the program. The theoretical and numerical background for the finite element computer program, COYOTE, is presented in detail. COYOTE is designed for the multi-dimensional analysis of nonlinear heat conduction problems
Energy Technology Data Exchange (ETDEWEB)
Haverkort, J.W. [Centrum Wiskunde & Informatica, P.O. Box 94079, 1090 GB Amsterdam (Netherlands); Dutch Institute for Fundamental Energy Research, P.O. Box 6336, 5600 HH Eindhoven (Netherlands); Blank, H.J. de [Dutch Institute for Fundamental Energy Research, P.O. Box 6336, 5600 HH Eindhoven (Netherlands); Huysmans, G.T.A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Pratt, J. [Dutch Institute for Fundamental Energy Research, P.O. Box 6336, 5600 HH Eindhoven (Netherlands); Koren, B., E-mail: b.koren@tue.nl [Eindhoven University of Technology, P.O. Box 513, 5600 MB Eindhoven (Netherlands)
2016-07-01
Numerical simulations form an indispensable tool to understand the behavior of a hot plasma that is created inside a tokamak for providing nuclear fusion energy. Various aspects of tokamak plasmas have been successfully studied through the reduced magnetohydrodynamic (MHD) model. The need for more complete modeling through the full MHD equations is addressed here. Our computational method is presented along with measures against possible problems regarding pollution, stability, and regularity. The problem of ensuring continuity of solutions in the center of a polar grid is addressed in the context of a finite element discretization of the full MHD equations. A rigorous and generally applicable solution is proposed here. Useful analytical test cases are devised to verify the correct implementation of the momentum and induction equation, the hyperdiffusive terms, and the accuracy with which highly anisotropic diffusion can be simulated. A striking observation is that highly anisotropic diffusion can be treated with the same order of accuracy as isotropic diffusion, even on non-aligned grids, as long as these grids are generated with sufficient care. This property is shown to be associated with our use of a magnetic vector potential to describe the magnetic field. Several well-known instabilities are simulated to demonstrate the capabilities of the new method. The linear growth rate of an internal kink mode and a tearing mode are benchmarked against the results of a linear MHD code. The evolution of a tearing mode and the resulting magnetic islands are simulated well into the nonlinear regime. The results are compared with predictions from the reduced MHD model. Finally, a simulation of a ballooning mode illustrates the possibility to use our method as an ideal MHD method without the need to add any physical dissipation.
Efficient Proximity Computation Techniques Using ZIP Code Data for Smart Cities †
Directory of Open Access Journals (Sweden)
Muhammad Harist Murdani
2018-03-01
Full Text Available In this paper, we are interested in computing ZIP code proximity from two perspectives, proximity between two ZIP codes (Ad-Hoc and neighborhood proximity (Top-K. Such a computation can be used for ZIP code-based target marketing as one of the smart city applications. A naïve approach to this computation is the usage of the distance between ZIP codes. We redefine a distance metric combining the centroid distance with the intersecting road network between ZIP codes by using a weighted sum method. Furthermore, we prove that the results of our combined approach conform to the characteristics of distance measurement. We have proposed a general and heuristic approach for computing Ad-Hoc proximity, while for computing Top-K proximity, we have proposed a general approach only. Our experimental results indicate that our approaches are verifiable and effective in reducing the execution time and search space.
Efficient Proximity Computation Techniques Using ZIP Code Data for Smart Cities †.
Murdani, Muhammad Harist; Kwon, Joonho; Choi, Yoon-Ho; Hong, Bonghee
2018-03-24
In this paper, we are interested in computing ZIP code proximity from two perspectives, proximity between two ZIP codes ( Ad-Hoc ) and neighborhood proximity ( Top-K ). Such a computation can be used for ZIP code-based target marketing as one of the smart city applications. A naïve approach to this computation is the usage of the distance between ZIP codes. We redefine a distance metric combining the centroid distance with the intersecting road network between ZIP codes by using a weighted sum method. Furthermore, we prove that the results of our combined approach conform to the characteristics of distance measurement. We have proposed a general and heuristic approach for computing Ad-Hoc proximity, while for computing Top-K proximity, we have proposed a general approach only. Our experimental results indicate that our approaches are verifiable and effective in reducing the execution time and search space.
SWAAM-LT: The long-term, sodium/water reaction analysis method computer code
International Nuclear Information System (INIS)
Shin, Y.W.; Chung, H.H.; Wiedermann, A.H.; Tanabe, H.
1993-01-01
The SWAAM-LT Code, developed for analysis of long-term effects of sodium/water reactions, is discussed. The theoretical formulation of the code is described, including the introduction of system matrices for ease of computer programming as a general system code. Also, some typical results of the code predictions for available large scale tests are presented. Test data for the steam generator design with the cover-gas feature and without the cover-gas feature are available and analyzed. The capabilities and limitations of the code are then discussed in light of the comparison between the code prediction and the test data
Compendium of computer codes for the safety analysis of fast breeder reactors
International Nuclear Information System (INIS)
1977-10-01
The objective of the compendium is to provide the reader with a guide which briefly describes many of the computer codes used for liquid metal fast breeder reactor safety analyses, since it is for this system that most of the codes have been developed. The compendium is designed to address the following frequently asked questions from individuals in licensing and research and development activities: (1) What does the code do. (2) To what safety problems has it been applied. (3) What are the code's limitations. (4) What is being done to remove these limitations. (5) How does the code compare with experimental observations and other code predictions. (6) What reference documents are available
COMPUTER EXPERIMENTS WITH FINITE ELEMENTS OF HIGHER ORDER
Directory of Open Access Journals (Sweden)
Khomchenko A.
2017-12-01
Full Text Available The paper deals with the problem of constructing the basic functions of a quadrilateral finite element of the fifth order by the means of the computer algebra system Maple. The Lagrangian approximation of such a finite element contains 36 nodes: 20 nodes perimeter and 16 internal nodes. Alternative models with reduced number of internal nodes are considered. Graphs of basic functions and cognitive portraits of lines of zero level are presented. The work is aimed at studying the possibilities of using modern information technologies in the teaching of individual mathematical disciplines.
MMA, A Computer Code for Multi-Model Analysis
Poeter, Eileen P.; Hill, Mary C.
2007-01-01
This report documents the Multi-Model Analysis (MMA) computer code. MMA can be used to evaluate results from alternative models of a single system using the same set of observations for all models. As long as the observations, the observation weighting, and system being represented are the same, the models can differ in nearly any way imaginable. For example, they may include different processes, different simulation software, different temporal definitions (for example, steady-state and transient models could be considered), and so on. The multiple models need to be calibrated by nonlinear regression. Calibration of the individual models needs to be completed before application of MMA. MMA can be used to rank models and calculate posterior model probabilities. These can be used to (1) determine the relative importance of the characteristics embodied in the alternative models, (2) calculate model-averaged parameter estimates and predictions, and (3) quantify the uncertainty of parameter estimates and predictions in a way that integrates the variations represented by the alternative models. There is a lack of consensus on what model analysis methods are best, so MMA provides four default methods. Two are based on Kullback-Leibler information, and use the AIC (Akaike Information Criterion) or AICc (second-order-bias-corrected AIC) model discrimination criteria. The other two default methods are the BIC (Bayesian Information Criterion) and the KIC (Kashyap Information Criterion) model discrimination criteria. Use of the KIC criterion is equivalent to using the maximum-likelihood Bayesian model averaging (MLBMA) method. AIC, AICc, and BIC can be derived from Frequentist or Bayesian arguments. The default methods based on Kullback-Leibler information have a number of theoretical advantages, including that they tend to favor more complicated models as more data become available than do the other methods, which makes sense in many situations. Many applications of MMA will
Determination of Major and Minor Elements in the Code River Sediments
International Nuclear Information System (INIS)
Sri Murniasih; Sukirno; Bambang Irianto
2007-01-01
Analyze major and minor elements in the Code river sediments has been done. The aim of this research is to determine the concentration of major and minor elements in the Code river sediments from upstream to downstream. The instrument used were X-ray Fluorescence using Si(Li) detector. The results show that major elements were Fe (1.66 ± 0.1% - 4.20 ± 0.7%) and Ca (4.43 ± 0.6% - 9.08 ± 1.3%); while minor elements were Ba (178.791 ± 21.1 ppm - 616.56 ± 59.4 ppm); Sr (148.22 ± 21.9 ppm - 410.25 ± 30.5 ppm); and Zr (9.71 ± 1.1 ppm - 22.11 ± 3.4 ppm). ANAVA method (confidence level of α 0.05 ) for statistic test was used. It was showed that there were significant influence of the sampling location difference on the concentration of major and minor elements in the sediment samples. (author)
ABAQUS/EPGEN - a general purpose finite element code with emphasis on nonlinear applications
International Nuclear Information System (INIS)
Hibbitt, H.D.
1984-01-01
The article contains a summary description of ABAQUS, a finite element program designed for general use in nonlinear as well as linear structural problems, in the context of its application to nuclear structural integrity analysis. The article begins with a discussion of the design criteria and methods upon which the code development has been based. The engineering modelling capabilities, currently implemented in the program - elements, constitutive models and analysis procedures - are then described. Finally, a few demonstration examples are presented, to illustrate some of the program's features that are of interest in structural integrity analysis associated with nuclear power plants. (orig.)
DYNAPCON: a computer code for dynamic analysis of prestressed concrete structures
International Nuclear Information System (INIS)
Marchertas, A.H.
1982-09-01
A finite element computer code for the transient analysis of prestressed concrete reactor vessels (PCRVs) for LMFBR containment is described. The method assumes rotational symmetry of the structure. Time integration is by an explicit method. The quasistatic prestressing operation of the PCRV model is performed by a dynamic relaxation technique. The material model accounts for the crushing and tensile cracking in arbitrary direction in concrete and the elastic-plastic behavior of reinforcing steel. The variation of the concrete tensile cracking and compressive crushing limits with strain rate is taken into account. Relative slip is permitted between the concrete and tendons. Several example solutions are presented and compared with experimental results. These sample problems range from simply supported beams to small scale models of PCRV's. It is shown that the analytical methods correlate quite well with experimental results, although in the vicinity of the failure load the response of the models tend to be quite sensitive to input parameters
DANIO-CODE: Toward an Encyclopedia of DNA Elements in Zebrafish.
Tan, Haihan; Onichtchouk, Daria; Winata, Cecilia
2016-02-01
The zebrafish has emerged as a model organism for genomics studies. The symposium "Toward an encyclopedia of DNA elements in zebrafish" held in London in December 2014, was coorganized by Ferenc Müller and Fiona Wardle. This meeting is a follow-up of a similar previous workshop held 2 years earlier and represents a push toward the formalization of a community effort to annotate functional elements in the zebrafish genome. The meeting brought together zebrafish researchers, bioinformaticians, as well as members of established consortia, to exchange scientific findings and experience, as well as to discuss the initial steps toward the formation of a DANIO-CODE consortium. In this study, we provide the latest updates on the current progress of the consortium's efforts, opening up a broad invitation to researchers to join in and contribute to DANIO-CODE.
K. V. Pagalthivarthi; R. J. Visintainer
2013-01-01
Multi-size particulate dense slurry flow through three-dimensional rectangular channel is modeled using penalty finite elements with 8-noded hexahedral elements. The methodology previously developed for two-dimensional channel is extended. The computed eddy viscosity of the pure carrier flow is modified to account for the presence of solid particles. Predictions from Spalart-Almaras and k-ε turbulence models are compared to show consistency of trends in results. Results are also found to comp...
Fuel behavior modeling using the MARS computer code
International Nuclear Information System (INIS)
Faya, S.C.S.; Faya, A.J.G.
1983-01-01
The fuel behaviour modeling code MARS against experimental data, was evaluated. Two cases were selected: an early comercial PWR rod (Maine Yankee rod) and an experimental rod from the Canadian BWR program (Canadian rod). The MARS predictions are compared with experimental data and predictions made by other fuel modeling codes. Improvements are suggested for some fuel behaviour models. Mars results are satisfactory based on the data available. (Author) [pt
The spectral-element method, Beowulf computing, and global seismology.
Komatitsch, Dimitri; Ritsema, Jeroen; Tromp, Jeroen
2002-11-29
The propagation of seismic waves through Earth can now be modeled accurately with the recently developed spectral-element method. This method takes into account heterogeneity in Earth models, such as three-dimensional variations of seismic wave velocity, density, and crustal thickness. The method is implemented on relatively inexpensive clusters of personal computers, so-called Beowulf machines. This combination of hardware and software enables us to simulate broadband seismograms without intrinsic restrictions on the level of heterogeneity or the frequency content.
Finite element computation of natural convection in enclosures
International Nuclear Information System (INIS)
Kushwaha, H.S.
1982-01-01
Compared to U-V-P-T formulation and stream-vorticity temperature formulation, penalty function formulation is simple and computationally competitive. Incremental New-Raphons method employed in this study is effective and efficient. From this study it is established that very fine mesh is not required for a low Rayleigh number considered in this study. The upwinding finite element may be necessary to avoid oscillations for higher Rayleigh numbers. (author)
Photodeposited diffractive optical elements of computer generated masks
International Nuclear Information System (INIS)
Mirchin, N.; Peled, A.; Baal-Zedaka, I.; Margolin, R.; Zagon, M.; Lapsker, I.; Verdyan, A.; Azoulay, J.
2005-01-01
Diffractive optical elements (DOE) were synthesized on plastic substrates using the photodeposition (PD) technique by depositing amorphous selenium (a-Se) films with argon lasers and UV spectra light. The thin films were deposited typically onto polymethylmethacrylate (PMMA) substrates at room temperature. Scanned beam and contact mask modes were employed using computer-designed DOE lenses. Optical and electron micrographs characterize the surface details. The films were typically 200 nm thick
Photodeposited diffractive optical elements of computer generated masks
Energy Technology Data Exchange (ETDEWEB)
Mirchin, N. [Electrical and Electronics Engineering Department, Holon Academic Institute of Technology, 52 Golomb Street, Holon 58102 (Israel)]. E-mail: mirchin@hait.ac.il; Peled, A. [Electrical and Electronics Engineering Department, Holon Academic Institute of Technology, 52 Golomb Street, Holon 58102 (Israel); Baal-Zedaka, I. [Electrical and Electronics Engineering Department, Holon Academic Institute of Technology, 52 Golomb Street, Holon 58102 (Israel); Margolin, R. [Electrical and Electronics Engineering Department, Holon Academic Institute of Technology, 52 Golomb Street, Holon 58102 (Israel); Zagon, M. [Electrical and Electronics Engineering Department, Holon Academic Institute of Technology, 52 Golomb Street, Holon 58102 (Israel); Lapsker, I. [Physics Department, Holon Academic Institute of Technology, 52 Golomb Street, Holon 58102 (Israel); Verdyan, A. [Physics Department, Holon Academic Institute of Technology, 52 Golomb Street, Holon 58102 (Israel); Azoulay, J. [Physics Department, Holon Academic Institute of Technology, 52 Golomb Street, Holon 58102 (Israel)
2005-07-30
Diffractive optical elements (DOE) were synthesized on plastic substrates using the photodeposition (PD) technique by depositing amorphous selenium (a-Se) films with argon lasers and UV spectra light. The thin films were deposited typically onto polymethylmethacrylate (PMMA) substrates at room temperature. Scanned beam and contact mask modes were employed using computer-designed DOE lenses. Optical and electron micrographs characterize the surface details. The films were typically 200 nm thick.
Energy Technology Data Exchange (ETDEWEB)
Candel, A.; Kabel, A.; Lee, L.; Li, Z.; Limborg, C.; Ng, C.; Prudencio, E.; Schussman, G.; Uplenchwar, R.; Ko, K.; /SLAC
2009-06-19
Over the past years, SLAC's Advanced Computations Department (ACD), under SciDAC sponsorship, has developed a suite of 3D (2D) parallel higher-order finite element (FE) codes, T3P (T2P) and Pic3P (Pic2P), aimed at accurate, large-scale simulation of wakefields and particle-field interactions in radio-frequency (RF) cavities of complex shape. The codes are built on the FE infrastructure that supports SLAC's frequency domain codes, Omega3P and S3P, to utilize conformal tetrahedral (triangular)meshes, higher-order basis functions and quadratic geometry approximation. For time integration, they adopt an unconditionally stable implicit scheme. Pic3P (Pic2P) extends T3P (T2P) to treat charged-particle dynamics self-consistently using the PIC (particle-in-cell) approach, the first such implementation on a conformal, unstructured grid using Whitney basis functions. Examples from applications to the International Linear Collider (ILC), Positron Electron Project-II (PEP-II), Linac Coherent Light Source (LCLS) and other accelerators will be presented to compare the accuracy and computational efficiency of these codes versus their counterparts using structured grids.
Multi keno-VAX a modified version of the reactor computer code Multi keno-2
International Nuclear Information System (INIS)
Imam, M.
1995-01-01
The reactor computer code Multi keno-2 is developed in Japan from the original Monte Carlo Keno-IV. By applications of this code on some real problems, fatal errors were detected. These errors are related to the restart option in the code. The restart option is essential for solving time-consuming problems on mini-computer like VAX-6320. These errors were corrected and other modifications were carried out in the code. Because of these modifications new input data description was written for the code. Thus a new VAX/VMS version for the program was developed which is also adaptable for mini-mainframes. This new developed program, called Multi keno-VAX is accepted in the Nea-IAEA data bank and is added to its international computer codes library. 1 fig
Application of computational fluid dynamics methods to improve thermal hydraulic code analysis
Sentell, Dennis Shannon, Jr.
A computational fluid dynamics code is used to model the primary natural circulation loop of a proposed small modular reactor for comparison to experimental data and best-estimate thermal-hydraulic code results. Recent advances in computational fluid dynamics code modeling capabilities make them attractive alternatives to the current conservative approach of coupled best-estimate thermal hydraulic codes and uncertainty evaluations. The results from a computational fluid dynamics analysis are benchmarked against the experimental test results of a 1:3 length, 1:254 volume, full pressure and full temperature scale small modular reactor during steady-state power operations and during a depressurization transient. A comparative evaluation of the experimental data, the thermal hydraulic code results and the computational fluid dynamics code results provides an opportunity to validate the best-estimate thermal hydraulic code's treatment of a natural circulation loop and provide insights into expanded use of the computational fluid dynamics code in future designs and operations. Additionally, a sensitivity analysis is conducted to determine those physical phenomena most impactful on operations of the proposed reactor's natural circulation loop. The combination of the comparative evaluation and sensitivity analysis provides the resources for increased confidence in model developments for natural circulation loops and provides for reliability improvements of the thermal hydraulic code.
International Nuclear Information System (INIS)
VOOGD, J.A.
1999-01-01
An analysis of three software proposals is performed to recommend a computer code for immobilized low activity waste flow and transport modeling. The document uses criteria restablished in HNF-1839, ''Computer Code Selection Criteria for Flow and Transport Codes to be Used in Undisturbed Vadose Zone Calculation for TWRS Environmental Analyses'' as the basis for this analysis
Computer codes for simulating atomic-displacement cascades in solids subject to irradiation
International Nuclear Information System (INIS)
Asaoka, Takumi; Taji, Yukichi; Tsutsui, Tsuneo; Nakagawa, Masayuki; Nishida, Takahiko
1979-03-01
In order to study atomic displacement cascades originating from primary knock-on atoms in solids subject to incident radiation, the simulation code CASCADE/CLUSTER is adapted for use on FACOM/230-75 computer system. In addition, the code is modified so as to plot the defect patterns in crystalline solids. As other simulation code of the cascade process, MARLOWE is also available for use on the FACOM system. To deal with the thermal annealing of point defects produced in the cascade process, the code DAIQUIRI developed originally for body-centered cubic crystals is modified to be applicable also for face-centered cubic lattices. By combining CASCADE/CLUSTER and DAIQUIRI, we then prepared a computer code system CASCSRB to deal with heavy irradiation or saturation damage state of solids at normal temperature. Furthermore, a code system for the simulation of heavy irradiations CASCMARL is available, in which MARLOWE code is substituted for CASCADE in the CASCSRB system. (author)
Byun, Chansup; Guruswamy, Guru P.; Kutler, Paul (Technical Monitor)
1994-01-01
In recent years significant advances have been made for parallel computers in both hardware and software. Now parallel computers have become viable tools in computational mechanics. Many application codes developed on conventional computers have been modified to benefit from parallel computers. Significant speedups in some areas have been achieved by parallel computations. For single-discipline use of both fluid dynamics and structural dynamics, computations have been made on wing-body configurations using parallel computers. However, only a limited amount of work has been completed in combining these two disciplines for multidisciplinary applications. The prime reason is the increased level of complication associated with a multidisciplinary approach. In this work, procedures to compute aeroelasticity on parallel computers using direct coupling of fluid and structural equations will be investigated for wing-body configurations. The parallel computer selected for computations is an Intel iPSC/860 computer which is a distributed-memory, multiple-instruction, multiple data (MIMD) computer with 128 processors. In this study, the computational efficiency issues of parallel integration of both fluid and structural equations will be investigated in detail. The fluid and structural domains will be modeled using finite-difference and finite-element approaches, respectively. Results from the parallel computer will be compared with those from the conventional computers using a single processor. This study will provide an efficient computational tool for the aeroelastic analysis of wing-body structures on MIMD type parallel computers.
V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009
Energy Technology Data Exchange (ETDEWEB)
Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.
2010-07-15
V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)
Gupta, K. K.
1997-01-01
A multidisciplinary, finite element-based, highly graphics-oriented, linear and nonlinear analysis capability that includes such disciplines as structures, heat transfer, linear aerodynamics, computational fluid dynamics, and controls engineering has been achieved by integrating several new modules in the original STARS (STructural Analysis RoutineS) computer program. Each individual analysis module is general-purpose in nature and is effectively integrated to yield aeroelastic and aeroservoelastic solutions of complex engineering problems. Examples of advanced NASA Dryden Flight Research Center projects analyzed by the code in recent years include the X-29A, F-18 High Alpha Research Vehicle/Thrust Vectoring Control System, B-52/Pegasus Generic Hypersonics, National AeroSpace Plane (NASP), SR-71/Hypersonic Launch Vehicle, and High Speed Civil Transport (HSCT) projects. Extensive graphics capabilities exist for convenient model development and postprocessing of analysis results. The program is written in modular form in standard FORTRAN language to run on a variety of computers, such as the IBM RISC/6000, SGI, DEC, Cray, and personal computer; associated graphics codes use OpenGL and IBM/graPHIGS language for color depiction. This program is available from COSMIC, the NASA agency for distribution of computer programs.
Qualification of the new version of HAMMER computer code
International Nuclear Information System (INIS)
Chia, C.T.
1984-06-01
(HTEC) code were tested with a great number of diferent type of experiments. This experiments covers the most important parameters in neutronic calculations, such as the cell geometry and composition. The HTEC code results have been analysed and compared with experimental data and results given by the literature and simulated by HAMMER and LEOPARD codes. The quantities used for analysis were Keff and the following integral parameters: R28 - ratio of epicadmium-to-subcadmium 238 U captures; D25 - ratio of epicadmium-to-subcadmium 235 U fission; D28 - ratio of 238 U fissions to 235 U fissions; C - ratio of 238 U captures to 235 U fissions; RC02 - ratio of epicadmium-to-subcadmium 232 Th capture. The analysis shows that the results given by the code are in good agreement with the experimental data and the results given by the other codes. The calculation that have been done with the detailed ressonance profile tabulations of plutonium isotopes shows worst results than that obtained with the ressonance parameters. Almost all the simulated cases, shows that the HTEC results are closest to the experimental data than the HAMMER results, when one do not use the detailed ressonance profile tabulations of the plutonium isotopes. (Author) [pt
Basic data, computer codes and integral experiments: The tools for modelling in nuclear technology
International Nuclear Information System (INIS)
Sartori, E.
2001-01-01
When studying applications in nuclear technology we need to understand and be able to predict the behavior of systems manufactured by human enterprise. First, the underlying basic physical and chemical phenomena need to be understood. We have then to predict the results from the interplay of the large number of the different basic events: i.e. the macroscopic effects. In order to be able to build confidence in our modelling capability, we need then to compare these results against measurements carried out on such systems. The different levels of modelling require the solution of different types of equations using different type of parameters. The tools required for carrying out a complete validated analysis are: - The basic nuclear or chemical data; - The computer codes, and; - The integral experiments. This article describes the role each component plays in a computational scheme designed for modelling purposes. It describes also which tools have been developed and are internationally available. The role of the OECD/NEA Data Bank, the Radiation Shielding Information Computational Center (RSICC), and the IAEA Nuclear Data Section are playing in making these elements available to the community of scientists and engineers is described. (author)
Algorithms and data structures for massively parallel generic adaptive finite element codes
Bangerth, Wolfgang
2011-12-01
Today\\'s largest supercomputers have 100,000s of processor cores and offer the potential to solve partial differential equations discretized by billions of unknowns. However, the complexity of scaling to such large machines and problem sizes has so far prevented the emergence of generic software libraries that support such computations, although these would lower the threshold of entry and enable many more applications to benefit from large-scale computing. We are concerned with providing this functionality for mesh-adaptive finite element computations. We assume the existence of an "oracle" that implements the generation and modification of an adaptive mesh distributed across many processors, and that responds to queries about its structure. Based on querying the oracle, we develop scalable algorithms and data structures for generic finite element methods. Specifically, we consider the parallel distribution of mesh data, global enumeration of degrees of freedom, constraints, and postprocessing. Our algorithms remove the bottlenecks that typically limit large-scale adaptive finite element analyses. We demonstrate scalability of complete finite element workflows on up to 16,384 processors. An implementation of the proposed algorithms, based on the open source software p4est as mesh oracle, is provided under an open source license through the widely used deal.II finite element software library. © 2011 ACM 0098-3500/2011/12-ART10 $10.00.
Quantum computation with topological codes from qubit to topological fault-tolerance
Fujii, Keisuke
2015-01-01
This book presents a self-consistent review of quantum computation with topological quantum codes. The book covers everything required to understand topological fault-tolerant quantum computation, ranging from the definition of the surface code to topological quantum error correction and topological fault-tolerant operations. The underlying basic concepts and powerful tools, such as universal quantum computation, quantum algorithms, stabilizer formalism, and measurement-based quantum computation, are also introduced in a self-consistent way. The interdisciplinary fields between quantum information and other fields of physics such as condensed matter physics and statistical physics are also explored in terms of the topological quantum codes. This book thus provides the first comprehensive description of the whole picture of topological quantum codes and quantum computation with them.
Fowler, S. J.; Driesner, T.; Kulik, D.; Wagner, T.
2010-12-01
We present a novel computational tool for modelling temporally and spatially varying chemical interactions between hydrothermal fluids and rocks that may affect the long-term performance of geothermal reservoirs. The code is written in C++. It incorporates fluid-rock interaction and scale formation self-consistently, via a modular coupling approach that combines the Complex System Modelling Platform (CSMP++) code for fluid flow in porous and fractured media (Matthai et al., 2007) with the numerical kernel (GEMIPM2K) of the GEM-Selektor Gibbs free energy minimization package (Kulik, Wagner et al., 2007). CSMP++ uses finite element-finite volume spatial discretization, implicit or explicit time discretization, and an operator splitting approach to solve equations. The GEM-Selektor package supports a wide range of equation of state and activity models, facilitating calculation of complex fluid-mineral equilibria. Coupled code input includes temperature, pressure, a charge balance, and total amounts of system chemical elements, as well as domain and boundary condition specifications. Speciation, thermodynamic, and physical properties of the system are output. Critical advantages of the coupled code compared to existing hydrothermal reactive transport models are: (1) simultaneous consideration of complex solid solutions (e.g., clay minerals) and non-ideal aqueous solutions (GEMIPM2K), and (2) a discretization scheme that can be applied to mass and heat transport in irregular, geologically realistic geometries (CSMP++). Each coupled simulation results in a thermodynamically-based description of the geochemical and physical state of a hydrothermal system evolving along a complex P-T-X path. The code design allows for efficient and flexible incorporation of numerical and thermodynamic database improvements. We apply the coupled code to a number of geologic applications to test its accuracy and performance. Kulik, D., Wagner, T. et al. (2007). GEM-Selektor (GEMS-PSI) home
Validation of thermohydraulic codes by comparison of experimental results with computer simulations
International Nuclear Information System (INIS)
Madeira, A.A.; Galetti, M.R.S.; Pontedeiro, A.C.
1989-01-01
The results obtained by simulation of three cases from CANON depressurization experience, using the TRAC-PF1 computer code, version 7.6, implanted in the VAX-11/750 computer of Brazilian CNEN, are presented. The CANON experience was chosen as first standard problem in thermo-hydraulic to be discussed at ENFIR for comparing results from different computer codes with results obtained experimentally. The ability of TRAC-PF1 code to prevent the depressurization phase of a loss of primary collant accident in pressurized water reactors is evaluated. (M.C.K.) [pt
Development of the computer code system for the analyses of PWR core
International Nuclear Information System (INIS)
Tsujimoto, Iwao; Naito, Yoshitaka.
1992-11-01
This report is one of the materials for the work titled 'Development of the computer code system for the analyses of PWR core phenomena', which is performed under contracts between Shikoku Electric Power Company and JAERI. In this report, the numerical method adopted in our computer code system are described, that is, 'The basic course and the summary of the analysing method', 'Numerical method for solving the Boltzmann equation', 'Numerical method for solving the thermo-hydraulic equations' and 'Description on the computer code system'. (author)
International Nuclear Information System (INIS)
Goncalves Filho, O.J.A.
1987-01-01
This work aims to describe the computer code EVP2D developed for the elastoviscoplastic-damage analysis of mettalic components, with particular emphasis dedicated to the problem of creep damage and rupture. After a brief introduction of the basic concepts and procedures of Continuum Damage Mechanics, the constitutive equations implemented are presented. Next, the finite element approximation proposed for solution of the initial boundary value problem of interest is discussed, particularly the numerical algorithms used for time integration of the creep strain rate and damage rate equations, and the numerical procedures adopted for dealing with the presense of partially or fully ruptured finite elements in the mesh. As a pratical application, the rupture behaviour of a biaxially tension loaded plate containing a central circular hole is examined. Finally, future developments of the code, which include as prioritiesthe treatment of ciyclic loads and the description of the anisotropic feature of creep damage evolution, are briefly introduced. (author) [pt
Code of Ethical Conduct for Computer-Using Educators: An ICCE Policy Statement.
Computing Teacher, 1987
1987-01-01
Prepared by the International Council for Computers in Education's Ethics and Equity Committee, this code of ethics for educators using computers covers nine main areas: curriculum issues, issues relating to computer access, privacy/confidentiality issues, teacher-related issues, student issues, the community, school organizational issues,…
Low Computational Complexity Network Coding For Mobile Networks
DEFF Research Database (Denmark)
Heide, Janus
2012-01-01
cooperation among receivers. In meshed networks, to simplify routing schemes and to increase robustness toward node failures. This thesis deals with implementation issues of one NC technique namely Random Linear Network Coding (RLNC) which can be described as a highly decentralized non-deterministic intra...
Development of a Computer Code for the Estimation of Fuel Rod Failure
Energy Technology Data Exchange (ETDEWEB)
Rhee, I.H.; Ahn, H.J. [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)
1997-12-31
Much research has already been performed to obtain the information on the degree of failed fuel rods from the primary coolant activities of operating PWRs in the last few decades. The computer codes that are currently in use for domestic nuclear power plants, such as CADE code and ABB-CE codes developed by Westinghouse and ABB-CE, respectively, still give significant overall errors in estimating the failed fuel rods. In addition, with the CADE code, it is difficult to predict the degree of fuel rod failures during the transient period of nuclear reactor operation, where as the ABB-CE codes are relatively more difficult to use for end-users. In particular, the rapid progresses made recently in the area of the computer hardware and software systems that their computer programs be more versatile and user-friendly. While the MS windows system that is centered on the graphic user interface and multitasking is now in widespread use, the computer codes currently employed at the nuclear power plants, such as CADE and ABB-CE codes, can only be run on the DOS system. Moreover, it is desirable to have a computer code for the fuel rod failure estimation that can directly use the radioactivity data obtained from the on-line monitoring system of the primary coolant activity. The main purpose of this study is, therefore, to develop a Windows computer code that can predict the location, the number of failed fuel rods,and the degree of failures using the radioactivity data obtained from the primary coolant activity for PWRs. Another objective is to combine this computer code with the on-line monitoring system of the primary coolant radioactivity at Kori 3 and 4 operating nuclear power plants and enable their combined use for on-line evaluation of the number and degree of fuel rod failures. (author). 49 refs., 85 figs., 30 tabs.
KOELSCH, a computer code simulating mechanical HCDA consequences in LMFBR primary containment
International Nuclear Information System (INIS)
Lange, L.; Gerling, W.; Schlein, I.
1981-01-01
The computer code KOELSCH (Kontinuumsmechanik in Euler-Lagrange Darstellung und Schalentheorie) calculates problems of continuum mechanics in Euler-Lagrangian representation and with the use of shell theory. It is being developed for the advanced simulation of the mechanical loading and the response of the primary containment of a LMFBR in the case of an HCDA. The essential features of the code and examples of calculations are presented. For the hydrodynamics the theoretical basis is the 3D Euler-Lagrangian formulation of the hydrodynamic equations in general curved coordinates. This starting point assure optimum adaptation of the physical problem to the geometry to be considered. The options available in KOELSCH are - H1: quasi- 1D Euler-Lagrangian - H2: 2D Euler - H3: 3D Euler (only rigid boundaries possible). The structural dynamics is treated by 2D Lagrangian formalism in cylindrical coordinates. Two options are foreseen: - thin shells: membrane or shell theory - massive structures: full tensor properties of materials are included. Flexibility with respect to complex geometries is gained by surface integration of the governing equations over finite elements with 3 - 8 corners. (orig./GL)
A three-dimensional computer code for the nonlinear dynamic response of an HTGR core
International Nuclear Information System (INIS)
Subudhi, M.; Lasker, L.; Koplik, B.; Curreri, J.; Goradia, H.
1979-01-01
A three-dimensional dynamic code has been developed to determine the nonlinear response of an HTGR core. The HTGR core consists of several thousands of hexagonal core blocks. These are arranged in layers stacked together. Each layer contains many core blocks surrounded on their outer periphery by reflector blocks. The entire assembly is contained within a prestressed concrete reactor vessel. Gaps exist between adjacent blocks in any horizontal plane. Each core block in a given layer is connected to the blocks directly above and below it via three dowell pins. The present analytical study is directed towards an investigation of the nonlinear response of the reactor core blocks in the event of a seismic occurrence. The computer code is developed for a specific mathematical model which represents a vertical arrangement of layers of blocks. This comprises a 'block module' of core elements which would be obtained by cutting a cylindrical portion consisting of seven fuel blocks per layer. It is anticipated that a number of such modules properly arranged could represent the entire core. Hence, the predicted response of this module would exhibit the response characteristics of the core. (orig.)
International Nuclear Information System (INIS)
Sheron, B.W.; Rosztoczy, Z.R.
1980-08-01
As a result of a request from Commissioner V. Gilinsky to investigate in detail the causes of an error discovered in a vendor Emergency Core Cooling System (ECCS) computer code in March, 1978, the staff undertook an extensive investigation of the vendor quality assurance practices applied to safety analysis computer code development and use. This investigation included inspections of code development and use practices of the four major Light Water Reactor Nuclear Steam Supply System vendors and a major reload fuel supplier. The conclusion reached by the staff as a result of the investigation is that vendor practices for code development and use are basically sound. A number of areas were identified, however, where improvements to existing vendor procedures should be made. In addition, the investigation also addressed the quality assurance (QA) review and inspection process for computer codes and identified areas for improvement
Development and validation of GWHEAD, a three-dimensional groundwater head computer code
International Nuclear Information System (INIS)
Beckmeyer, R.R.; Root, R.W.; Routt, K.R.
1980-03-01
A computer code has been developed to solve the groundwater flow equation in three dimensions. The code has finite-difference approximations solved by the strongly implicit solution procedure. Input parameters to the code include hydraulic conductivity, specific storage, porosity, accretion (recharge), and initial hydralic head. These parameters may be input as varying spatially. The hydraulic conductivity may be input as isotropic or anisotropic. The boundaries either may permit flow across them or may be impermeable. The code has been used to model leaky confined groundwater conditions and spherical flow to a continuous point sink, both of which have exact analytical solutions. The results generated by the computer code compare well with those of the analytical solutions. The code was designed to be used to model groundwater flow beneath fuel reprocessing and waste storage areas at the Savannah River Plant
Current status of the transient integral fuel element performance code URANUS
International Nuclear Information System (INIS)
Preusser, T.; Lassmann, K.
1983-01-01
To investigate the behavior of fuel pins during normal and off-normal operation, the integral fuel rod code URANUS has been extended to include a transient version. The paper describes the current status of the program system including a presentation of newly developed models for hypothetical accident investigation. The main objective of current development work is to improve the modelling of fuel and clad material behavior during fast transients. URANUS allows detailed analysis of experiments until the onset of strong material transport phenomena. Transient fission gas analysis is carried out due to the coupling with a special version of the LANGZEIT-KURZZEIT-code (KfK). Fuel restructuring and grain growth kinetics models have been improved recently to better characterize pre-experimental steady-state operation; transient models are under development. Extensive verification of the new version has been carried out by comparison with analytical solutions, experimental evidence, and code-to-code evaluation studies. URANUS, with all these improvements, has been successfully applied to difficult fast breeder fuel rod analysis including TOP, LOF, TUCOP, local coolant blockage and specific carbide fuel experiments. Objective of further studies is the description of transient PCMI. It is expected that the results of these developments will contribute significantly to the understanding of fuel element structural behavior during severe transients. (orig.)
Dimitrieva, Slavica; Bucher, Philipp
2012-09-15
Genomic context analysis, also known as phylogenetic profiling, is widely used to infer functional interactions between proteins but rarely applied to non-coding cis-regulatory DNA elements. We were wondering whether this approach could provide insights about utlraconserved non-coding elements (UCNEs). These elements are organized as large clusters, so-called gene regulatory blocks (GRBs) around key developmental genes. Their molecular functions and the reasons for their high degree of conservation remain enigmatic. In a special setting of genomic context analysis, we analyzed the fate of GRBs after a whole-genome duplication event in five fish genomes. We found that in most cases all UCNEs were retained together as a single block, whereas the corresponding target genes were often retained in two copies, one completely devoid of UCNEs. This 'winner-takes-all' pattern suggests that UCNEs of a GRB function in a highly cooperative manner. We propose that the multitude of interactions between UCNEs is the reason for their extreme sequence conservation. Supplementary data are available at Bioinformatics online and at http://ccg.vital-it.ch/ucne/
Verification of Advective Bar Elements Implemented in the Aria Thermal Response Code.
Energy Technology Data Exchange (ETDEWEB)
Mills, Brantley [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
2016-01-01
A verification effort was undertaken to evaluate the implementation of the new advective bar capability in the Aria thermal response code. Several approaches to the verification process were taken : a mesh refinement study to demonstrate solution convergence in the fluid and the solid, visually examining the mapping of the advective bar element nodes to the surrounding surfaces, and a comparison of solutions produced using the advective bars for simple geometries with solutions from commercial CFD software . The mesh refinement study has shown solution convergence for simple pipe flow in both temperature and velocity . Guidelines were provided to achieve appropriate meshes between the advective bar elements and the surrounding volume. Simulations of pipe flow using advective bars elements in Aria have been compared to simulations using the commercial CFD software ANSYS Fluent (r) and provided comparable solutions in temperature and velocity supporting proper implementation of the new capability. Verification of Advective Bar Elements iv Acknowledgements A special thanks goes to Dean Dobranich for his guidance and expertise through all stages of this effort . His advice and feedback was instrumental to its completion. Thanks also goes to Sam Subia and Tolu Okusanya for helping to plan many of the verification activities performed in this document. Thank you to Sam, Justin Lamb and Victor Brunini for their assistance in resolving issues encountered with running the advective bar element model. Finally, thanks goes to Dean, Sam, and Adam Hetzler for reviewing the document and providing very valuable comments.
GATE: computation code for medical imagery, radiotherapy and dosimetry
International Nuclear Information System (INIS)
Jan, S.
2010-01-01
The author presents the GATE code, a simulation software based on the Geant4 development environment developed by the CERN (the European organization for nuclear research) which enables Monte-Carlo type simulation to be developed for tomography imagery using ionizing radiation, and radiotherapy examinations (conventional and hadron therapy) to be simulated. The authors concentrate on the use of medical imagery in carcinology. They comment some results obtained in nuclear imagery and in radiotherapy
Energy Technology Data Exchange (ETDEWEB)
Gould, R K; Srivastava, R
1979-12-01
Models and computer codes which may be used to describe flow reactors in which high purity, solar grade silicon is produced via reduction of gaseous silicon halides are described. A prominent example of the type of process which may be studied using the codes developed in this program is the SiCl/sub 4//Na reactor currently being developed by the Westinghouse Electric Corp. During this program two large computer codes were developed. The first is the CHEMPART code, an axisymmetric, marching code which treats two-phase flows with models describing detailed gas-phase chemical kinetics, particle formation, and particle growth. This code, based on the AeroChem LAPP (Low Altitude Plume Program) code can be used to describe flow reactors in which reactants mix, react, and form a particulate phase. Detailed radial gas-phase composition, temperature, velocity, and particle size distribution profiles are computed. Also, depositon of heat, momentum, and mass (either particulate or vapor) on reactor walls is described. The second code is a modified version of the GENMIX boundary layer code which is used to compute rates of heat, momentum, and mass transfer to the reactor walls. This code lacks the detailed chemical kinetics and particle handling features of the CHEMPART code but has the virtue of running much more rapidly than CHEMPART, while treating the phenomena occurring in the boundary layer in more detail than can be afforded using CHEMPART. These two codes have been used in this program to predict particle formation characteristics and wall collection efficiencies for SiCl/sub 4//Na flow reactors. Results are described.
Development of a computer code for Dalat research reactor transient analysis
International Nuclear Information System (INIS)
Le Vinh Vinh; Nguyen Thai Sinh; Huynh Ton Nghiem; Luong Ba Vien; Pham Van Lam; Nguyen Kien Cuong
2003-01-01
DRSIM (Dalat Reactor SIMulation) computer code has been developed for Dalat reactor transient analysis. It is basically a coupled neutronics-hydrodynamics-heat transfer code employing point kinetics, one dimensional hydrodynamics and one dimensional heat transfer. The work was financed by VAEC and DNRI in the framework of institutional R and D programme. Some transient problems related to reactivity and loss of coolant flow was carried out by DRSIM using temperature and void coefficients calculated by WIMS and HEXNOD2D codes. (author)
Code and papers: computing publication patterns in the LHC era
CERN. Geneva
2012-01-01
Publications in scholarly journals establish the body of knowledge deriving from scientific research; they also play a fundamental role in the career path of scientists and in the evaluation criteria of funding agencies. This presentation reviews the evolution of computing-oriented publications in HEP following the start of operation of LHC. Quantitative analyses are illustrated, which document the production of scholarly papers on computing-related topics by HEP experiments and core tools projects (including distributed computing R&D), and the citations they receive. Several scientometric indicators are analyzed to characterize the role of computing in HEP literature. Distinctive features of scholarly publication production in the software-oriented and hardware-oriented experimental HEP communities are highlighted. Current patterns and trends are compared to the situation in previous generations' HEP experiments at LEP, Tevatron and B-factories. The results of this scientometric analysis document objec...
Computer code for thermal-hydraulic simulation of heat pressurizer tanks operation (Simterm-H)
International Nuclear Information System (INIS)
Sellos, R.F.
1987-01-01
It is presented the Simtherm-H computer code, developed for calculating the thermodynamic properties of the high pressure heating system and the feedwater tank in transient state for PWR nuclear power plants (1300 MWe). (E.G.) [pt
FRAP-T1: a computer code for the transient analysis of oxide fuel rods
International Nuclear Information System (INIS)
Dearien, J.A.; Miller, R.L.; Hobbins, R.R.; Siefken, L.J.; Baston, V.F.; Coleman, D.R.
1977-02-01
FRAP-T is a FORTRAN IV computer code which can be used to solve for the transient response of a light water reactor (LWR) fuel rod during accident transients such as loss-of-coolant accident (LOCA) or a power-cooling-mismatch (PCM). The coupled effects of mechanical, thermal, internal gas, and material property response on the behavior of the fuel rod are considered. FRAP-T is a modular code with each major computational model isolated within the code and coupled to the main code by subroutine calls and data transfer through argument lists. FRAP-T is coupled to a materials properties subcode (MATPRO) which is used to provide gas, fuel, and cladding properties to the FRAP-T computational subcodes. No material properties need be supplied by the code user. The needed water properties are stored in tables built into the code. Critical heat flux (CHF) and heat transfer correlations for a wide range of coolant conditions are contained in modular subroutines. FRAP-T has been evaluated by making extensive comparisons between predictions of the code and experimental data. Comparison of predicted and experimental results are presented for a range of FRAP-T calculated parameters. The code is presently programmed and running on an IBM-360/75 and a CDC 7600 computer
PAD: a one-dimensional, coupled neutronic-thermodynamic-hydrodynamic computer code
International Nuclear Information System (INIS)
Peterson, D.M.; Stratton, W.R.; McLaughlin, T.P.
1976-12-01
Theoretical and numerical foundations, utilization guide, sample problems, and program listing and glossary are given for the PAD computer code which describes dynamic systems with interactive neutronics, thermodynamics, and hydrodynamics in one-dimensional spherical, cylindrical, and planar geometries. The code has been applied to prompt critical excursions in various fissioning systems (solution, metal, LMFBR, etc.) as well as to nonfissioning systems
Computer codes for tasks in the fields of isotope and radiation research
International Nuclear Information System (INIS)
Friedrich, K.; Gebhardt, O.
1978-11-01
Concise descriptions of computer codes developed for solving problems in the fields of isotope and radiation research at the Zentralinstitut fuer Isotopen- und Strahlenforschung (ZfI) are compiled. In part two the structure of the ZfI program library MABIF is outlined and a complete list of all codes available is given
Moral, Cristian; de Antonio, Angelica; Ferre, Xavier; Lara, Graciela
2015-01-01
Introduction: In this article we propose a qualitative analysis tool--a coding system--that can support the formalisation of the information-seeking process in a specific field: research in computer science. Method: In order to elaborate the coding system, we have conducted a set of qualitative studies, more specifically a focus group and some…
HADES. A computer code for fast neutron cross section from the Optical Model
International Nuclear Information System (INIS)
Guasp, J.; Navarro, C.
1973-01-01
A FORTRAN V computer code for UNIVAC 1108/6 using a local Optical Model with spin-orbit interaction is described. The code calculates fast neutron cross sections, angular distribution, and Legendre moments for heavy and intermediate spherical nuclei. It allows for the possibility of automatic variation of potential parameters for experimental data fitting. (Author) 55 refs
Directory of Open Access Journals (Sweden)
K. V. Pagalthivarthi
2013-03-01
Full Text Available Multi-size particulate dense slurry flow through three-dimensional rectangular channel is modeled using penalty finite elements with 8-noded hexahedral elements. The methodology previously developed for two-dimensional channel is extended. The computed eddy viscosity of the pure carrier flow is modified to account for the presence of solid particles. Predictions from Spalart-Almaras and k-ε turbulence models are compared to show consistency of trends in results. Results are also found to compare well with experimental results from the literature.
Improvement of the computing speed of the FBR fuel pin bundle deformation analysis code 'BAMBOO'
International Nuclear Information System (INIS)
Ito, Masahiro; Uwaba, Tomoyuki
2005-04-01
JNC has developed a coupled analysis system of a fuel pin bundle deformation analysis code 'BAMBOO' and a thermal hydraulics analysis code ASFRE-IV' for the purpose of evaluating the integrity of a subassembly under the BDI condition. This coupled analysis took much computation time because it needs convergent calculations to obtain numerically stationary solutions for thermal and mechanical behaviors. We improved the computation time of the BAMBOO code analysis to make the coupled analysis practicable. 'BAMBOO' is a FEM code and as such its matrix calculations consume large memory area to temporarily stores intermediate results in the solution of simultaneous linear equations. The code used the Hard Disk Drive (HDD) for the virtual memory area to save Random Access Memory (RAM) of the computer. However, the use of the HDD increased the computation time because Input/Output (I/O) processing with the HDD took much time in data accesses. We improved the code in order that it could conduct I/O processing only with the RAM in matrix calculations and run with in high-performance computers. This improvement considerably increased the CPU occupation rate during the simulation and reduced the total simulation time of the BAMBOO code to about one-seventh of that before the improvement. (author)
Computing the Feng-Rao distances for codes from order domains
DEFF Research Database (Denmark)
Ruano Benito, Diego
2007-01-01
We compute the Feng–Rao distance of a code coming from an order domain with a simplicial value semigroup. The main tool is the Apéry set of a semigroup that can be computed using a Gröbner basis.......We compute the Feng–Rao distance of a code coming from an order domain with a simplicial value semigroup. The main tool is the Apéry set of a semigroup that can be computed using a Gröbner basis....
Icarus: A 2-D Direct Simulation Monte Carlo (DSMC) Code for Multi-Processor Computers
International Nuclear Information System (INIS)
BARTEL, TIMOTHY J.; PLIMPTON, STEVEN J.; GALLIS, MICHAIL A.
2001-01-01
Icarus is a 2D Direct Simulation Monte Carlo (DSMC) code which has been optimized for the parallel computing environment. The code is based on the DSMC method of Bird[11.1] and models from free-molecular to continuum flowfields in either cartesian (x, y) or axisymmetric (z, r) coordinates. Computational particles, representing a given number of molecules or atoms, are tracked as they have collisions with other particles or surfaces. Multiple species, internal energy modes (rotation and vibration), chemistry, and ion transport are modeled. A new trace species methodology for collisions and chemistry is used to obtain statistics for small species concentrations. Gas phase chemistry is modeled using steric factors derived from Arrhenius reaction rates or in a manner similar to continuum modeling. Surface chemistry is modeled with surface reaction probabilities; an optional site density, energy dependent, coverage model is included. Electrons are modeled by either a local charge neutrality assumption or as discrete simulational particles. Ion chemistry is modeled with electron impact chemistry rates and charge exchange reactions. Coulomb collision cross-sections are used instead of Variable Hard Sphere values for ion-ion interactions. The electro-static fields can either be: externally input, a Langmuir-Tonks model or from a Green's Function (Boundary Element) based Poison Solver. Icarus has been used for subsonic to hypersonic, chemically reacting, and plasma flows. The Icarus software package includes the grid generation, parallel processor decomposition, post-processing, and restart software. The commercial graphics package, Tecplot, is used for graphics display. All of the software packages are written in standard Fortran
Energy Technology Data Exchange (ETDEWEB)
Joshua J. Cogliati; Abderrafi M. Ougouag
2006-10-01
A comprehensive, high fidelity model for pebble flow has been developed and embodied in the PEBBLES computer code. In this paper, a description of the physical artifacts included in the model is presented and some results from using the computer code for predicting the features of pebble flow and packing in a realistic pebble bed reactor design are shown. The sensitivity of models to various physical parameters is also discussed.
Modification in the CITATION computer code: change of microscopic cross sections by zone
International Nuclear Information System (INIS)
Yamaguchi, M.; Kosaka, N.
1983-01-01
Some modifications done in the CITATION computer code are presented, aiming to calculate the accumulated burnup for each reactor zone in each step of burnup and allow changing the microscopic cross sections for each zone in accordance to the burnup accumulated after each step of burnup. Some input data were put in the computer code. The alterations were tested and the results were compared with and without modifications. (E.G.) [pt
Distribution of absorbed dose in human eye simulated by SRNA-2KG computer code
International Nuclear Information System (INIS)
Ilic, R.; Pesic, M.; Pavlovic, R.; Mostacci, D.
2003-01-01
Rapidly increasing performances of personal computers and development of codes for proton transport based on Monte Carlo methods will allow, very soon, the introduction of the computer planning proton therapy as a normal activity in regular hospital procedures. A description of SRNA code used for such applications and results of calculated distributions of proton-absorbed dose in human eye are given in this paper. (author)
International Nuclear Information System (INIS)
Maudlin, P.J.; Tome, C.N.; Kaschner, G.C.; Gray, G.T. III
1998-01-01
In this work the authors simulate the compressive deformation of heavily textured zirconium sheet using a finite element code with the constitutive response given by a polycrystal self-consistent model. They show that the strong anisotropy of the response can be explained in terms of the texture and the relative activity of prismatic (easy) and pyramidal (hard) slip modes. The simulations capture the yield anisotropy observed for so-called through-thickness and in-plane compression tests in terMs of the loading curves and final specimen geometries
Implementation of thermo-viscoplastic constitutive equations into the finite element code ABAQUS
International Nuclear Information System (INIS)
Youn, Sam Son; Lee, Soon Bok; Kim, Jong Bum; Lee, Hyeong Yeon; Yoo, Bong
1998-01-01
Sophisticated viscoplatic constitutive laws describing material behavior at high temperature have been implemented in the general-purpose finite element code ABAQUS to predict the viscoplastic response of structures to cyclic loading. Because of the complexity of viscoplastic constitutive equation, the general implementation methods are developed. The solution of the non-linear system of algebraic equations arising from time discretization is determined using line-search and back-tracking in combination with Newton method. The time integration method of the constitutive equations is based on semi-implicit method with efficient time step control. For numerical examples, the viscoplastic model proposed by Chaboche is implemented and several applications are illustrated
Development of computer models for fuel element behaviour in water reactors
International Nuclear Information System (INIS)
Gittus, J.H.
1987-03-01
Description of fuel behaviour during normal operation transients and accident conditions has always represented a most challenging and important problem. Reliable predictions constitute a basic demand for safety based calculations, for design purposes and for fuel performance. Therefore, computer codes based on deterministic and probabilistic models were developed. Possibility of comprehensive descriptions of the phenomena is precluded in view of the great number of individual processes, involving physical, chemical, thermohydraulical and mechanical parameters, to be considered in a wide range of situations. In case of fast thermal transients predictive capability is limited by the kinetics of evolution of the system and its eventual dynamic behaviour. Evidently, probabilistic approaches are also limited by the sparcity and limited breadth of the impirical data base. Code predictions have to be evaluated against power reactor data, results from simulation experiments and, if possible, include cross validation of different codes and validation of sub-models. Progress on this subject is reviewed in this report, which completes the co-ordinated research programme on 'Development of Computer Models for Fuel Element Behaviour in Water Reactors' (D-COM), initiated under the auspices of the IAEA in 1981
Challenges in the twentieth century and beyond: Computer codes and data
International Nuclear Information System (INIS)
Kirk, B.L.
1995-01-01
The second half of the twentieth century has seen major changes in computer architecture. From the early fifties to the early seventies, the word open-quotes computerclose quotes demanded reverence, respect, and even fear. Computers, then, were almost open-quotes untouchable.close quotes Computers have become the mainstream of communication on rapidly expanding communication highways. They have become necessities of life. This report describes computer codes and packaging, as well as compilers and operating systems
Solution of the neutronics code dynamic benchmark by finite element method
Avvakumov, A. V.; Vabishchevich, P. N.; Vasilev, A. O.; Strizhov, V. F.
2016-10-01
The objective is to analyze the dynamic benchmark developed by Atomic Energy Research for the verification of best-estimate neutronics codes. The benchmark scenario includes asymmetrical ejection of a control rod in a water-type hexagonal reactor at hot zero power. A simple Doppler feedback mechanism assuming adiabatic fuel temperature heating is proposed. The finite element method on triangular calculation grids is used to solve the three-dimensional neutron kinetics problem. The software has been developed using the engineering and scientific calculation library FEniCS. The matrix spectral problem is solved using the scalable and flexible toolkit SLEPc. The solution accuracy of the dynamic benchmark is analyzed by condensing calculation grid and varying degree of finite elements.
Plutonium explosive dispersal modeling using the MACCS2 computer code
Energy Technology Data Exchange (ETDEWEB)
Steele, C.M.; Wald, T.L.; Chanin, D.I.
1998-11-01
The purpose of this paper is to derive the necessary parameters to be used to establish a defensible methodology to perform explosive dispersal modeling of respirable plutonium using Gaussian methods. A particular code, MACCS2, has been chosen for this modeling effort due to its application of sophisticated meteorological statistical sampling in accordance with the philosophy of Nuclear Regulatory Commission (NRC) Regulatory Guide 1.145, ``Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants``. A second advantage supporting the selection of the MACCS2 code for modeling purposes is that meteorological data sets are readily available at most Department of Energy (DOE) and NRC sites. This particular MACCS2 modeling effort focuses on the calculation of respirable doses and not ground deposition. Once the necessary parameters for the MACCS2 modeling are developed and presented, the model is benchmarked against empirical test data from the Double Tracks shot of project Roller Coaster (Shreve 1965) and applied to a hypothetical plutonium explosive dispersal scenario. Further modeling with the MACCS2 code is performed to determine a defensible method of treating the effects of building structure interaction on the respirable fraction distribution as a function of height. These results are related to the Clean Slate 2 and Clean Slate 3 bunkered shots of Project Roller Coaster. Lastly a method is presented to determine the peak 99.5% sector doses on an irregular site boundary in the manner specified in NRC Regulatory Guide 1.145 (1983). Parametric analyses are performed on the major analytic assumptions in the MACCS2 model to define the potential errors that are possible in using this methodology.
Plutonium explosive dispersal modeling using the MACCS2 computer code
International Nuclear Information System (INIS)
Steele, C.M.; Wald, T.L.; Chanin, D.I.
1998-01-01
The purpose of this paper is to derive the necessary parameters to be used to establish a defensible methodology to perform explosive dispersal modeling of respirable plutonium using Gaussian methods. A particular code, MACCS2, has been chosen for this modeling effort due to its application of sophisticated meteorological statistical sampling in accordance with the philosophy of Nuclear Regulatory Commission (NRC) Regulatory Guide 1.145, ''Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants''. A second advantage supporting the selection of the MACCS2 code for modeling purposes is that meteorological data sets are readily available at most Department of Energy (DOE) and NRC sites. This particular MACCS2 modeling effort focuses on the calculation of respirable doses and not ground deposition. Once the necessary parameters for the MACCS2 modeling are developed and presented, the model is benchmarked against empirical test data from the Double Tracks shot of project Roller Coaster (Shreve 1965) and applied to a hypothetical plutonium explosive dispersal scenario. Further modeling with the MACCS2 code is performed to determine a defensible method of treating the effects of building structure interaction on the respirable fraction distribution as a function of height. These results are related to the Clean Slate 2 and Clean Slate 3 bunkered shots of Project Roller Coaster. Lastly a method is presented to determine the peak 99.5% sector doses on an irregular site boundary in the manner specified in NRC Regulatory Guide 1.145 (1983). Parametric analyses are performed on the major analytic assumptions in the MACCS2 model to define the potential errors that are possible in using this methodology
Status Report on Hydrogen Management and Related Computer Codes
International Nuclear Information System (INIS)
Liang, Z.; Chan, C.K.; Sonnenkalb, M.; Bentaib, A.; Malet, J.; Sangiorgi, M.; Gryffroy, D.; Gyepi-Garbrah, S.; Duspiva, J.; Sevon, T.; Kelm, S.; Reinecke, E.A.; Xu, Z.J.; Cervone, A.; Utsuno, H.; Hotta, A.; Hong, S.W.; Kim, J.T.; Visser, D.C.; Stempniewicz, M.M.; Kuriene, L.; Prusinski, P.; Martin-Valdepenas, J.M.; Frid, W.; Isaksson, P.; Dreier, J.; Paladino, D.; Algama, D.; Notafrancesco, A.; Amri, A.; Kissane, M.; )
2014-01-01
In follow-up to the Fukushima Daiichi NPP accident, the Committee on the Safety of Nuclear Installations (CSNI) decided to launch several high priority activities. At the 14. plenary meeting of the Working Group on Analysis and Management of Accidents (WGAMA), a proposal for a status paper on hydrogen generation, transport and mitigation under severe accident conditions was approved. The proposed activity is in line with the WGAMA mandate and it was considered to be needed to revisit the hydrogen issue. The report is broken down into five Chapters and two appendixes. Chapter 1 provides background information for this activity and expected topics defined by the WGAMA members. A general understanding of hydrogen behavior and control in severe accidents is discussed. A brief literature review is included in this chapter to summarize the progress obtained from the early US NRC sponsored research on hydrogen and recent international OECD or EC sponsored projects on hydrogen related topics (generation, distribution, combustion and mitigation). Chapter 2 provides a general overview of the various reactor designs of Western PWRs, BWRs, Eastern European VVERs and PHWRs (CANDUs). The purpose is to understand the containment design features in relation to hydrogen management measures. Chapter 3 provides a detailed description of national requirements on hydrogen management and hydrogen mitigation measures inside the containment and other places (e.g., annulus space, secondary buildings, spent fuel pool, etc.). Discussions are followed on hydrogen analysis approaches, application of safety systems (e.g., spray, containment ventilation, local air cooler, suppression pool, and latch systems), hydrogen measurement strategies as well as lessons learnt from the Fukushima Daiichi nuclear power accident. Chapter 4 provides an overview of various codes that are being used for hydrogen risk assessment, and the codes capabilities and validation status in terms of hydrogen related
Computational fluid mechanics qualification calculations for the code TEACH
International Nuclear Information System (INIS)
DeGrazia, M.C.; Fitzsimmons, L.B.; Reynolds, J.T.
1979-11-01
KAPL is developing a predictive method for three-dimensional (3-D) turbulent fluid flow configurations typically encountered in the thermal-hydraulic design of a nuclear reactor. A series of experiments has been selected for analysis to investigate the adequacy of the two-equation turbulence model developed at Imperial College, London, England for predicting the flow patterns in simple geometries. The analysis of these experiments is described with the two-dimensional (2-D) turbulent fluid flow code TEACH. This work qualifies TEACH for a variety of geometries and flow conditions
CONDOR: a database resource of developmentally associated conserved non-coding elements
Directory of Open Access Journals (Sweden)
Smith Sarah
2007-08-01
Full Text Available Abstract Background Comparative genomics is currently one of the most popular approaches to study the regulatory architecture of vertebrate genomes. Fish-mammal genomic comparisons have proved powerful in identifying conserved non-coding elements likely to be distal cis-regulatory modules such as enhancers, silencers or insulators that control the expression of genes involved in the regulation of early development. The scientific community is showing increasing interest in characterizing the function, evolution and language of these sequences. Despite this, there remains little in the way of user-friendly access to a large dataset of such elements in conjunction with the analysis and the visualization tools needed to study them. Description Here we present CONDOR (COnserved Non-coDing Orthologous Regions available at: http://condor.fugu.biology.qmul.ac.uk. In an interactive and intuitive way the website displays data on > 6800 non-coding elements associated with over 120 early developmental genes and conserved across vertebrates. The database regularly incorporates results of ongoing in vivo zebrafish enhancer assays of the CNEs carried out in-house, which currently number ~100. Included and highlighted within this set are elements derived from duplication events both at the origin of vertebrates and more recently in the teleost lineage, thus providing valuable data for studying the divergence of regulatory roles between paralogs. CONDOR therefore provides a number of tools and facilities to allow scientists to progress in their own studies on the function and evolution of developmental cis-regulation. Conclusion By providing access to data with an approachable graphics interface, the CONDOR database presents a rich resource for further studies into the regulation and evolution of genes involved in early development.
Ivanov, V A; Melnikov, A A; Siunov, A V; Fodor, I I; Ilyin, Y V
1991-01-01
The mobile element jockey is similar in structural organization and coding potential to the LINEs of various organisms. It is transcribed at different stages of Drosophila ontogenesis. The Drosophila LINE family includes active transposable elements. Current models for the mechanism of transposition involve reverse transcription of an RNA intermediate and utilization of element-encoded proteins. As demonstrated here, a 2.23 kb DNA fragment from the region of jockey encoding the putative reverse transcriptase was stably introduced into an expression system under inducible control of the Escherichia coli lac regulatory elements. We describe the expression of the 92 kDa protein and identify this polypeptide alone as the authentic jockey reverse transcriptase based on some of its physical and enzymic properties. The jockey polymerase demonstrates RNA and DNA-directed DNA polymerase activities but lacks detectable RNase H, has a temperature optimum at 26 degrees C, requires Mg2+ or Mn2+ as a cofactor and is inactivated by sulphydryl reagent. The enzyme prefers poly(rC) and poly(rA) as template and 'activated' DNA is not effective. Images PMID:1714378
Computational promoter analysis of mouse, rat and human antimicrobial peptide-coding genes
Directory of Open Access Journals (Sweden)
Kai Chikatoshi
2006-12-01
Full Text Available Abstract Background Mammalian antimicrobial peptides (AMPs are effectors of the innate immune response. A multitude of signals coming from pathways of mammalian pathogen/pattern recognition receptors and other proteins affect the expression of AMP-coding genes (AMPcgs. For many AMPcgs the promoter elements and transcription factors that control their tissue cell-specific expression have yet to be fully identified and characterized. Results Based upon the RIKEN full-length cDNA and public sequence data derived from human, mouse and rat, we identified 178 candidate AMP transcripts derived from 61 genes belonging to 29 AMP families. However, only for 31 mouse genes belonging to 22 AMP families we were able to determine true orthologous relationships with 30 human and 15 rat sequences. We screened the promoter regions of AMPcgs in the three species for motifs by an ab initio motif finding method and analyzed the derived promoter characteristics. Promoter models were developed for alpha-defensins, penk and zap AMP families. The results suggest a core set of transcription factors (TFs that regulate the transcription of AMPcg families in mouse, rat and human. The three most frequent core TFs groups include liver-, nervous system-specific and nuclear hormone receptors (NHRs. Out of 440 motifs analyzed, we found that three represent potentially novel TF-binding motifs enriched in promoters of AMPcgs, while the other four motifs appear to be species-specific. Conclusion Our large-scale computational analysis of promoters of 22 families of AMPcgs across three mammalian species suggests that their key transcriptional regulators are likely to be TFs of the liver-, nervous system-specific and NHR groups. The computationally inferred promoter elements and potential TF binding motifs provide a rich resource for targeted experimental validation of TF binding and signaling studies that aim at the regulation of mouse, rat or human AMPcgs.
TPASS: a gamma-ray spectrum analysis and isotope identification computer code
International Nuclear Information System (INIS)
Dickens, J.K.
1981-03-01
The gamma-ray spectral data-reduction and analysis computer code TPASS is described. This computer code is used to analyze complex Ge(Li) gamma-ray spectra to obtain peak areas corrected for detector efficiencies, from which are determined gamma-ray yields. These yields are compared with an isotope gamma-ray data file to determine the contributions to the observed spectrum from decay of specific radionuclides. A complete FORTRAN listing of the code and a complex test case are given
The ATOMek computer code - graphical interface for the ROOT
International Nuclear Information System (INIS)
Gburek, T.; Siwek, A.
2002-05-01
In this report we describe a computer program ATOMek. It is a data analysis and visualization program written in C++ which utilizes the Linux XWindow system and classes of ROOT data analysis framework. It replace hardcore C++ programming by set of graphic controls (buttons, menus, tabs etc.) and mouse operations - therefore it can be treated as an overlay for ROOT. (author)
Shape analysis of neutron transmission resonances - a computer code
International Nuclear Information System (INIS)
Giacobbe, P.; Magnani, M.
1979-01-01
A computer programme for shape-analysis of time-of-flight neutron transmission spectra of non-fissile nuclei in the resonance region is described. The main features are: partial independence of the structure of the programme for the formalism used, accurate description of the resolution function, use of ''a priori'' information on the fitted parameters and simultaneous analysis of many spectra
Development of easy-to-use interface for nuclear transmutation computing, VCINDER code
Directory of Open Access Journals (Sweden)
Oyeon Kum
2018-02-01
Full Text Available The CINDER code has about 60 years of development history, and is thus one of the world's best transmutation computing codes to date. Unfortunately, it is complex and cumbersome to use. Preparing auxiliary input files for activation computation from MCNPX output and executing them using Perl script (activation script is the first difficulty, and separation of gamma source computing script (gamma script, which analyzes the spectra files produced by CINDER code and creates source definition format for MCNPX code, is the second difficulty. In addition, for highly nonlinear problems, multiple human interventions may increase the possibility of errors. Postprocessing such as making plots with large text outputs is also time consuming. One way to improve these limitations is to make a graphical user interface wrapper that includes all codes, such as MCNPX and CINDER, and all scripts with a visual C#.NET tool. The graphical user interface merges all the codes and provides easy postprocessing of graphics data and Microsoft office tools, such as Excel sheets, which make the CINDER code easy to use. This study describes the VCINDER code (with visual C#.NET and gives a typical application example.
Fuel element thermo-mechanical analysis during transient events using the FMS and FETMA codes
International Nuclear Information System (INIS)
Hernandez Lopez Hector; Hernandez Martinez Jose Luis; Ortiz Villafuerte Javier
2005-01-01
In the Instituto Nacional de Investigaciones Nucleares of Mexico, the Fuel Management System (FMS) software package has been used for long time to simulate the operation of a BWR nuclear power plant in steady state, as well as in transient events. To evaluate the fuel element thermo-mechanical performance during transient events, an interface between the FMS codes and our own Fuel Element Thermo Mechanical Analysis (FETMA) code is currently being developed and implemented. In this work, the results of the thermo-mechanical behavior of fuel rods in the hot channel during the simulation of transient events of a BWR nuclear power plant are shown. The transient events considered for this work are a load rejection and a feedwater control failure, which among the most important events that can occur in a BWR. The results showed that conditions leading to fuel rod failure at no time appeared for both events. Also, it is shown that a transient due load rejection is more demanding on terms of safety that the failure of a controller of the feedwater. (authors)
Comparison of various NLTE codes in computing the charge-state populations of an argon plasma
Energy Technology Data Exchange (ETDEWEB)
Stone, S.R.; Weisheit, J.C.
1984-11-01
A comparison among nine computer codes shows surprisingly large differences where it had been believed that the theroy was well understood. Each code treats an argon plasma, optically thin and with no external photon flux; temperatures vary around 1 keV and ion densities vary from 6 x 10/sup 17/ cm/sup -3/ to 6 x 10/sup 21/ cm/sup -3/. At these conditions most ions have three or fewer bound electrons. The calculated populations of 0-, 1-, 2-, and 3-electron ions differ from code to code by typical factors of 2, in some cases by factors greater than 300. These differences depend as sensitively on how may Rydberg states a code allows as they do on variations among computed collision rates. 29 refs., 23 figs.
Proceedings of the conference on computer codes and the linear accelerator community
International Nuclear Information System (INIS)
Cooper, R.K.
1990-07-01
The conference whose proceedings you are reading was envisioned as the second in a series, the first having been held in San Diego in January 1988. The intended participants were those people who are actively involved in writing and applying computer codes for the solution of problems related to the design and construction of linear accelerators. The first conference reviewed many of the codes both extant and under development. This second conference provided an opportunity to update the status of those codes, and to provide a forum in which emerging new 3D codes could be described and discussed. The afternoon poster session on the second day of the conference provided an opportunity for extended discussion. All in all, this conference was felt to be quite a useful interchange of ideas and developments in the field of 3D calculations, parallel computation, higher-order optics calculations, and code documentation and maintenance for the linear accelerator community. A third conference is planned
Equivalence of computer codes for calculation of coincidence summing correction factors
International Nuclear Information System (INIS)
Vidmar, T.; Capogni, M.; Hult, M.; Hurtado, S.; Kastlander, J.; Lutter, G.; Lépy, M.-C.; Martinkovič, J.; Ramebäck, H.; Sima, O.; Tzika, F.; Vidmar, G.
2014-01-01
The aim of the study was to check for equivalence of computer codes that can perform calculations of true coincidence summing correction factors. All calculations were performed for a set of well-defined detector and sample parameters, without any reference to empirical data. For a p-type detector model the application of different codes resulted in satisfactory agreement in the calculated correction factors. For high-efficiency geometries in combination with an n-type detector and a radionuclide emitting abundant X-rays the results were scattered. - Highlights: • Measured and calculated true coincidence corrections differ in recent study. • We checked for equivalence of computer codes that can perform such calculations. • Codes compared with each other for a set of well-defined detector and sample models. • Satisfactory agreement between the codes for a p-type detector model. • Large differences for an n-type detector model and X-ray emitting radionuclides
Proceedings of the conference on computer codes and the linear accelerator community
Energy Technology Data Exchange (ETDEWEB)
Cooper, R.K. (comp.)
1990-07-01
The conference whose proceedings you are reading was envisioned as the second in a series, the first having been held in San Diego in January 1988. The intended participants were those people who are actively involved in writing and applying computer codes for the solution of problems related to the design and construction of linear accelerators. The first conference reviewed many of the codes both extant and under development. This second conference provided an opportunity to update the status of those codes, and to provide a forum in which emerging new 3D codes could be described and discussed. The afternoon poster session on the second day of the conference provided an opportunity for extended discussion. All in all, this conference was felt to be quite a useful interchange of ideas and developments in the field of 3D calculations, parallel computation, higher-order optics calculations, and code documentation and maintenance for the linear accelerator community. A third conference is planned.
Visualization of elastic wavefields computed with a finite difference code
Energy Technology Data Exchange (ETDEWEB)
Larsen, S. [Lawrence Livermore National Lab., CA (United States); Harris, D.
1994-11-15
The authors have developed a finite difference elastic propagation model to simulate seismic wave propagation through geophysically complex regions. To facilitate debugging and to assist seismologists in interpreting the seismograms generated by the code, they have developed an X Windows interface that permits viewing of successive temporal snapshots of the (2D) wavefield as they are calculated. The authors present a brief video displaying the generation of seismic waves by an explosive source on a continent, which propagate to the edge of the continent then convert to two types of acoustic waves. This sample calculation was part of an effort to study the potential of offshore hydroacoustic systems to monitor seismic events occurring onshore.
Spent fuel management fee methodology and computer code user's manual
International Nuclear Information System (INIS)
Engel, R.L.; White, M.K.
1982-01-01
The methodology and computer model described here were developed to analyze the cash flows for the federal government taking title to and managing spent nuclear fuel. The methodology has been used by the US Department of Energy (DOE) to estimate the spent fuel disposal fee that will provide full cost recovery. Although the methodology was designed to analyze interim storage followed by spent fuel disposal, it could be used to calculate a fee for reprocessing spent fuel and disposing of the waste. The methodology consists of two phases. The first phase estimates government expenditures for spent fuel management. The second phase determines the fees that will result in revenues such that the government attains full cost recovery assuming various revenue collection philosophies. These two phases are discussed in detail in subsequent sections of this report. Each of the two phases constitute a computer module, called SPADE (SPent fuel Analysis and Disposal Economics) and FEAN (FEe ANalysis), respectively
Computer code for general analysis of radon risks (GARR)
International Nuclear Information System (INIS)
Ginevan, M.
1984-09-01
This document presents a computer model for general analysis of radon risks that allow the user to specify a large number of possible models with a small number of simple commands. The model is written in a version of BASIC which conforms closely to the American National Standards Institute (ANSI) definition for minimal BASIC and thus is readily modified for use on a wide variety of computers and, in particular, microcomputers. Model capabilities include generation of single-year life tables from 5-year abridged data, calculation of multiple-decrement life tables for lung cancer for the general population, smokers, and nonsmokers, and a cohort lung cancer risk calculation that allows specification of level and duration of radon exposure, the form of the risk model, and the specific population assumed at risk. 36 references, 8 figures, 7 tables
Characterization of the MCNPX computer code in micro processed architectures
International Nuclear Information System (INIS)
Almeida, Helder C.; Dominguez, Dany S.; Orellana, Esbel T.V.; Milian, Felix M.
2009-01-01
The MCNPX (Monte Carlo N-Particle extended) can be used to simulate the transport of several types of nuclear particles, using probabilistic methods. The technique used for MCNPX is to follow the history of each particle from its origin to its extinction that can be given by absorption, escape or other reasons. To obtain accurate results in simulations performed with the MCNPX is necessary to process a large number of histories, which demand high computational cost. Currently the MCNPX can be installed in virtually all computing platforms available, however there is virtually no information on the performance of the application in each. This paper studies the performance of MCNPX, to work with electrons and photons in phantom Faux on two platforms used by most researchers, Windows and Li nux. Both platforms were tested on the same computer to ensure the reliability of the hardware in the measures of performance. The performance of MCNPX was measured by time spent to run a simulation, making the variable time the main measure of comparison. During the tests the difference in performance between the two platforms MCNPX was evident. In some cases we were able to gain speed more than 10% only with the exchange platforms, without any specific optimization. This shows the relevance of the study to optimize this tool on the platform most appropriate for its use. (author)
International Nuclear Information System (INIS)
Lundsager, P.; Krenk, S.
1975-08-01
The static and dynamic response of a cylindrical/ spherical containment to a Boeing 720 impact is computed using 3 different linear elastic computer codes: FINEL, SAP and STARDYNE. Stress and displacement fields are shown together with time histories for a point in the impact zone. The main conclusions from this study are: - In this case the maximum dynamic load factors for stress and displacements were close to 1, but a static analysis alone is not fully sufficient. - More realistic load time histories should be considered. - The main effects seem to be local. The present study does not indicate general collapse from elastic stresses alone. - Further study of material properties at high rates is needed. (author)
Compendium of computer codes for the safety analysis of fast breeder reactors
Energy Technology Data Exchange (ETDEWEB)
1977-10-01
The objective of the compendium is to provide the reader with a guide which briefly describes many of the computer codes used for liquid metal fast breeder reactor safety analyses, since it is for this system that most of the codes have been developed. The compendium is designed to address the following frequently asked questions from individuals in licensing and research and development activities: (1) What does the code do. (2) To what safety problems has it been applied. (3) What are the code's limitations. (4) What is being done to remove these limitations. (5) How does the code compare with experimental observations and other code predictions. (6) What reference documents are available.
Benchmark testing and independent verification of the VS2DT computer code
International Nuclear Information System (INIS)
McCord, J.T.
1994-11-01
The finite difference flow and transport simulator VS2DT was benchmark tested against several other codes which solve the same equations (Richards equation for flow and the Advection-Dispersion equation for transport). The benchmark problems investigated transient two-dimensional flow in a heterogeneous soil profile with a localized water source at the ground surface. The VS2DT code performed as well as or better than all other codes when considering mass balance characteristics and computational speed. It was also rated highly relative to the other codes with regard to ease-of-use. Following the benchmark study, the code was verified against two analytical solutions, one for two-dimensional flow and one for two-dimensional transport. These independent verifications show reasonable agreement with the analytical solutions, and complement the one-dimensional verification problems published in the code's original documentation
User's manual for the G.T.M.-1 computer code
International Nuclear Information System (INIS)
Prado-Herrero, P.
1992-01-01
This document describes the GTM-1 ( Geosphere Transport Model, release-1) computer code and is intended to provide the reader with enough detailed information in order to use the code. GTM-1 was developed for the assessment of radionuclide migration by the ground water through geologic deposits whose properties can change along the pathway.GTM-1 solves the transport equation by the finite differences method ( Crank-Nicolson scheme ). It was developped for specific use within Probabilistic System Assessment (PSA) Monte Carlo Method codes; in this context the first application of GTM-1 was within the LISA (Long Term Isolation System Assessment) code. GTM-1 is also available as an independent model, which includes various submodels simulating a multi-barrier disposal system. The code has been tested with the PSACOIN ( Probabilistic System Assessment Codes intercomparison) benchmarks exercises from PSAC User Group (OECD/NEA). 10 refs., 6 Annex., 2 tabs
International Nuclear Information System (INIS)
Rudolph, Juergen; Goetz, Andreas; Hilpert, Roland
2012-01-01
The procedures of fatigue analyses of several relevant nuclear and conventional design codes (ASME, KTA, EN, AD) for power plant components differentiate between an elastic, simplified elastic-plastic and elastic-plastic fatigue check. As a rule, operational load levels will exclude the purely elastic fatigue check. The application of the code procedure of the simplified elastic-plastic fatigue check is common practice. Nevertheless, resulting cumulative usage factors may be overly conservative mainly due to high code based plastification penalty factors Ke. As a consequence, the more complex and still code conforming general elastic-plastic fatigue analysis methodology based on non-linear finite element analysis (FEA) is applied for fatigue design as an alternative. The requirements of the FEA and the material law to be applied have to be clarified in a first step. Current design codes only give rough guidelines on these relevant items. While the procedure for the simplified elastic-plastic fatigue analysis and the associated code passages are based on stress related cycle counting and the determination of pseudo elastic equivalent stress ranges, an adaptation to elastic-plastic strains and strain ranges is required for the elastic-plastic fatigue check. The associated requirements are explained in detail in the paper. If the established and implemented evaluation mechanism (cycle counting according to the peak and valley respectively the rainflow method, calculation of stress ranges from arbitrary load-time histories and determination of cumulative usage factors based on all load events) is to be retained, a conversion of elastic-plastic strains and strain ranges into pseudo elastic stress ranges is required. The algorithm to be applied is described in the paper. It has to be implemented in the sense of an extended post processing operation of FEA e.g. by APDL scripts in ANSYS registered . Variations of principal stress (strain) directions during the loading
The extensive international use of commercial computational fluid dynamics (CFD) codes
International Nuclear Information System (INIS)
Hartmut Wider
2005-01-01
What are the main reasons for the extensive international success of commercial CFD codes? This is due to their ability to calculate the fine structures of the investigated processes due to their versatility, their numerical stability and that they can guarantee the proper solution in most cases. This was made possible by the constantly increasing computer power at an ever more affordable prize. Furthermore it is much more efficient to have researchers use a CFD code rather than to develop a similar code system due to the time consuming nature of this activity and the high probability of hidden coding errors. The centralized development and upgrading makes these reliable CFD codes possible and affordable. However, the CFD companies' developments are naturally concentrated on the most profitable areas, and thus, if one works in a 'non-priority' field one cannot use them. Moreover, the prize of renting CFD codes, applications to complex systems such as whole nuclear reactors and the need to teach students gives the development of self-made codes still plenty of room. But CFD codes can model detailed aspects of large systems and subroutines generated by users can be added. Since there are only a few heavily used CFD codes such as FLUENT, STAR-CD, ANSYS CFX, these are used in many countries. Also international training courses are given and the news bulletins of these codes help to spread the news on further developments. A larger number of international codes would increase the competition but would at the same time make it harder to select the most appropriate CFD code for a given problem. Examples will be presented of uses of CFD codes as more detailed system codes for the decay heat removal from reactors, the application to aerosol physics and the application to heavy metal fluids using different turbulence models. (author)
GIANT: a computer code for General Interactive ANalysis of Trajectories
International Nuclear Information System (INIS)
Jaeger, J.; Lee, M.; Servranckx, R.; Shoaee, H.
1985-04-01
Many model-driven diagnostic and correction procedures have been developed at SLAC for the on-line computer controlled operation of SPEAR, PEP, the LINAC, and the Electron Damping Ring. In order to facilitate future applications and enhancements, these procedures are being collected into a single program, GIANT. The program allows interactive diagnosis as well as performance optimization of any beam transport line or circular machine. The test systems for GIANT are those of the SLC project. The organization of this program and some of the recent applications of the procedures will be described in this paper
ASAS: Computational code for Analysis and Simulation of Atomic Spectra
Directory of Open Access Journals (Sweden)
Jhonatha R. dos Santos
2017-01-01
Full Text Available The laser isotopic separation process is based on the selective photoionization principle and, because of this, it is necessary to know the absorption spectrum of the desired atom. Computational resource has become indispensable for the planning of experiments and analysis of the acquired data. The ASAS (Analysis and Simulation of Atomic Spectra software presented here is a helpful tool to be used in studies involving atomic spectroscopy. The input for the simulations is friendly and essentially needs a database containing the energy levels and spectral lines of the atoms subjected to be studied.
Verification of thermal-hydraulic computer codes against standard problems for WWER reflooding
International Nuclear Information System (INIS)
Alexander D Efanov; Vladimir N Vinogradov; Victor V Sergeev; Oleg A Sudnitsyn
2005-01-01
Full text of publication follows: The computational assessment of reactor core components behavior under accident conditions is impossible without knowledge of the thermal-hydraulic processes occurring in this case. The adequacy of the results obtained using the computer codes to the real processes is verified by carrying out a number of standard problems. In 2000-2003, the fulfillment of three Russian standard problems on WWER core reflooding was arranged using the experiments on full-height electrically heated WWER 37-rod bundle model cooldown in regimes of bottom (SP-1), top (SP-2) and combined (SP-3) reflooding. The representatives from the eight MINATOM's organizations took part in this work, in the course of which the 'blind' and posttest calculations were performed using various versions of the RELAP5, ATHLET, CATHARE, COBRA-TF, TRAP, KORSAR computer codes. The paper presents a brief description of the test facility, test section, test scenarios and conditions as well as the basic results of computational analysis of the experiments. The analysis of the test data revealed a significantly non-one-dimensional nature of cooldown and rewetting of heater rods heated up to a high temperature in a model bundle. This was most pronounced at top and combined reflooding. The verification of the model reflooding computer codes showed that most of computer codes fairly predict the peak rod temperature and the time of bundle cooldown. The exception is provided by the results of calculations with the ATHLET and CATHARE codes. The nature and rate of rewetting front advance in the lower half of the bundle are fairly predicted practically by all computer codes. The disagreement between the calculations and experimental results for the upper half of the bundle is caused by the difficulties of computational simulation of multidimensional effects by 1-D computer codes. In this regard, a quasi-two-dimensional computer code COBRA-TF offers certain advantages. Overall, the closest
The Uncertainty Test for the MAAP Computer Code
International Nuclear Information System (INIS)
Park, S. H.; Song, Y. M.; Park, S. Y.; Ahn, K. I.; Kim, K. R.; Lee, Y. J.
2008-01-01
After the Three Mile Island Unit 2 (TMI-2) and Chernobyl accidents, safety issues for a severe accident are treated in various aspects. Major issues in our research part include a level 2 PSA. The difficulty in expanding the level 2 PSA as a risk information activity is the uncertainty. In former days, it attached a weight to improve the quality in a internal accident PSA, but the effort is insufficient for decrease the phenomenon uncertainty in the level 2 PSA. In our country, the uncertainty degree is high in the case of a level 2 PSA model, and it is necessary to secure a model to decrease the uncertainty. We have not yet experienced the uncertainty assessment technology, the assessment system itself depends on advanced nations. In advanced nations, the severe accident simulator is implemented in the hardware level. But in our case, basic function in a software level can be implemented. In these circumstance at home and abroad, similar instances are surveyed such as UQM and MELCOR. Referred to these instances, SAUNA (Severe Accident UNcertainty Analysis) system is being developed in our project to assess and decrease the uncertainty in a level 2 PSA. It selects the MAAP code to analyze the uncertainty in a severe accident
Automatic Parallelization Tool: Classification of Program Code for Parallel Computing
Directory of Open Access Journals (Sweden)
Mustafa Basthikodi
2016-04-01
Full Text Available Performance growth of single-core processors has come to a halt in the past decade, but was re-enabled by the introduction of parallelism in processors. Multicore frameworks along with Graphical Processing Units empowered to enhance parallelism broadly. Couples of compilers are updated to developing challenges forsynchronization and threading issues. Appropriate program and algorithm classifications will have advantage to a great extent to the group of software engineers to get opportunities for effective parallelization. In present work we investigated current species for classification of algorithms, in that related work on classification is discussed along with the comparison of issues that challenges the classification. The set of algorithms are chosen which matches the structure with different issues and perform given task. We have tested these algorithms utilizing existing automatic species extraction toolsalong with Bones compiler. We have added functionalities to existing tool, providing a more detailed characterization. The contributions of our work include support for pointer arithmetic, conditional and incremental statements, user defined types, constants and mathematical functions. With this, we can retain significant data which is not captured by original speciesof algorithms. We executed new theories into the device, empowering automatic characterization of program code.
Overview of adaptive finite element analysis in computational geodynamics
May, D. A.; Schellart, W. P.; Moresi, L.
2013-10-01
The use of numerical models to develop insight and intuition into the dynamics of the Earth over geological time scales is a firmly established practice in the geodynamics community. As our depth of understanding grows, and hand-in-hand with improvements in analytical techniques and higher resolution remote sensing of the physical structure and state of the Earth, there is a continual need to develop more efficient, accurate and reliable numerical techniques. This is necessary to ensure that we can meet the challenge of generating robust conclusions, interpretations and predictions from improved observations. In adaptive numerical methods, the desire is generally to maximise the quality of the numerical solution for a given amount of computational effort. Neither of these terms has a unique, universal definition, but typically there is a trade off between the number of unknowns we can calculate to obtain a more accurate representation of the Earth, and the resources (time and computational memory) required to compute them. In the engineering community, this topic has been extensively examined using the adaptive finite element (AFE) method. Recently, the applicability of this technique to geodynamic processes has started to be explored. In this review we report on the current status and usage of spatially adaptive finite element analysis in the field of geodynamics. The objective of this review is to provide a brief introduction to the area of spatially adaptive finite analysis, including a summary of different techniques to define spatial adaptation and of different approaches to guide the adaptive process in order to control the discretisation error inherent within the numerical solution. An overview of the current state of the art in adaptive modelling in geodynamics is provided, together with a discussion pertaining to the issues related to using adaptive analysis techniques and perspectives for future research in this area. Additionally, we also provide a
A restructuring of the MELCOR fission product packages for the MIDAS computer code
International Nuclear Information System (INIS)
Park, S.H.; Kim, K.R.; Kim, D.H.
2004-01-01
The RN1/RN2 packages, which are the fission product-related packages in MELCOR, have been restructured for the MIDAS computer code. MIDAS is being developed as an integrated severe accident analysis code with a user-friendly graphical user interface and a modernized data structure. To do this, the data transferring methods of the current MELCOR code are modified and adopted into the RN1/RN2 package. The data structure of the current MELCOR code using FORTRAN77 has a difficulty in grasping the meaning of the variables as well as waste of memory. New features of FORTRAN90 make it possible to allocate the storage dynamically and to user-defined data type, which leads to an efficient memory treatment and an easy understanding of the code. Restructuring of the RN1/RN2 package addressed in this paper includes a module development, subroutine modification, and the treatment of MELGEN, which generates the data file, as well as MELCOR, which is processing the calculation. The verification has been done by comparing the results of the modified code with those of the existing code. As the trends are similar to each other, it implies that the same approach could be extended to the entire code package. It is expected that the code restructuring will accelerate the code domestication thanks to a direct understanding of each variable and an easy implementation of the modified or newly developed models. (author)
ABAQUS-EPGEN: a general-purpose finite element code. Volume 3. Example problems manual
International Nuclear Information System (INIS)
Hibbitt, H.D.; Karlsson, B.I.; Sorensen, E.P.
1983-03-01
This volume is the Example and Verification Problems Manual for ABAQUS/EPGEN. Companion volumes are the User's, Theory and Systems Manuals. This volume contains two major parts. The bulk of the manual (Sections 1-8) contains worked examples that are discussed in detail, while Appendix A documents a large set of basic verification cases that provide the fundamental check of the elements in the code. The examples in Sections 1-8 illustrate and verify significant aspects of the program's capability. Most of these problems provide verification, but they have also been chosen to allow discussion of modeling and analysis techniques. Appendix A contains basic verification cases. Each of these cases verifies one element in the program's library. The verification consists of applying all possible load or flux types (including thermal loading of stress elements), and all possible foundation or film/radiation conditions, and checking the resulting force and stress solutions or flux and temperature results. This manual provides program verification. All of the problems described in the manual are run and the results checked, for each release of the program, and these verification results are made available
Predoi, Mihai Valentin
2014-09-01
The dispersion curves for hollow multilayered cylinders are prerequisites in any practical guided waves application on such structures. The equations for homogeneous isotropic materials have been established more than 120 years ago. The difficulties in finding numerical solutions to analytic expressions remain considerable, especially if the materials are orthotropic visco-elastic as in the composites used for pipes in the last decades. Among other numerical techniques, the semi-analytical finite elements method has proven its capability of solving this problem. Two possibilities exist to model a finite elements eigenvalue problem: a two-dimensional cross-section model of the pipe or a radial segment model, intersecting the layers between the inner and the outer radius of the pipe. The last possibility is here adopted and distinct differential problems are deduced for longitudinal L(0,n), torsional T(0,n) and flexural F(m,n) modes. Eigenvalue problems are deduced for the three modes classes, offering explicit forms of each coefficient for the matrices used in an available general purpose finite elements code. Comparisons with existing solutions for pipes filled with non-linear viscoelastic fluid or visco-elastic coatings as well as for a fully orthotropic hollow cylinder are all proving the reliability and ease of use of this method. Copyright © 2014 Elsevier B.V. All rights reserved.
Issues in computational fluid dynamics code verification and validation
Energy Technology Data Exchange (ETDEWEB)
Oberkampf, W.L.; Blottner, F.G.
1997-09-01
A broad range of mathematical modeling errors of fluid flow physics and numerical approximation errors are addressed in computational fluid dynamics (CFD). It is strongly believed that if CFD is to have a major impact on the design of engineering hardware and flight systems, the level of confidence in complex simulations must substantially improve. To better understand the present limitations of CFD simulations, a wide variety of physical modeling, discretization, and solution errors are identified and discussed. Here, discretization and solution errors refer to all errors caused by conversion of the original partial differential, or integral, conservation equations representing the physical process, to algebraic equations and their solution on a computer. The impact of boundary conditions on the solution of the partial differential equations and their discrete representation will also be discussed. Throughout the article, clear distinctions are made between the analytical mathematical models of fluid dynamics and the numerical models. Lax`s Equivalence Theorem and its frailties in practical CFD solutions are pointed out. Distinctions are also made between the existence and uniqueness of solutions to the partial differential equations as opposed to the discrete equations. Two techniques are briefly discussed for the detection and quantification of certain types of discretization and grid resolution errors.
International Nuclear Information System (INIS)
Zyvoloski, G.A.; Robinson, B.A.; Dash, Z.V.; Trease, L.L.
1997-07-01
The mathematical models and numerical methods employed by the FEHM application, a finite-element heat- and mass-transfer computer code that can simulate nonisothermal multiphase multi-component flow in porous media, are described. The use of this code is applicable to natural-state studies of geothermal systems and groundwater flow. A primary use of the FEHM application will be to assist in the understanding of flow fields and mass transport in the saturated and unsaturated zones below the proposed Yucca Mountain nuclear waste repository in Nevada. The component models of FEHM are discussed. The first major component, Flow- and Energy-Transport Equations, deals with heat conduction; heat and mass transfer with pressure- and temperature-dependent properties, relative permeabilities and capillary pressures; isothermal air-water transport; and heat and mass transfer with noncondensible gas. The second component, Dual-Porosity and Double-Porosity/Double-Permeability Formulation, is designed for problems dominated by fracture flow. Another component, The Solute-Transport Models, includes both a reactive-transport model that simulates transport of multiple solutes with chemical reaction and a particle-tracking model. Finally, the component, Constitutive Relationships, deals with pressure- and temperature-dependent fluid/air/gas properties, relative permeabilities and capillary pressures, stress dependencies, and reactive and sorbing solutes. Each of these components is discussed in detail, including purpose, assumptions and limitations, derivation, applications, numerical method type, derivation of numerical model, location in the FEHM code flow, numerical stability and accuracy, and alternative approaches to modeling the component
Algorithms and computer codes for atomic and molecular quantum scattering theory
Energy Technology Data Exchange (ETDEWEB)
Thomas, L. (ed.)
1979-01-01
This workshop has succeeded in bringing up 11 different coupled equation codes on the NRCC computer, testing them against a set of 24 different test problems and making them available to the user community. These codes span a wide variety of methodologies, and factors of up to 300 were observed in the spread of computer times on specific problems. A very effective method was devised for examining the performance of the individual codes in the different regions of the integration range. Many of the strengths and weaknesses of the codes have been identified. Based on these observations, a hybrid code has been developed which is significantly superior to any single code tested. Thus, not only have the original goals been fully met, the workshop has resulted directly in an advancement of the field. All of the computer programs except VIVS are available upon request from the NRCC. Since an improved version of VIVS is contained in the hybrid program, VIVAS, it was not made available for distribution. The individual program LOGD is, however, available. In addition, programs which compute the potential energy matrices of the test problems are also available. The software library names for Tests 1, 2 and 4 are HEH2, LICO, and EN2, respectively.
The computer code SEURBNUK/EURDYN (Release 1). Input and output specification
International Nuclear Information System (INIS)
Broadhouse, B.J.; Yerkess, A.
1986-05-01
SEURBNUK/EURODYN is an extension of SEURBNUK-2, a two dimensional, axisymmetric, Eulerian, finite element containment code in which the finite difference thin shell treatment is replaced by a finite element calculation for both thin and thick structures. These codes are designed to model the hydrodynamic development in time of a hypothetical core disruptive accident (HCDA) in a fast breeder reactor. This manual describes the input data specifications needed for the execution of SEURBNUK/EURDYN calculations, with information on output facilities, and aid to users to avoid some common difficulties. (UK)
Energy Technology Data Exchange (ETDEWEB)
Carbajo, Juan (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin, Madison, WI); Schmidt, Rodney Cannon; Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Ludewig, Hans (Brookhaven National Laboratory, Upton, NY); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache %3CU%2B2013%3E CEA, France)
2011-06-01
This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the
Development of computing code system for level 3 PSA
Energy Technology Data Exchange (ETDEWEB)
Jeong, Jong Tae; Yu, Dong Han; Kim, Seung Hwan
1997-07-01
Among the various research areas of the level 3 PSA, the effect of terrain on the transport of radioactive material was investigated through wind tunnel experiment. These results will give a physical insight in the development of a new dispersion model. Because there are some discrepancies between the results from Gaussian plume model and those from field test, the effect of terrain on the atmospheric dispersion was investigated by using CTDMPLUS code. Through this study we find that the model which can treat terrain effect is essential in the atmospheric dispersion of radioactive materials and the CTDMPLUS model can be used as a useful tool. And it is suggested that modification of a model and experimental study should be made through the continuous effort. The health effect assessment near the Yonggwang site by using IPE (Individual plant examination) results and its site data was performed. The health effect assessment is an important part of consequence analysis of a nuclear power plant site. The MACCS was used in the assessment. Based on the calculation of CCDF for each risk measure, it is shown that CCDF has a slow slope and thus wide probability distribution in cases of early fatality, early injury, total early fatality risk, and total weighted early fatality risk. And in cases of cancer fatality and population dose within 48km and 80km, the CCDF curve have a steep slope and thus narrow probability distribution. The establishment of methodologies for necessary models for consequence analysis resulting form a server accident in the nuclear power plant was made and a program for consequence analysis was developed. The models include atmospheric transport and diffusion, calculation of exposure doses for various pathways, and assessment of health effects and associated risks. Finally, the economic impact resulting form an accident in a nuclear power plant was investigated. In this study, estimation models for each cost terms that considered in economic
Barendregt, W; Bekker, M M
2006-08-01
This article describes the development and assessment of a coding scheme for finding both usability and fun problems through observations of young children playing computer games during user tests. The proposed coding scheme is based on an existing list of breakdown indication types of the detailed video analysis method (DEVAN). This method was developed to detect usability problems in task-based products for adults. However, the new coding scheme for children's computer games takes into account that in games, fun, in addition to usability, is an important factor and that children behave differently from adults. Therefore, the proposed coding scheme uses 8 of the 14 original breakdown indications and has 7 new indications. The article first discusses the development of the new coding scheme. Subsequently, the article describes the reliability assessment of the coding scheme. The any-two agreement measure of 38.5% shows that thresholds for when certain user behavior is worth coding will be different for different evaluators. However, the any-two agreement of .92 for a fixed list of observation points shows that the distinction between the available codes is clear to most evaluators. Finally, a pilot study shows that training can increase any-two agreement considerably by decreasing the number of unique observations, in comparison with the number of agreed upon observations.
FIRAC - a computer code to predict fire accident effects in nuclear facilities
International Nuclear Information System (INIS)
Bolstad, J.W.; Foster, R.D.; Gregory, W.S.
1983-01-01
FIRAC is a medium-sized computer code designed to predict fire-induced flows, temperatures, and material transport within the ventilating systems and other airflow pathways in nuclear-related facilities. The code is designed to analyze the behavior of interconnected networks of rooms and typical ventilation system components. This code is one in a family of computer codes that is designed to provide improved methods of safety analysis for the nuclear industry. The structure of this code closely follows that of the previously developed TVENT and EVENT codes. Because a lumped-parameter formulation is used, this code is particularly suitable for calculating the effects of fires in the far field (that is, in regions removed from the fire compartment), where the fire may be represented parametrically. However, a fire compartment model to simulate conditions in the enclosure is included. This model provides transport source terms to the ventilation system that can affect its operation and in turn affect the fire. A basic material transport capability that features the effects of convection, deposition, entrainment, and filtration of material is included. The interrelated effects of filter plugging, heat transfer, gas dynamics, and material transport are taken into account. In this paper the authors summarize the physical models used to describe the gas dynamics, material transport, and heat transfer processes. They also illustrate how a typical facility is modeled using the code
A computer code for calculations in the algebraic collective model of the atomic nucleus
Welsh, T. A.; Rowe, D. J.
2014-01-01
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 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 functi...
UCODE, a computer code for universal inverse modeling
Poeter, Eileen P.; Hill, Mary C.
1999-05-01
This article presents the US Geological Survey computer program UCODE, which was developed in collaboration with the US Army Corps of Engineers Waterways Experiment Station and the International Ground Water Modeling Center of the Colorado School of Mines. UCODE performs inverse modeling, posed as a parameter-estimation problem, using nonlinear regression. Any application model or set of models can be used; the only requirement is that they have numerical (ASCII or text only) input and output files and that the numbers in these files have sufficient significant digits. Application models can include preprocessors and postprocessors as well as models related to the processes of interest (physical, chemical and so on), making UCODE extremely powerful for model calibration. Estimated parameters can be defined flexibly with user-specified functions. Observations to be matched in the regression can be any quantity for which a simulated equivalent value can be produced, thus simulated equivalent values are calculated using values that appear in the application model output files and can be manipulated with additive and multiplicative functions, if necessary. Prior, or direct, information on estimated parameters also can be included in the regression. The nonlinear regression problem is solved by minimizing a weighted least-squares objective function with respect to the parameter values using a modified Gauss-Newton method. Sensitivities needed for the method are calculated approximately by forward or central differences and problems and solutions related to this approximation are discussed. Statistics are calculated and printed for use in (1) diagnosing inadequate data or identifying parameters that probably cannot be estimated with the available data, (2) evaluating estimated parameter values, (3) evaluating the model representation of the actual processes and (4) quantifying the uncertainty of model simulated values. UCODE is intended for use on any computer operating
Matrix element method for high performance computing platforms
Grasseau, G.; Chamont, D.; Beaudette, F.; Bianchini, L.; Davignon, O.; Mastrolorenzo, L.; Ochando, C.; Paganini, P.; Strebler, T.
2015-12-01
Lot of efforts have been devoted by ATLAS and CMS teams to improve the quality of LHC events analysis with the Matrix Element Method (MEM). Up to now, very few implementations try to face up the huge computing resources required by this method. We propose here a highly parallel version, combining MPI and OpenCL, which makes the MEM exploitation reachable for the whole CMS datasets with a moderate cost. In the article, we describe the status of two software projects under development, one focused on physics and one focused on computing. We also showcase their preliminary performance obtained with classical multi-core processors, CUDA accelerators and MIC co-processors. This let us extrapolate that with the help of 6 high-end accelerators, we should be able to reprocess the whole LHC run 1 within 10 days, and that we have a satisfying metric for the upcoming run 2. The future work will consist in finalizing a single merged system including all the physics and all the parallelism infrastructure, thus optimizing implementation for best hardware platforms.
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.
ABAQUS-EPGEN: a general-purpose finite-element code. Volume 1. User's manual
International Nuclear Information System (INIS)
Hibbitt, H.D.; Karlsson, B.I.; Sorensen, E.P.
1982-10-01
This document is the User's Manual for ABAQUS/EPGEN, a general purpose finite element computer program, designed specifically to serve advanced structural analysis needs. The program contains very general libraries of elements, materials and analysis procedures, and is highly modular, so that complex combinations of features can be put together to model physical problems. The program is aimed at production analysis needs, and for this purpose aspects such as ease-of-use, reliability, flexibility and efficiency have received maximum attention. The input language is designed to make it straightforward to describe complicated models; the analysis procedures are highly automated with the program choosing time or load increments based on user supplied tolerances and controls; and the program offers a wide range of post-processing options for display of the analysis results
PANAMA. A computer code to predict TRISO particle failure under accident conditions
International Nuclear Information System (INIS)
Verfondern, K.; Nabielek, H.
1985-02-01
The computer code PANAMA and its underlying modeling assumptions are presented. The models are based on independent measurements of the properties of TRISO particles with a SiC interlayer. Essential features are the calculation of internal gas pressure, of coating strength and its decrease during irradiation and its weakening due to fission product interaction during accidents. At very high temperatures, particle life is determined by SiC thermal decomposition. Good comparison is obtained in the temperature range 1600 - 2500 0 C when applying PANAMA to a wide variation of existing accident simulation experiments with spherical fuel elements. At lower temperatures, PANAMA tends to be over-conservative. Predictions of particle failure during the depressurized accident sequence with the worst temperatures of the 200 MWsub(th) side-by-side Modular Reactor System remain below the level of normal operations. The same holds true for the HTR-500 MWsub(e) accident sequence with the system under pressure. In the depressurized case, however, failure of all particles has to be expected after approximately 100 hours in the least favourable core position. (orig.) [de
Mathematical model and computer code for coated particles performance at normal operating conditions
International Nuclear Information System (INIS)
Golubev, I.; Kadarmetov, I.; Makarov, V.
2002-01-01
Computer modeling of thermo-mechanical behavior of coated particles during operating both at normal and off-normal conditions has a very significant role particularly on a stage of new reactors development. In Russia a big experience has been accumulated on fabrication and reactor tests of CP and fuel elements with UO 2 kernels. However, this experience cannot be using in full volume for development of a new reactor installation GT-MHR. This is due to very deep burn-up of the fuel based on plutonium oxide (up to 70% fima). Therefore the mathematical modeling of CP thermal-mechanical behavior and failure prediction becomes particularly important. The authors have a clean understanding that serviceability of fuel with high burn-ups are defined not only by thermo-mechanics, but also by structured changes in coating materials, thermodynamics of chemical processes, 'amoeba-effect', formation CO etc. In the report the first steps of development of integrate code for numerical modeling of coated particles behavior and some calculating results concerning the influence of various design parameters on fuel coated particles endurance for GT-MHR normal operating conditions are submitted. A failure model is developed to predict the fraction of TRISO-coated particles. In this model it is assumed that the failure of CP depends not only on probability of SiC-layer fracture but also on the PyC-layers damage. The coated particle is considered as a uniform design. (author)
KC-A Kinectic computer code for investigation of parametric plasma instabilities
International Nuclear Information System (INIS)
Olshansky, V.
1995-07-01
In the frame of a joint research program of the Institute of Plasma Physics of the NationaI Science Center 'Kharkov Institute of Physics and Technology' (Kh IPT), Ukraine, and the plasma physics group of the Austrian Research Center Seibersdorf (FZS) a kinetic computer code with the acronym KC for investigation of paramarametric plasma instabilities has been implemented at the computer facilities of FZS as a starting point for further research in this field. This code based on a macroparticle technique is appropriate for studying the evolution of instabilities in a turbulent plasma including saturation. The results can be of interest for heating of tokamaks of the next generation, i.g. ITER. The present report describes the underlying physical models and numerical methods as well as the code structure and how to use the code as a reference of forthcoming joint papers. (author)
Vectorization of nuclear codes on FACOM 230-75 APU computer
International Nuclear Information System (INIS)
Harada, Hiroo; Higuchi, Kenji; Ishiguro, Misako; Tsutsui, Tsuneo; Fujii, Minoru
1983-02-01
To provide for the future usage of supercomputer, we have investigated the vector processing efficiency of nuclear codes which are being used at JAERI. The investigation is performed by using FACOM 230-75 APU computer. The codes are CITATION (3D neutron diffusion), SAP5 (structural analysis), CASCMARL (irradiation damage simulation). FEM-BABEL (3D neutron diffusion by FEM), GMSCOPE (microscope simulation). DWBA (cross section calculation at molecular collisions). A new type of cell density calculation for particle-in-cell method is also investigated. For each code we have obtained a significant speedup which ranges from 1.8 (CASCMARL) to 7.5 (GMSCOPE), respectively. We have described in this report the running time dynamic profile analysis of the codes, numerical algorithms used, program restructuring for the vectorization, numerical experiments of the iterative process, vectorized ratios, speedup ratios on the FACOM 230-75 APU computer, and some vectorization views. (author)
Compendium of computer codes for the researcher in magnetic fusion energy
Energy Technology Data Exchange (ETDEWEB)
Porter, G.D. (ed.)
1989-03-10
This is a compendium of computer codes, which are available to the fusion researcher. It is intended to be a document that permits a quick evaluation of the tools available to the experimenter who wants to both analyze his data, and compare the results of his analysis with the predictions of available theories. This document will be updated frequently to maintain its usefulness. I would appreciate receiving further information about codes not included here from anyone who has used them. The information required includes a brief description of the code (including any special features), a bibliography of the documentation available for the code and/or the underlying physics, a list of people to contact for help in running the code, instructions on how to access the code, and a description of the output from the code. Wherever possible, the code contacts should include people from each of the fusion facilities so that the novice can talk to someone ''down the hall'' when he first tries to use a code. I would also appreciate any comments about possible additions and improvements in the index. I encourage any additional criticism of this document. 137 refs.
International Nuclear Information System (INIS)
Keshavarz, Mohammad Hossein; Gharagheizi, Farhad; Shokrolahi, Arash; Zakinejad, Sajjad
2012-01-01
Highlights: ► A novel method is introduced for desk calculation of toxicity of benzoic acid derivatives. ► There is no need to use QSAR and QSTR methods, which are based on computer codes. ► The predicted results of 58 compounds are more reliable than those predicted by QSTR method. ► The present method gives good predictions for further 324 benzoic acid compounds. - Abstract: Most of benzoic acid derivatives are toxic, which may cause serious public health and environmental problems. Two novel simple and reliable models are introduced for desk calculations of the toxicity of benzoic acid compounds in mice via oral LD 50 with more reliance on their answers as one could attach to the more complex outputs. They require only elemental composition and molecular fragments without using any computer codes. The first model is based on only the number of carbon and hydrogen atoms, which can be improved by several molecular fragments in the second model. For 57 benzoic compounds, where the computed results of quantitative structure–toxicity relationship (QSTR) were recently reported, the predicted results of two simple models of present method are more reliable than QSTR computations. The present simple method is also tested with further 324 benzoic acid compounds including complex molecular structures, which confirm good forecasting ability of the second model.
COMMUNICATIVE ASPECTS OF MULTILINGUAL CODE SWITCHING IN COMPUTER-MEDIATED COMMUNICATION
Pilar Caparas; Leah Gustilo
2017-01-01
The quintessential role of language has been punctiliously studied relative to intercultural communication, cultural heritage, social development, education, identity construction and many more domains. One forum wherein language is investigated is the Computer-mediated Communication (CMC) which provides a fertile ground for linguistic and sociolinguistic analyses. The present study aims at investigating the preferred codes used in code switching (CS), functions of CS, and the motives of user...
Fault-tolerant quantum computation with asymmetric Bacon-Shor codes
Brooks, Peter; Preskill, John
2013-03-01
We develop a scheme for fault-tolerant quantum computation based on asymmetric Bacon-Shor codes, which works effectively against highly biased noise dominated by dephasing. We find the optimal Bacon-Shor block size as a function of the noise strength and the noise bias, and estimate the logical error rate and overhead cost achieved by this optimal code. Our fault-tolerant gadgets, based on gate teleportation, are well suited for hardware platforms with geometrically local gates in two dimensions.
Computer code 'WAGEN' for the generation of artificial earthquake ground motions
International Nuclear Information System (INIS)
Fujita, Shigeki; Baba, Osamu; Suzuki, Takehiro.
1988-10-01
WAGEN is a computer code for the generation of artificial Earthquake Ground Motions utilized for the aseismic design of nuclear power facilities. WAGEN is developed on the basis of the procedure proposed by Dr. Ohsaki. An artificial earthquake ground motion generated in such a way that the response spectrum shows good fitting to the design response spectrum. This report presents main features, numerical procedure, manual of the code and an example of usage. (author)
SAMDIST A Computer Code for Calculating Statistical Distributions for R-Matrix Resonance Parameters
Leal, L C
1995-01-01
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.
The development of an intelligent interface to a computational fluid dynamics flow-solver code
Williams, Anthony D.
1988-01-01
Researchers at NASA Lewis are currently developing an 'intelligent' interface to aid in the development and use of large, computational fluid dynamics flow-solver codes for studying the internal fluid behavior of aerospace propulsion systems. This paper discusses the requirements, design, and implementation of an intelligent interface to Proteus, a general purpose, three-dimensional, Navier-Stokes flow solver. The interface is called PROTAIS to denote its introduction of artificial intelligence (AI) concepts to the Proteus code.
POPCYCLE: a computer code for calculating nuclear and fossil plant levelized life-cycle power costs
International Nuclear Information System (INIS)
Hardie, R.W.
1982-02-01
POPCYCLE, a computer code designed to calculate levelized life-cycle power costs for nuclear and fossil electrical generating plants is described. Included are (1) derivations of the equations and a discussion of the methodology used by POPCYCLE, (2) a description of the input required by the code, (3) a listing of the input for a sample case, and (4) the output for a sample case
International Nuclear Information System (INIS)
Mann, F.M.
1998-01-01
The Tank Waste Remediation System (TWRS) is responsible for the safe storage, retrieval, and disposal of waste currently being held in 177 underground tanks at the Hanford Site. In order to successfully carry out its mission, TWRS must perform environmental analyses describing the consequences of tank contents leaking from tanks and associated facilities during the storage, retrieval, or closure periods and immobilized low-activity tank waste contaminants leaving disposal facilities. Because of the large size of the facilities and the great depth of the dry zone (known as the vadose zone) underneath the facilities, sophisticated computer codes are needed to model the transport of the tank contents or contaminants. This document presents the code selection criteria for those vadose zone analyses (a subset of the above analyses) where the hydraulic properties of the vadose zone are constant in time the geochemical behavior of the contaminant-soil interaction can be described by simple models, and the geologic or engineered structures are complicated enough to require a two-or three dimensional model. Thus, simple analyses would not need to use the fairly sophisticated codes which would meet the selection criteria in this document. Similarly, those analyses which involve complex chemical modeling (such as those analyses involving large tank leaks or those analyses involving the modeling of contaminant release from glass waste forms) are excluded. The analyses covered here are those where the movement of contaminants can be relatively simply calculated from the moisture flow. These code selection criteria are based on the information from the low-level waste programs of the US Department of Energy (DOE) and of the US Nuclear Regulatory Commission as well as experience gained in the DOE Complex in applying these criteria. Appendix table A-1 provides a comparison between the criteria in these documents and those used here. This document does not define the models (that
Computer codes used in the calculation of high-temperature thermodynamic properties of sodium
Energy Technology Data Exchange (ETDEWEB)
Fink, J.K.
1979-12-01
Three computer codes - SODIPROP, NAVAPOR, and NASUPER - were written in order to calculate a self-consistent set of thermodynamic properties for saturated, subcooled, and superheated sodium. These calculations incorporate new critical parameters (temperature, pressure, and density) and recently derived single equations for enthalpy and vapor pressure. The following thermodynamic properties have been calculated in these codes: enthalpy, heat capacity, entropy, vapor pressure, heat of vaporization, density, volumetric thermal expansion coefficient, compressibility, and thermal pressure coefficient. In the code SODIPROP, these properties are calculated for saturated and subcooled liquid sodium. Thermodynamic properties of saturated sodium vapor are calculated in the code NAVAPOR. The code NASUPER calculates thermodynamic properties for super-heated sodium vapor only for low (< 1644 K) temperatures. No calculations were made for the supercritical region.
Computer codes used in the calculation of high-temperature thermodynamic properties of sodium
International Nuclear Information System (INIS)
Fink, J.K.
1979-12-01
Three computer codes - SODIPROP, NAVAPOR, and NASUPER - were written in order to calculate a self-consistent set of thermodynamic properties for saturated, subcooled, and superheated sodium. These calculations incorporate new critical parameters (temperature, pressure, and density) and recently derived single equations for enthalpy and vapor pressure. The following thermodynamic properties have been calculated in these codes: enthalpy, heat capacity, entropy, vapor pressure, heat of vaporization, density, volumetric thermal expansion coefficient, compressibility, and thermal pressure coefficient. In the code SODIPROP, these properties are calculated for saturated and subcooled liquid sodium. Thermodynamic properties of saturated sodium vapor are calculated in the code NAVAPOR. The code NASUPER calculates thermodynamic properties for super-heated sodium vapor only for low (< 1644 K) temperatures. No calculations were made for the supercritical region
Abstracts of digital computer code packages. Assembled by the Radiation Shielding Information Center
International Nuclear Information System (INIS)
McGill, B.; Maskewitz, B.F.; Anthony, C.M.; Comolander, H.E.; Hendrickson, H.R.
1976-01-01
The term ''code package'' is used to describe a miscellaneous grouping of materials which, when interpreted in connection with a digital computer, enables the scientist--user to solve technical problems in the area for which the material was designed. In general, a ''code package'' consists of written material--reports, instructions, flow charts, listings of data, and other useful material and IBM card decks (or, more often, a reel of magnetic tape) on which the source decks, sample problem input (including libraries of data) and the BCD/EBCDIC output listing from the sample problem are written. In addition to the main code, and any available auxiliary routines are also included. The abstract format was chosen to give to a potential code user several criteria for deciding whether or not he wishes to request the code package
A first accident simulation for Angra-1 power plant using the ALMOD computer code
International Nuclear Information System (INIS)
Camargo, C.T.M.
1981-02-01
The acquisition of the Almod computer code from GRS-Munich to CNEN has permited doing calculations of transients in PWR nuclear power plants, in which doesn't occur loss of coolant. The implementation of the german computer code Almod and its application in the calculation of Angra-1, a nuclear power plant different from the KWU power plants, demanded study and models adaptation; and due to economic reasons simplifications and optimizations were necessary. The first results define the analytical potential of the computer code, confirm the adequacy of the adaptations done and provide relevant conclusions about the Angra-1 safety analysis, showing at the same time areas in which the model can be applied or simply improved. (Author) [pt
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.
SCALE: A modular code system for performing standardized computer analyses for licensing evaluation
International Nuclear Information System (INIS)
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
Modifications in the UNIMUG computer code (Addendum to technical note EAV/NT--003/80)
International Nuclear Information System (INIS)
Alcantara, H.G. de; Rosa, M.A.P.
1983-01-01
The modifications done in the UNIMUG computer code are described. The UNIMUG-VERSION 3 has options for condensating macroscopic cross sections in regions and microscopic cross sections of isotopes contained in each region. The code also calculates: spectral index, ratio between two specificated reactions of chosen isotopes in indicated groups and regions, average buckling by region/group, by region, reactor buckling and average flux of reactor. The code has no restrictions for number of regions, number of groups, number of intervals, and table size of macroscopic and microscopic cross sections. (M.C.K.) [pt
International Nuclear Information System (INIS)
Kolev, N.I.
1991-12-01
This report describes the input and output ov IVA3 computer code and the procedure how to compile, link, and run the code. The common blocs recorded for restarts files and post processing are described in detail as well as the IVA3 interface for thermodynamic and thermo physical properties. Some recommendations for the input preparation together with some detailed comments on some architectural and functional features of the code are given in order to give some insight of the caused actions by changing some control parameters. (orig.) [de
A computer code 'BEAM' for the ion optics calculation of the JAERI tandem accelerator system
International Nuclear Information System (INIS)
Kikuchi, Shiroh; Takeuchi, Suehiro
1987-11-01
The computer code BEAM is described, together with an outline of the formalism used for the ion optics calculation. The purpose of the code is to obtain the optimum parameters of devices, with which the ion beam is transported through the system without losses. The procedures of the calculation, especially those of searching for the parameters of quadrupole lenses, are discussed in detail. The flow of the code is illustrated as a whole and its constituent subroutines are explained individually. A few resultant beam trajectories and the parameters used to obtain them are shown as examples. (author)
Computing the effects of a contained sodium sheet fire: The 'FEUNA' code
International Nuclear Information System (INIS)
Duverger De Cuy, G.
1979-01-01
FEUNA is a computer code developed to calculate the thermodynamic effects of a sodium fire in a ventilated or unventilated containment volume. Developed jointly by the CEA/DSN and Novatome, the FEUNA code involves two oxide formation reactions, aerosol generation and deposits, heat transfer by convection, conduction and radiation, gas inflow and outflow through the ventilation system and the relief valves. The code was validated by comparing calculated values with the results of an actual sodium fire in a 400m 3 caisson. (author)
PyLith: A Finite-Element Code for Modeling Quasi-Static and Dynamic Crustal Deformation
Aagaard, B.; Williams, C. A.; Knepley, M. G.
2011-12-01
We have developed open-source finite-element software for 2-D and 3-D dynamic and quasi-static modeling of crustal deformation. This software, PyLith (current release is version 1.6) can be used for quasi-static viscoelastic modeling, dynamic spontaneous rupture and/or ground-motion modeling. Unstructured and structured finite-element discretizations allow for spatial scales ranging from tens of meters to hundreds of kilometers with temporal scales in dynamic problems ranging from milliseconds to minutes and temporal scales in quasi-static problems ranging from minutes to thousands of years. PyLith development is part of the NSF funded Computational Infrastructure for Geodynamics (CIG) and the software runs on a wide variety of platforms (laptops, workstations, and Beowulf clusters). Binaries (Linux, Darwin, and Windows systems) and source code are available from geodynamics.org. PyLith uses a suite of general, parallel, graph data structures called Sieve for storing and manipulating finite-element meshes. This permits use of a variety of 2-D and 3-D cell types including triangles, quadrilaterals, hexahedra, and tetrahedra. Current PyLith features include prescribed fault ruptures with multiple earthquakes and aseismic creep, spontaneous fault ruptures with a variety of fault constitutive models, time-dependent Dirichlet and Neumann boundary conditions, absorbing boundary conditions, time-dependent point forces, and gravitational body forces. PyLith supports infinitesimal and small strain formulations for linear elastic rheologies, linear and generalized Maxwell viscoelastic rheologies, power-law viscoelastic rheologies, and Drucker-Prager elastoplastic rheologies. Current software development focuses on coupling quasi-static and dynamic simulations to resolve multi-scale deformation across the entire seismic cycle and the coupling of elasticity to heat and/or fluid flow.
CEDNBR: a computer code for transient thermal margin analysis of a reactor core
International Nuclear Information System (INIS)
Shesler, A.T.; Lehmann, C.R.
1976-09-01
The report describes the CEDNBR computer code. This code was developed for the transient thermal analysis of a pressurized water reactor core or a critical heat flux test. Included are the code structure, conservation equations, and correlations utilized by CEDNBR. The methods of modelling a reactor core and hot channel and a CHF test are presented. Comparisons of CEDNBR calculations are made with both empirical pressure loss data and simulated loss of flow test data. The code solves the one-dimensional conservation of mass, energy, and momentum equations and the equation of state for the fluid for either steady-state or transient conditions. Tabular time dependent functions of inlet temperatures, pressure, mass velocity, axial heat flux distributions, normalized heat flux, radial peaking factors, and incremental mixing factors are required input to the code. Transient effects are included in the calculation of enthalpy rise and fluid properties. The Departure from Nucleate Boiling Ratio (DNBR) is calculated by applying a Critical Heat Flux (CHF) correlation to the computed local fluid properties. A code user's guide is provided for preparing input to the code. In addition, descriptions of the sub-routines used by CEDNBR are given
Computer code TOBUNRAD for PWR fuel bundle heat-up calculations
International Nuclear Information System (INIS)
Shimooke, Takanori; Yoshida, Kazuo
1979-05-01
The computer code TOBUNRAD developed is for analysis of ''fuel-bundle'' heat-up phenomena in a loss-of-coolant accident of PWR. The fuel bundle consists of fuel pins in square lattice; its behavior is different from that of individual pins during heat-up. The code is based on the existing TOODEE2 code which analyzes heat-up phenomena of single fuel pins, so that the basic models of heat conduction and transfer and coolant flow are the same as the TOODEE2's. In addition to the TOODEE2 features, unheated rods are modeled and radiation heat loss is considered between fuel pins, a fuel pin and other heat sinks. The TOBUNRAD code is developed by a new FORTRAN technique which makes it possible to interrupt a flow of program controls wherever desired, thereby attaching several subprograms to the main code. Users' manual for TOBUNRAD is presented: The basic program-structure by interruption method, physical and computational model in each sub-code, usage of the code and sample problems. (author)
Accuracy assessment of a new Monte Carlo based burnup computer code
International Nuclear Information System (INIS)
El Bakkari, B.; ElBardouni, T.; Nacir, B.; ElYounoussi, C.; Boulaich, Y.; Meroun, O.; Zoubair, M.; Chakir, E.
2012-01-01
Highlights: ► A new burnup code called BUCAL1 was developed. ► BUCAL1 uses the MCNP tallies directly in the calculation of the isotopic inventories. ► Validation of BUCAL1 was done by code to code comparison using VVER-1000 LEU Benchmark Assembly. ► Differences from BM value were found to be ± 600 pcm for k ∞ and ±6% for the isotopic compositions. ► The effect on reactivity due to the burnup of Gd isotopes is well reproduced by BUCAL1. - Abstract: This study aims to test for the suitability and accuracy of a new home-made Monte Carlo burnup code, called BUCAL1, by investigating and predicting the neutronic behavior of a “VVER-1000 LEU Assembly Computational Benchmark”, at lattice level. BUCAL1 uses MCNP tally information directly in the computation; this approach allows performing straightforward and accurate calculation without having to use the calculated group fluxes to perform transmutation analysis in a separate code. ENDF/B-VII evaluated nuclear data library was used in these calculations. Processing of the data library is performed using recent updates of NJOY99 system. Code to code comparisons with the reported Nuclear OECD/NEA results are presented and analyzed.
Energy Technology Data Exchange (ETDEWEB)
Chan, M.K.; Ballinger, M.Y.; Owczarski, P.C.
1989-02-01
This manual describes the technical bases and use of the computer code FIRIN. This code was developed to estimate the source term release of smoke and radioactive particles from potential fires in nuclear fuel cycle facilities. FIRIN is a product of a broader study, Fuel Cycle Accident Analysis, which Pacific Northwest Laboratory conducted for the US Nuclear Regulatory Commission. The technical bases of FIRIN consist of a nonradioactive fire source term model, compartment effects modeling, and radioactive source term models. These three elements interact with each other in the code affecting the course of the fire. This report also serves as a complete FIRIN user's manual. Included are the FIRIN code description with methods/algorithms of calculation and subroutines, code operating instructions with input requirements, and output descriptions. 40 refs., 5 figs., 31 tabs.
FEMAXI-III, a computer code for fuel rod performance analysis
International Nuclear Information System (INIS)
Ito, K.; Iwano, Y.; Ichikawa, M.; Okubo, T.
1983-01-01
This paper presents a method of fuel rod thermal-mechanical performance analysis used in the FEMAXI-III code. The code incorporates the models describing thermal-mechanical processes such as pellet-cladding thermal expansion, pellet irradiation swelling, densification, relocation and fission gas release as they affect pellet-cladding gap thermal conductance. The code performs the thermal behavior analysis of a full-length fuel rod within the framework of one-dimensional multi-zone modeling. The mechanical effects including ridge deformation is rigorously analyzed by applying the axisymmetric finite element method. The finite element geometrical model is confined to a half-pellet-height region with the assumption that pellet-pellet interaction is symmetrical. The 8-node quadratic isoparametric ring elements are adopted for obtaining accurate finite element solutions. The Newton-Raphson iteration with an implicit algorithm is applied to perform the analysis of non-linear material behaviors accurately and stably. The pellet-cladding interaction mechanism is exactly treated using the nodal continuity conditions. The code is applicable to the thermal-mechanical analysis of water reactor fuel rods experiencing variable power histories. (orig.)
Validation of Mean Drift Forces Computed with the BEM Code NEMOH
DEFF Research Database (Denmark)
Thomsen, Jonas Bjerg
This report covers a simple investigation of mean drift forces found by use of the boundary element method code NEMOH. The results from NEMOH are compared to analytical results from literature and to numerical values found from the commercial software package WADAM by DNV-GL. The work was conducted...
A finite element method for the computation of transonic flow past airfoils
Eberle, A.
1980-01-01
A finite element method for the computation of the transonic flow with shocks past airfoils is presented using the artificial viscosity concept for the local supersonic regime. Generally, the classic element types do not meet the accuracy requirements of advanced numerical aerodynamics requiring special attention to the choice of an appropriate element. A series of computed pressure distributions exhibits the usefulness of the method.
Computer-aided manufacturing for freeform optical elements by ultraprecision micromilling
Stoebenau, Sebastian; Kleindienst, Roman; Hofmann, Meike; Sinzinger, Stefan
2011-09-01
The successful fabrication of several freeform optical elements by ultraprecision micromilling is presented in this article. We discuss in detail the generation of the tool paths using different variations of a computer-aided manufacturing (CAM) process. Following a classical CAM approach, a reflective beam shaper was fabricated. The approach is based on a solid model calculated by optical design software. As no analytical description of the surface is needed, this procedure is the most general solution for the programming of the tool paths. A second approach is based on the same design data. But instead of a solid model, a higher order polynomial was fitted to the data using computational methods. Taking advantage of the direct programming capabilities of state-of-the-art computerized numerical control units, the mathematics to calculate the polynomial based tool paths on-the-fly during the machining process are implemented in a highly flexible CNC code. As another example for this programming method, the fabrication of a biconic lens from a closed analytical description directly derived from the optical design is shown. We provide details about the different programming methods and the fabrication processes as well as the results of characterizations concerning surface quality and shape accuracy of the freeform optical elements.
Energy Technology Data Exchange (ETDEWEB)
Baes, C.F. III; Sharp, R.D.; Sjoreen, A.L.; Hermann, O.W.
1984-11-01
TERRA is a computer code which calculates concentrations of radionuclides and ingrowing daughters in surface and root-zone soil, produce and feed, beef, and milk from a given deposition rate at any location in the conterminous United States. The code is fully integrated with seven other computer codes which together comprise a Computerized Radiological Risk Investigation System, CRRIS. Output from either the long range (> 100 km) atmospheric dispersion code RETADD-II or the short range (<80 km) atmospheric dispersion code ANEMOS, in the form of radionuclide air concentrations and ground deposition rates by downwind location, serves as input to TERRA. User-defined deposition rates and air concentrations may also be provided as input to TERRA through use of the PRIMUS computer code. The environmental concentrations of radionuclides predicted by TERRA serve as input to the ANDROS computer code which calculates population and individual intakes, exposures, doses, and risks. TERRA incorporates models to calculate uptake from soil and atmospheric deposition on four groups of produce for human consumption and four groups of livestock feeds. During the environmental transport simulation, intermediate calculations of interception fraction for leafy vegetables, produce directly exposed to atmospherically depositing material, pasture, hay, and silage are made based on location-specific estimates of standing crop biomass. Pasture productivity is estimated by a model which considers the number and types of cattle and sheep, pasture area, and annual production of other forages (hay and silage) at a given location. Calculations are made of the fraction of grain imported from outside the assessment area. TERRA output includes the above calculations and estimated radionuclide concentrations in plant produce, milk, and a beef composite by location.
International Nuclear Information System (INIS)
Baes, C.F. III; Sharp, R.D.; Sjoreen, A.L.; Hermann, O.W.
1984-11-01
TERRA is a computer code which calculates concentrations of radionuclides and ingrowing daughters in surface and root-zone soil, produce and feed, beef, and milk from a given deposition rate at any location in the conterminous United States. The code is fully integrated with seven other computer codes which together comprise a Computerized Radiological Risk Investigation System, CRRIS. Output from either the long range (> 100 km) atmospheric dispersion code RETADD-II or the short range (<80 km) atmospheric dispersion code ANEMOS, in the form of radionuclide air concentrations and ground deposition rates by downwind location, serves as input to TERRA. User-defined deposition rates and air concentrations may also be provided as input to TERRA through use of the PRIMUS computer code. The environmental concentrations of radionuclides predicted by TERRA serve as input to the ANDROS computer code which calculates population and individual intakes, exposures, doses, and risks. TERRA incorporates models to calculate uptake from soil and atmospheric deposition on four groups of produce for human consumption and four groups of livestock feeds. During the environmental transport simulation, intermediate calculations of interception fraction for leafy vegetables, produce directly exposed to atmospherically depositing material, pasture, hay, and silage are made based on location-specific estimates of standing crop biomass. Pasture productivity is estimated by a model which considers the number and types of cattle and sheep, pasture area, and annual production of other forages (hay and silage) at a given location. Calculations are made of the fraction of grain imported from outside the assessment area. TERRA output includes the above calculations and estimated radionuclide concentrations in plant produce, milk, and a beef composite by location
Independent verification and validation testing of the FLASH computer code, Versiion 3.0
International Nuclear Information System (INIS)
Martian, P.; Chung, J.N.
1992-06-01
Independent testing of the FLASH computer code, Version 3.0, was conducted to determine if the code is ready for use in hydrological and environmental studies at various Department of Energy sites. This report describes the technical basis, approach, and results of this testing. Verification tests, and validation tests, were used to determine the operational status of the FLASH computer code. These tests were specifically designed to test: correctness of the FORTRAN coding, computational accuracy, and suitability to simulating actual hydrologic conditions. This testing was performed using a structured evaluation protocol which consisted of: blind testing, independent applications, and graduated difficulty of test cases. Both quantitative and qualitative testing was performed through evaluating relative root mean square values and graphical comparisons of the numerical, analytical, and experimental data. Four verification test were used to check the computational accuracy and correctness of the FORTRAN coding, and three validation tests were used to check the suitability to simulating actual conditions. These tests cases ranged in complexity from simple 1-D saturated flow to 2-D variably saturated problems. The verification tests showed excellent quantitative agreement between the FLASH results and analytical solutions. The validation tests showed good qualitative agreement with the experimental data. Based on the results of this testing, it was concluded that the FLASH code is a versatile and powerful two-dimensional analysis tool for fluid flow. In conclusion, all aspects of the code that were tested, except for the unit gradient bottom boundary condition, were found to be fully operational and ready for use in hydrological and environmental studies
International Nuclear Information System (INIS)
McGill, B.L.; Roussin, R.W.; Trubey, D.K.; Maskewitz, B.F.
1980-01-01
The Radiation Shielding Information Center (RSIC), established in 1962 to collect, package, analyze, and disseminate information, computer codes, and data in the area of radiation transport related to fission, is now being utilized to support fusion neutronics technology. The major activities include: (1) answering technical inquiries on radiation transport problems, (2) collecting, packaging, testing, and disseminating computing technology and data libraries, and (3) reviewing literature and operating a computer-based information retrieval system containing material pertinent to radiation transport analysis. The computer codes emphasize methods for solving the Boltzmann equation such as the discrete ordinates and Monte Carlo techniques, both of which are widely used in fusion neutronics. The data packages include multigroup coupled neutron-gamma-ray cross sections and kerma coefficients, other nuclear data, and radiation transport benchmark problem results
Some questions of using coding theory and analytical calculation methods on computers
International Nuclear Information System (INIS)
Nikityuk, N.M.
1987-01-01
Main results of investigations devoted to the application of theory and practice of correcting codes are presented. These results are used to create very fast units for the selection of events registered in multichannel detectors of nuclear particles. Using this theory and analytical computing calculations, practically new combination devices, for example, parallel encoders, have been developed. Questions concerning the creation of a new algorithm for the calculation of digital functions by computers and problems of devising universal, dynamically reprogrammable logic modules are discussed
International Nuclear Information System (INIS)
Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.
1979-07-01
User input data requirements are presented for certain special processors in a nuclear reactor computation system. These processors generally read data in formatted form and generate binary interface data files. Some data processing is done to convert from the user oriented form to the interface file forms. The VENTURE diffusion theory neutronics code and other computation modules in this system use the interface data files which are generated
Steam generator transient studies using a simplified two-fluid computer code
International Nuclear Information System (INIS)
Munshi, P.; Bhatnagar, R.; Ram, K.S.
1985-01-01
A simplified two-fluid computer code has been used to simulate reactor-side (or primary-side) transients in a PWR steam generator. The disturbances are modelled as ramp inputs for pressure, internal energy and mass flow-rate for the primary fluid. The CPU time for a transient duration of 4 s is approx. 10 min on a DEC-1090 computer system. The results are thermodynamically consistent and encouraging for further studies. (author)
1995-09-01
this subprogram defines the relationship bet~een the ·actuator position c louver angle and open area for the end vend of the LB /TS RWE c c...ARMY RESEARCH UBORA TORY Computer Programs for LB {fS Test Design: Technical Description, Usage Instructions and Source Code Listings ARL-MR-260...COVERED Final, Oct 94 - May 95 4. TITLE AND SUBTITLE 5. FUNDING NUMBERS Computer Programs for LB /TS Test Design: Technical Description, Usage
PIC codes for plasma accelerators on emerging computer architectures (GPUS, Multicore/Manycore CPUS)
Vincenti, Henri
2016-03-01
The advent of exascale computers will enable 3D simulations of a new laser-plasma interaction regimes that were previously out of reach of current Petasale computers. However, the paradigm used to write current PIC codes will have to change in order to fully exploit the potentialities of these new computing architectures. Indeed, achieving Exascale computing facilities in the next decade will be a great challenge in terms of energy consumption and will imply hardware developments directly impacting our way of implementing PIC codes. As data movement (from die to network) is by far the most energy consuming part of an algorithm future computers will tend to increase memory locality at the hardware level and reduce energy consumption related to data movement by using more and more cores on each compute nodes (''fat nodes'') that will have a reduced clock speed to allow for efficient cooling. To compensate for frequency decrease, CPU machine vendors are making use of long SIMD instruction registers that are able to process multiple data with one arithmetic operator in one clock cycle. SIMD register length is expected to double every four years. GPU's also have a reduced clock speed per core and can process Multiple Instructions on Multiple Datas (MIMD). At the software level Particle-In-Cell (PIC) codes will thus have to achieve both good memory locality and vectorization (for Multicore/Manycore CPU) to fully take advantage of these upcoming architectures. In this talk, we present the portable solutions we implemented in our high performance skeleton PIC code PICSAR to both achieve good memory locality and cache reuse as well as good vectorization on SIMD architectures. We also present the portable solutions used to parallelize the Pseudo-sepctral quasi-cylindrical code FBPIC on GPUs using the Numba python compiler.
International Nuclear Information System (INIS)
Heeb, C.M.
1991-03-01
The ORIGEN2 computer code is the primary calculational tool for computing isotopic source terms for the Hanford Environmental Dose Reconstruction (HEDR) Project. The ORIGEN2 code computes the amounts of radionuclides that are created or remain in spent nuclear fuel after neutron irradiation and radioactive decay have occurred as a result of nuclear reactor operation. ORIGEN2 was chosen as the primary code for these calculations because it is widely used and accepted by the nuclear industry, both in the United States and the rest of the world. Its comprehensive library of over 1,600 nuclides includes any possible isotope of interest to the HEDR Project. It is important to evaluate the uncertainties expected from use of ORIGEN2 in the HEDR Project because these uncertainties may have a pivotal impact on the final accuracy and credibility of the results of the project. There are three primary sources of uncertainty in an ORIGEN2 calculation: basic nuclear data uncertainty in neutron cross sections, radioactive decay constants, energy per fission, and fission product yields; calculational uncertainty due to input data; and code uncertainties (i.e., numerical approximations, and neutron spectrum-averaged cross-section values from the code library). 15 refs., 5 figs., 5 tabs
Just-in-time compilation-inspired methodology for parallelization of compute intensive java code
International Nuclear Information System (INIS)
Mustafa, G.; Ghani, M.U.
2017-01-01
Compute intensive programs generally consume significant fraction of execution time in a small amount of repetitive code. Such repetitive code is commonly known as hotspot code. We observed that compute intensive hotspots often possess exploitable loop level parallelism. A JIT (Just-in-Time) compiler profiles a running program to identify its hotspots. Hotspots are then translated into native code, for efficient execution. Using similar approach, we propose a methodology to identify hotspots and exploit their parallelization potential on multicore systems. Proposed methodology selects and parallelizes each DOALL loop that is either contained in a hotspot method or calls a hotspot method. The methodology could be integrated in front-end of a JIT compiler to parallelize sequential code, just before native translation. However, compilation to native code is out of scope of this work. As a case study, we analyze eighteen JGF (Java Grande Forum) benchmarks to determine parallelization potential of hotspots. Eight benchmarks demonstrate a speedup of up to 7.6x on an 8-core system. (author)
Just-in-Time Compilation-Inspired Methodology for Parallelization of Compute Intensive Java Code
Directory of Open Access Journals (Sweden)
GHULAM MUSTAFA
2017-01-01
Full Text Available Compute intensive programs generally consume significant fraction of execution time in a small amount of repetitive code. Such repetitive code is commonly known as hotspot code. We observed that compute intensive hotspots often possess exploitable loop level parallelism. A JIT (Just-in-Time compiler profiles a running program to identify its hotspots. Hotspots are then translated into native code, for efficient execution. Using similar approach, we propose a methodology to identify hotspots and exploit their parallelization potential on multicore systems. Proposed methodology selects and parallelizes each DOALL loop that is either contained in a hotspot method or calls a hotspot method. The methodology could be integrated in front-end of a JIT compiler to parallelize sequential code, just before native translation. However, compilation to native code is out of scope of this work. As a case study, we analyze eighteen JGF (Java Grande Forum benchmarks to determine parallelization potential of hotspots. Eight benchmarks demonstrate a speedup of up to 7.6x on an 8-core system
Assessment of a hybrid finite element and finite volume code for turbulent incompressible flows
International Nuclear Information System (INIS)
Xia, Yidong; Wang, Chuanjin; Luo, Hong; Christon, Mark; Bakosi, Jozsef
2016-01-01
Hydra-TH is a hybrid finite-element/finite-volume incompressible/low-Mach flow simulation code based on the Hydra multiphysics toolkit being developed and used for thermal-hydraulics applications. In the present work, a suite of verification and validation (V&V) test problems for Hydra-TH was defined to meet the design requirements of the Consortium for Advanced Simulation of Light Water Reactors (CASL). The intent for this test problem suite is to provide baseline comparison data that demonstrates the performance of the Hydra-TH solution methods. The simulation problems vary in complexity from laminar to turbulent flows. A set of RANS and LES turbulence models were used in the simulation of four classical test problems. Numerical results obtained by Hydra-TH agreed well with either the available analytical solution or experimental data, indicating the verified and validated implementation of these turbulence models in Hydra-TH. Where possible, some form of solution verification has been attempted to identify sensitivities in the solution methods, and suggest best practices when using the Hydra-TH code. -- Highlights: •We performed a comprehensive study to verify and validate the turbulence models in Hydra-TH. •Hydra-TH delivers 2nd-order grid convergence for the incompressible Navier–Stokes equations. •Hydra-TH can accurately simulate the laminar boundary layers. •Hydra-TH can accurately simulate the turbulent boundary layers with RANS turbulence models. •Hydra-TH delivers high-fidelity LES capability for simulating turbulent flows in confined space.
Chen, Jian; Matuttis, Hans-Georg
2013-02-01
We report our experiences with the optimization and parallelization of a discrete element code for convex polyhedra on multi-core machines and introduce a novel variant of the sort-and-sweep neighborhood algorithm. While in theory the whole code in itself parallelizes ideally, in practice the results on different architectures with different compilers and performance measurement tools depend very much on the particle number and optimization of the code. After difficulties with the interpretation of the data for speedup and efficiency are overcome, respectable parallelization speedups could be obtained.
Solving linear systems in FLICA-4, thermohydraulic code for 3-D transient computations
International Nuclear Information System (INIS)
Allaire, G.
1995-01-01
FLICA-4 is a computer code, developed at the CEA (France), devoted to steady state and transient thermal-hydraulic analysis of nuclear reactor cores, for small size problems (around 100 mesh cells) as well as for large ones (more than 100000), on, either standard workstations or vector super-computers. As for time implicit codes, the largest time and memory consuming part of FLICA-4 is the routine dedicated to solve the linear system (the size of which is of the order of the number of cells). Therefore, the efficiency of the code is crucially influenced by the optimization of the algorithms used in assembling and solving linear systems: direct methods as the Gauss (or LU) decomposition for moderate size problems, iterative methods as the preconditioned conjugate gradient for large problems. 6 figs., 13 refs
Validation and uncertainty analysis of the Athlet thermal-hydraulic computer code
International Nuclear Information System (INIS)
Glaeser, H.
1995-01-01
The computer code ATHLET is being developed by GRS as an advanced best-estimate code for the simulation of breaks and transients in Pressurized Water Reactor (PWRs) and Boiling Water Reactor (BWRs) including beyond design basis accidents. A systematic validation of ATHLET is based on a well balanced set of integral and separate effects tests emphasizing the German combined Emergency Core Cooling (ECC) injection system. When using best estimate codes for predictions of reactor plant states during assumed accidents, qualification of the uncertainty in these calculations is highly desirable. A method for uncertainty and sensitivity evaluation has been developed by GRS where the computational effort is independent of the number of uncertain parameters. (author)
Modeling of BWR core meltdown accidents - for application in the MELRPI. MOD2 computer code
Energy Technology Data Exchange (ETDEWEB)
Koh, B R; Kim, S H; Taleyarkhan, R P; Podowski, M Z; Lahey, Jr, R T
1985-04-01
This report summarizes improvements and modifications made in the MELRPI computer code. A major difference between this new, updated version of the code, called MELRPI.MOD2, and the one reported previously, concerns the inclusion of a model for the BWR emergency core cooling systems (ECCS). This model and its computer implementation, the ECCRPI subroutine, account for various emergency injection modes, for both intact and rubblized geometries. Other changes to MELRPI deal with an improved model for canister wall oxidation, rubble bed modeling, and numerical integration of system equations. A complete documentation of the entire MELRPI.MOD2 code is also given, including an input guide, list of subroutines, sample input/output and program listing.
Algorithms and computer codes for atomic and molecular quantum scattering theory. Volume I
International Nuclear Information System (INIS)
Thomas, L.
1979-01-01
The goals of this workshop are to identify which of the existing computer codes for solving the coupled equations of quantum molecular scattering theory perform most efficiently on a variety of test problems, and to make tested versions of those codes available to the chemistry community through the NRCC software library. To this end, many of the most active developers and users of these codes have been invited to discuss the methods and to solve a set of test problems using the LBL computers. The first volume of this workshop report is a collection of the manuscripts of the talks that were presented at the first meeting held at the Argonne National Laboratory, Argonne, Illinois June 25-27, 1979. It is hoped that this will serve as an up-to-date reference to the most popular methods with their latest refinements and implementations
Algorithms and computer codes for atomic and molecular quantum scattering theory. Volume I
Energy Technology Data Exchange (ETDEWEB)
Thomas, L. (ed.)
1979-01-01
The goals of this workshop are to identify which of the existing computer codes for solving the coupled equations of quantum molecular scattering theory perform most efficiently on a variety of test problems, and to make tested versions of those codes available to the chemistry community through the NRCC software library. To this end, many of the most active developers and users of these codes have been invited to discuss the methods and to solve a set of test problems using the LBL computers. The first volume of this workshop report is a collection of the manuscripts of the talks that were presented at the first meeting held at the Argonne National Laboratory, Argonne, Illinois June 25-27, 1979. It is hoped that this will serve as an up-to-date reference to the most popular methods with their latest refinements and implementations.
RAP-2A Computer code for transients analysis in fast reactors
International Nuclear Information System (INIS)
Iftode, I.; Popescu, C.; Turcu, I.; Biro, L.
1975-10-01
The RAP-2A computer code is designed for analyzing thermohydraulic transients and/or steady state problems for large LMFBR cores. Physical and mathematical models, main input-output data, the flow chart of the code and a sample problem are given. RAP-2A calculates the power and the thermoydraulic transients initiated by a flow or reactivity changes, from a normal operating state of the reactor up to core disassembly. In this analysis a representative fuel pin is considered: a one-group space-independent (point) kinetics model to describe the neutron kinetics and a one-dimensional model describing the heat transfer (radial in the fuel and axial in the coolant) are used. Mechanical deformations due to temperature gradient, pressure losses, fuel melting, etc., are also calculated. The code is written in FORTRAN-4 language and is running on a IBM-370/135 computer
Modeling of BWR core meltdown accidents - for application in the MELRPI.MOD2 computer code
International Nuclear Information System (INIS)
Koh, B.R.; Kim, S.H.; Taleyarkhan, R.P.; Podowski, M.Z.; Lahey, R.T. Jr.
1985-04-01
This report summarizes improvements and modifications made in the MELRPI computer code. A major difference between this new, updated version of the code, called MELRPI.MOD2, and the one reported previously, concerns the inclusion of a model for the BWR emergency core cooling systems (ECCS). This model and its computer implementation, the ECCRPI subroutine, account for various emergency injection modes, for both intact and rubblized geometries. Other changes to MELRPI deal with an improved model for canister wall oxidation, rubble bed modeling, and numerical integration of system equations. A complete documentation of the entire MELRPI.MOD2 code is also given, including an input guide, list of subroutines, sample input/output and program listing
A code to compute borehole fluid conductivity profiles with multiple feed points
International Nuclear Information System (INIS)
Hale, F.V.; Tsang, C.F.
1988-03-01
It is of much current interest to determine the flow characteristics of fractures intersecting a wellbore in order to understand the hydrologic behavior of fractured rocks. Often inflow from these fractures into the wellbore is at very low rates. A new procedure has been proposed and a corresponding method of analysis developed to obtain fracture inflow parameters from a time sequence of electric conductivity logs of the borehole fluid. The present report is a companion document to NTB--88-13 giving the details of equations and computer code used to compute borehole fluid conductivity distributions. Verification of the code used and a listing of the code are also given. (author) 9 refs., 5 figs., 7 tabs
The computation of linear triangular matrices in the finite element ...
African Journals Online (AJOL)
An algorithm is developed for generating the system matrices for the Finite Element Method of solving some classes of second order partial differential equations problems using the linear triangular elements. This algorithm reduces the complexity normally associated with the finite element approximation and makes the ...
Garneli, Varvara; Chorianopoulos, Konstantinos
2018-01-01
Various aspects of computational thinking (CT) could be supported by educational contexts such as simulations and video-games construction. In this field study, potential differences in student motivation and learning were empirically examined through students' code. For this purpose, we performed a teaching intervention that took place over five…
INTERTRAN 2 - A computer code for calculating the risk from transportation of radioactive materials
International Nuclear Information System (INIS)
Ericsson, A.M.; Jaernry, C.
1993-01-01
In this paper a description of IAEA Coordinated Research Program (CRP) dealing with the updating of the computer code INTERTRAN is given. The paper includes a summary of the work performed by several member states within the CRP as well as gives a description of the final product that will be presented to the IAEA. (J.P.N.)
International Nuclear Information System (INIS)
Falikov, A.A.; Vakhrushev, V.V.; Kuul, V.S.; Samoilov, O.B.; Tarasov, G.I.
1997-01-01
The paper briefly reviews the specific thermal-hydraulic problems for AST-type NHRs, the experimental investigations that have been carried out in the RF, and the design procedures and computer codes used for AST-500 thermohydraulic characteristics and safety validation. (author). 13 refs, 10 figs, 1 tab
ALICE-87 (Livermore). Precompound Nuclear Model Code. Version for Personal Computer IBM/AT
International Nuclear Information System (INIS)
Blann, M.
1988-05-01
The precompound nuclear model code ALICE-87 from the Lawrence Livermore National Laboratory (USA) was implemented for use on personal computer. It is available on a set of high density diskettes from the Data Bank of Nuclear Energy Agency (Saclay) and the IAEA Nuclear Data Section. (author). Refs and figs
Independent validation testing of the FLAME computer code, Version 1.0
International Nuclear Information System (INIS)
Martian, P.; Chung, J.N.
1992-07-01
Independent testing of the FLAME computer code, Version 1.0, was conducted to determine if the code is ready for use in hydrological and environmental studies at Department of Energy sites. This report describes the technical basis, approach, and results of this testing. Validation tests, (i.e., tests which compare field data to the computer generated solutions) were used to determine the operational status of the FLAME computer code and were done on a qualitative basis through graphical comparisons of the experimental and numerical data. These tests were specifically designed to check: (1) correctness of the FORTRAN coding, (2) computational accuracy, and (3) suitability to simulating actual hydrologic conditions. This testing was performed using a structured evaluation protocol which consisted of: (1) independent applications, and (2) graduated difficulty of test cases. Three tests ranging in complexity from simple one-dimensional steady-state flow field problems under near-saturated conditions to two-dimensional transient flow problems with very dry initial conditions
International Nuclear Information System (INIS)
Costa, L.; Fremont, R. de; Mougniot, J.C.; Msika, D.
1976-01-01
In this paper the proble ms of corrosion and activity in the primary coolant circuit of a sodium cooled fast reactor are treated. The main features of Corona, a computer code which has been used to perform the calculations, are outlined
SIVAR - Computer code for simulation of fuel rod behavior in PWR during fast transients
International Nuclear Information System (INIS)
Dias, A.F.V.
1980-10-01
Fuel rod behavior during a stationary and a transitory operation, is studied. A computer code aiming at simulating PWR type rods, was developed; however, it can be adapted for simulating other type of rods. A finite difference method was used. (E.G.) [pt
The computer code Eurdyn - 1 M. (Release 1) Part 2: User's Manual
International Nuclear Information System (INIS)
Donea, J.; Giuliani, S.
1979-01-01
This report is the user's manual for the computer code Eurdyn-1 M developed at the J.R.C. Ispra for use in containment and fuel subassembly analyses for fast reactor safety studies. The input data are defined and a test problem is presented to illustrate both the input and the output of results
International Nuclear Information System (INIS)
King, C.M.; Wilhite, E.L.; Root, R.W. Jr.
1985-01-01
The Savannah River Laboratory DOSTOMAN code has been used since 1978 for environmental pathway analysis of potential migration of radionuclides and hazardous chemicals. The DOSTOMAN work is reviewed including a summary of historical use of compartmental models, the mathematical basis for the DOSTOMAN code, examples of exact analytical solutions for simple matrices, methods for numerical solution of complex matrices, and mathematical validation/calibration of the SRL code. The review includes the methodology for application to nuclear and hazardous chemical waste disposal, examples of use of the model in contaminant transport and pathway analysis, a user's guide for computer implementation, peer review of the code, and use of DOSTOMAN at other Department of Energy sites. 22 refs., 3 figs
Simulation of small break loss of coolant accident using relap 5/ MOD 2 computer code
International Nuclear Information System (INIS)
Megahed, M.M.
1992-01-01
An assessment of relap 5 / MOD 2/Cycle 36.05 best estimate computer code capabilities in predicting the thermohydraulic response of a PWR following a small break loss of coolant accident is presented. The experimental data base for the evaluation is the results of Test S-N H-3 performed in the semi scale MOD-2 c Test facility which modeled a 0.5% small break loss of coolant accident with an accompanying failure of the high pressure injection emergency core cooling system. A conclusion was reached that the code is capable of making small break loss of coolant accident calculations efficiently. However, some of the small break loss of coolant accident related phenomena were not properly predicted by the code, suggesting a need for code improvement.9 fig., 3 tab
TEMP: a computer code to calculate fuel pin temperatures during a transient
International Nuclear Information System (INIS)
Bard, F.E.; Christensen, B.Y.; Gneiting, B.C.
1980-04-01
The computer code TEMP calculates fuel pin temperatures during a transient. It was developed to accommodate temperature calculations in any system of axi-symmetric concentric cylinders. When used to calculate fuel pin temperatures, the code will handle a fuel pin as simple as a solid cylinder or as complex as a central void surrounded by fuel that is broken into three regions by two circumferential cracks. Any fuel situation between these two extremes can be analyzed along with additional cladding, heat sink, coolant or capsule regions surrounding the fuel. The one-region version of the code accurately calculates the solution to two problems having closed-form solutions. The code uses an implicit method, an explicit method and a Crank-Nicolson (implicit-explicit) method
International Nuclear Information System (INIS)
Glazer, S.; Todreas, N.; Rohsenow, W.; Sonin, A.
1981-02-01
This document is intended as a user/programmer manual for the TRANSENERGY-S computer code. The code represents an extension of the steady state ENERGY model, originally developed by E. Khan, to predict coolant and fuel pin temperatures in a single LMFBR core assembly during transient events. Effects which may be modelled in the analysis include temporal variation in gamma heating in the coolant and duct wall, rod power production, coolant inlet temperature, coolant flow rate, and thermal boundary conditions around the single assembly. Numerical formulations of energy equations in the fuel and coolant are presented, and the solution schemes and stability criteria are discussed. A detailed description of the input deck preparation is presented, as well as code logic flowcharts, and a complete program listing. TRANSENERGY-S code predictions are compared with those of two different versions of COBRA, and partial results of a 61 pin bundle test case are presented
JPDYN-IV: a computer code for JPDR-II dynamics analysis
International Nuclear Information System (INIS)
Yokobayashi, Masao; Ishizuka, Makoto; Kishi, Akimasa; Wakabayashi, Yoshimune.
1978-12-01
The computer code JPDYN-IV developed is for analyzing dynamic characteristics of the JPDR-II BWR plant. It treats transient phenomena caused by various perturbations in normal operation of the reactor plant. The code features (1) a void map scheme to take into consideration the effect of a non-uniform void fraction distribution, (2) the treatment of a pressure change due to behavior of the stagnant water, and (3) a set of algebraic equations for analytical estimation of the flow rate in each region. Performance of the code was examined by way of the measured data in commissioning test of the JPDR-II up to 50% power. In the transient behavior caused by small perturbations, the calculated results agree well with the measured ones. For large perturbations, though the data are limitted, the agreement between calculation and measurement shows wide applicability of the code. (author)
Recommendations for computer modeling codes to support the UMTRA groundwater restoration project
International Nuclear Information System (INIS)
Tucker, M.D.; Khan, M.A.
1996-04-01
The Uranium Mill Tailings Remediation Action (UMTRA) Project is responsible for the assessment and remedial action at the 24 former uranium mill tailings sites located in the US. The surface restoration phase, which includes containment and stabilization of the abandoned uranium mill tailings piles, has a specific termination date and is nearing completion. Therefore, attention has now turned to the groundwater restoration phase, which began in 1991. Regulated constituents in groundwater whose concentrations or activities exceed maximum contaminant levels (MCLs) or background levels at one or more sites include, but are not limited to, uranium, selenium, arsenic, molybdenum, nitrate, gross alpha, radium-226 and radium-228. The purpose of this report is to recommend computer codes that can be used to assist the UMTRA groundwater restoration effort. The report includes a survey of applicable codes in each of the following areas: (1) groundwater flow and contaminant transport modeling codes, (2) hydrogeochemical modeling codes, (3) pump and treat optimization codes, and (4) decision support tools. Following the survey of the applicable codes, specific codes that can best meet the needs of the UMTRA groundwater restoration program in each of the four areas are recommended
Recommendations for computer modeling codes to support the UMTRA groundwater restoration project
Energy Technology Data Exchange (ETDEWEB)
Tucker, M.D. [Sandia National Labs., Albuquerque, NM (United States); Khan, M.A. [IT Corp., Albuquerque, NM (United States)
1996-04-01
The Uranium Mill Tailings Remediation Action (UMTRA) Project is responsible for the assessment and remedial action at the 24 former uranium mill tailings sites located in the US. The surface restoration phase, which includes containment and stabilization of the abandoned uranium mill tailings piles, has a specific termination date and is nearing completion. Therefore, attention has now turned to the groundwater restoration phase, which began in 1991. Regulated constituents in groundwater whose concentrations or activities exceed maximum contaminant levels (MCLs) or background levels at one or more sites include, but are not limited to, uranium, selenium, arsenic, molybdenum, nitrate, gross alpha, radium-226 and radium-228. The purpose of this report is to recommend computer codes that can be used to assist the UMTRA groundwater restoration effort. The report includes a survey of applicable codes in each of the following areas: (1) groundwater flow and contaminant transport modeling codes, (2) hydrogeochemical modeling codes, (3) pump and treat optimization codes, and (4) decision support tools. Following the survey of the applicable codes, specific codes that can best meet the needs of the UMTRA groundwater restoration program in each of the four areas are recommended.
Energy Technology Data Exchange (ETDEWEB)
Maker, B.N.
1995-04-14
This report provides a user`s manual for NIKE3D, a fully implicit three-dimensional finite element code for analyzing the finite strain static and dynamic response of inelastic solids, shells, and beams. Spatial discretization is achieved by the use of 8-node solid elements, 2-node truss and beam elements, and 4-node membrane and shell elements. Over twenty constitutive models are available for representing a wide range of elastic, plastic, viscous, and thermally dependent material behavior. Contact-impact algorithms permit gaps, frictional sliding, and mesh discontinuities along material interfaces. Several nonlinear solution strategies are available, including Full-, Modified-, and Quasi-Newton methods. The resulting system of simultaneous linear equations is either solved iteratively by an element-by-element method, or directly by a factorization method, for which case bandwidth minimization is optional. Data may be stored either in or out of core memory to allow for large analyses.
International Nuclear Information System (INIS)
Nanty, Simon
2015-01-01
This work relates to the framework of uncertainty quantification for numerical simulators, and more precisely studies two industrial applications linked to the safety studies of nuclear plants. These two applications have several common features. The first one is that the computer code inputs are functional and scalar variables, functional ones being dependent. The second feature is that the probability distribution of functional variables is known only through a sample of their realizations. The third feature, relative to only one of the two applications, is the high computational cost of the code, which limits the number of possible simulations. The main objective of this work was to propose a complete methodology for the uncertainty analysis of numerical simulators for the two considered cases. First, we have proposed a methodology to quantify the uncertainties of dependent functional random variables from a sample of their realizations. This methodology enables to both model the dependency between variables and their link to another variable, called co-variate, which could be, for instance, the output of the considered code. Then, we have developed an adaptation of a visualization tool for functional data, which enables to simultaneously visualize the uncertainties and features of dependent functional variables. Second, a method to perform the global sensitivity analysis of the codes used in the two studied cases has been proposed. In the case of a computationally demanding code, the direct use of quantitative global sensitivity analysis methods is intractable. To overcome this issue, the retained solution consists in building a surrogate model or meta model, a fast-running model approximating the computationally expensive code. An optimized uniform sampling strategy for scalar and functional variables has been developed to build a learning basis for the meta model. Finally, a new approximation approach for expensive codes with functional outputs has been
Ioannides, A A
1987-01-01
The nature of the available computing tools strongly influences modern scientific investigations. The sources of well known problems associated with the use of procedural computer languages are traced and their consequences investigated. The likely impact of recent quantitative and qualitative advances in software and hardware is examined with emphasis on its relevance to the biomagnetic inverse problem. Gradual changes in the use of computers, some already employed in a recent study of a specific biomagnetic inverse problem, are outlined which take into account the large investment in conventional codes.
International Nuclear Information System (INIS)
Strenge, D.L.; Watson, E.C.; Droppo, J.G.
1976-06-01
The development of technological bases for siting nuclear fuel cycle facilities requires calculational models and computer codes for the evaluation of risks and the assessment of environmental impact of radioactive effluents. A literature search and review of available computer programs revealed that no one program was capable of performing all of the great variety of calculations (i.e., external dose, internal dose, population dose, chronic release, accidental release, etc.). Available literature on existing computer programs has been reviewed and a description of each program reviewed is given
Energy Technology Data Exchange (ETDEWEB)
Strenge, D.L.; Watson, E.C.; Droppo, J.G.
1976-06-01
The development of technological bases for siting nuclear fuel cycle facilities requires calculational models and computer codes for the evaluation of risks and the assessment of environmental impact of radioactive effluents. A literature search and review of available computer programs revealed that no one program was capable of performing all of the great variety of calculations (i.e., external dose, internal dose, population dose, chronic release, accidental release, etc.). Available literature on existing computer programs has been reviewed and a description of each program reviewed is given.
Finite elements numerical codes as primary tool to improve beam optics in NIO1
Baltador, C.; Cavenago, M.; Veltri, P.; Serianni, G.
2017-08-01
The RF negative ion source NIO1, built at Consorzio RFX in Padua (Italy), is aimed to investigate general issues on ion source physics in view of the full-size ITER injector MITICA as well as DEMO relevant solutions, like energy recovery and alternative neutralization systems, crucial for neutral beam injectors in future fusion experiments. NIO1 has been designed to produce 9 H-beamlets (in a 3x3 pattern) of 15mA each and 60keV, using a three electrodes system downstream the plasma source. At the moment the source is at its early operational stage and only operation at low power and low beam energy is possible. In particular, NIO1 presents a too strong set of SmCo co-extraction electron suppression magnets (CESM) in the extraction grid (EG) that will be replaced by a weaker set of Ferrite magnets. A completely new set of magnets will be also designed and mounted on the new EG that will be installed next year, replacing the present one. In this paper, the finite element code OPERA 3D is used to investigate the effects of the three sets of magnets on beamlet optics. A comparison of numerical results with measurements will be provided where possible.
Directory of Open Access Journals (Sweden)
Min Kyung Sung
2014-12-01
Full Text Available Genome-wide association studies have proven the highly polygenic architecture of complex diseases or traits; therefore, single-locus-based methods are usually unable to detect all involved loci, especially when individual loci exert small effects. Moreover, the majority of associated single-nucleotide polymorphisms resides in non-coding regions, making it difficult to understand their phenotypic contribution. In this work, we studied epistatic interactions associated with three common diseases using Korea Association Resource (KARE data: type 2 diabetes mellitus (DM, hypertension (HT, and coronary artery disease (CAD. We showed that epistatic single-nucleotide polymorphisms (SNPs were enriched in enhancers, as well as in DNase I footprints (the Encyclopedia of DNA Elements [ENCODE] Project Consortium 2012, which suggested that the disruption of the regulatory regions where transcription factors bind may be involved in the disease mechanism. Accordingly, to identify the genes affected by the SNPs, we employed whole-genome multiple-cell-type enhancer data which discovered using DNase I profiles and Cap Analysis Gene Expression (CAGE. Assigned genes were significantly enriched in known disease associated gene sets, which were explored based on the literature, suggesting that this approach is useful for detecting relevant affected genes. In our knowledge-based epistatic network, the three diseases share many associated genes and are also closely related with each other through many epistatic interactions. These findings elucidate the genetic basis of the close relationship between DM, HT, and CAD.
High-Fidelity Buckling Analysis of Composite Cylinders Using the STAGS Finite Element Code
Hilburger, Mark W.
2014-01-01
Results from previous shell buckling studies are presented that illustrate some of the unique and powerful capabilities in the STAGS finite element analysis code that have made it an indispensable tool in structures research at NASA over the past few decades. In particular, prototypical results from the development and validation of high-fidelity buckling simulations are presented for several unstiffened thin-walled compression-loaded graphite-epoxy cylindrical shells along with a discussion on the specific methods and user-defined subroutines in STAGS that are used to carry out the high-fidelity simulations. These simulations accurately account for the effects of geometric shell-wall imperfections, shell-wall thickness variations, local shell-wall ply-gaps associated with the fabrication process, shell-end geometric imperfections, nonuniform applied end loads, and elastic boundary conditions. The analysis procedure uses a combination of nonlinear quasi-static and transient dynamic solution algorithms to predict the prebuckling and unstable collapse response characteristics of the cylinders. Finally, the use of high-fidelity models in the development of analysis-based shell-buckling knockdown (design) factors is demonstrated.
Choudhary, Aniruddha
Code verifition is the process of ensuring, to the degree possible, that there are no algorithm deficiencies and coding mistakes (bugs) in a scientific computing simulation. In this work, techniques are presented for performing code verifition of boundary conditions commonly used in compressible and incompressible Computational Fluid Dynamics (CFD) codes. Using a compressible CFD code, this study assesses the subsonic in flow (isentropic and fixed-mass), subsonic out ow, supersonic out ow, no-slip wall (adiabatic and isothermal), and inviscid slip-wall. The use of simplified curved surfaces is proposed for easier generation of manufactured solutions during the verifition of certain boundary conditions involving many constraints. To perform rigorous code verifition, general grids with mixed cell types at the verified boundary are used. A novel approach is introduced to determine manufactured solutions for boundary condition verifition when the velocity-field is constrained to be divergence-free during the simulation in an incompressible CFD code. Order of accuracy testing using the Method of Manufactured Solutions (MMS) is employed here for code verifition of the major components of an open-source, multiphase ow code - MFIX. The presence of two-phase governing equations and a modified SIMPLE-based algorithm requiring divergence-free flows makes the selection of manufactured solutions more involved than for single-phase, compressible flows. Code verifition is performed here on 2D and 3D, uniform and stretched meshes for incompressible, steady and unsteady, single-phase and two-phase flows using the two-fluid model of MFIX. In a CFD simulation, truncation error (TE) is the difference between the continuous governing equation and its discrete approximation. Since TE can be shown to be the local source term for the discretization error, TE is proposed as the criterion for determining which regions of the computational mesh should be refined/coarsened. For mesh
TBCI and URMEL - New computer codes for wake field and cavity mode calculations
International Nuclear Information System (INIS)
Weiland, T.
1983-01-01
Wake force computation is important for any study of instabilities in high current accelerators and storage rings. These forces are generated by intense bunches of charged particles passing cylindrically symmetric structures on or off axis. The adequate method for computing such forces is the time domain approach. The computer Code TBCI computes for relativistic as well as for nonrelativistic bunches of arbitrary shape longitudinal and transverse wake forces up to the octupole component. TBCI is not limited to cavity-like objects and thus applicable to bellows, beam pipes with varying cross sections and any other nonresonant structures. For the accelerating cavities one also needs to know the resonant modes and frequencies for the study of instabilities and mode couplers. The complementary code named URMEL computes these fields for any azimuthal dependence of the fields in ascending order. The mathematical procedure being used is very safe and does not miss modes. Both codes together represent a unique tool for accelerator design and are easy to use
Directory of Open Access Journals (Sweden)
Chia-Chang Hu
2005-04-01
Full Text Available A novel space-time adaptive near-far robust code-synchronization array detector for asynchronous DS-CDMA systems is developed in this paper. There are the same basic requirements that are needed by the conventional matched filter of an asynchronous DS-CDMA system. For the real-time applicability, a computationally efficient architecture of the proposed detector is developed that is based on the concept of the multistage Wiener filter (MWF of Goldstein and Reed. This multistage technique results in a self-synchronizing detection criterion that requires no inversion or eigendecomposition of a covariance matrix. As a consequence, this detector achieves a complexity that is only a linear function of the size of antenna array (J, the rank of the MWF (M, the system processing gain (N, and the number of samples in a chip interval (S, that is, Ã°ÂÂ’Âª(JMNS. The complexity of the equivalent detector based on the minimum mean-squared error (MMSE or the subspace-based eigenstructure analysis is a function of Ã°ÂÂ’Âª((JNS3. Moreover, this multistage scheme provides a rapid adaptive convergence under limited observation-data support. Simulations are conducted to evaluate the performance and convergence behavior of the proposed detector with the size of the J-element antenna array, the amount of the L-sample support, and the rank of the M-stage MWF. The performance advantage of the proposed detector over other DS-CDMA detectors is investigated as well.
Hardman, R. R.; Mahan, J. R.; Smith, M. H.; Gelhausen, P. A.; Van Dalsem, W. R.
1991-01-01
The need for a validation technique for computational fluid dynamics (CFD) codes in STOVL applications has led to research efforts to apply infrared thermal imaging techniques to visualize gaseous flow fields. Specifically, a heated, free-jet test facility was constructed. The gaseous flow field of the jet exhaust was characterized using an infrared imaging technique in the 2 to 5.6 micron wavelength band as well as conventional pitot tube and thermocouple methods. These infrared images are compared to computer-generated images using the equations of radiative exchange based on the temperature distribution in the jet exhaust measured with the thermocouple traverses. Temperature and velocity measurement techniques, infrared imaging, and the computer model of the infrared imaging technique are presented and discussed. From the study, it is concluded that infrared imaging techniques coupled with the radiative exchange equations applied to CFD models are a valid method to qualitatively verify CFD codes used in STOVL applications.
Multiple frequencies sequential coding for SSVEP-based brain-computer interface.
Directory of Open Access Journals (Sweden)
Yangsong Zhang
Full Text Available BACKGROUND: Steady-state visual evoked potential (SSVEP-based brain-computer interface (BCI has become one of the most promising modalities for a practical noninvasive BCI system. Owing to both the limitation of refresh rate of liquid crystal display (LCD or cathode ray tube (CRT monitor, and the specific physiological response property that only a very small number of stimuli at certain frequencies could evoke strong SSVEPs, the available frequencies for SSVEP stimuli are limited. Therefore, it may not be enough to code multiple targets with the traditional frequencies coding protocols, which poses a big challenge for the design of a practical SSVEP-based BCI. This study aimed to provide an innovative coding method to tackle this problem. METHODOLOGY/PRINCIPAL FINDINGS: In this study, we present a novel protocol termed multiple frequencies sequential coding (MFSC for SSVEP-based BCI. In MFSC, multiple frequencies are sequentially used in each cycle to code the targets. To fulfill the sequential coding, each cycle is divided into several coding epochs, and during each epoch, certain frequency is used. Obviously, different frequencies or the same frequency can be presented in the coding epochs, and the different epoch sequence corresponds to the different targets. To show the feasibility of MFSC, we used two frequencies to realize four targets and carried on an offline experiment. The current study shows that: 1 MFSC is feasible and efficient; 2 the performance of SSVEP-based BCI based on MFSC can be comparable to some existed systems. CONCLUSIONS/SIGNIFICANCE: The proposed protocol could potentially implement much more targets with the limited available frequencies compared with the traditional frequencies coding protocol. The efficiency of the new protocol was confirmed by real data experiment. We propose that the SSVEP-based BCI under MFSC might be a promising choice in the future.
Zhao, Shengmei; Wang, Le; Liang, Wenqiang; Cheng, Weiwen; Gong, Longyan
2015-10-01
In this paper, we propose a high performance optical encryption (OE) scheme based on computational ghost imaging (GI) with QR code and compressive sensing (CS) technique, named QR-CGI-OE scheme. N random phase screens, generated by Alice, is a secret key and be shared with its authorized user, Bob. The information is first encoded by Alice with QR code, and the QR-coded image is then encrypted with the aid of computational ghost imaging optical system. Here, measurement results from the GI optical system's bucket detector are the encrypted information and be transmitted to Bob. With the key, Bob decrypts the encrypted information to obtain the QR-coded image with GI and CS techniques, and further recovers the information by QR decoding. The experimental and numerical simulated results show that the authorized users can recover completely the original image, whereas the eavesdroppers can not acquire any information about the image even the eavesdropping ratio (ER) is up to 60% at the given measurement times. For the proposed scheme, the number of bits sent from Alice to Bob are reduced considerably and the robustness is enhanced significantly. Meantime, the measurement times in GI system is reduced and the quality of the reconstructed QR-coded image is improved.
MOLOCH computer code for molecular-dynamics simulation of processes in condensed matter
Directory of Open Access Journals (Sweden)
Derbenev I.V.
2011-01-01
Full Text Available Theoretical and experimental investigation into properties of condensed matter is one of the mainstreams in RFNC-VNIITF scientific activity. The method of molecular dynamics (MD is an innovative method of theoretical materials science. Modern supercomputers allow the direct simulation of collective effects in multibillion atom sample, making it possible to model physical processes on the atomistic level, including material response to dynamic load, radiation damage, influence of defects and alloying additions upon material mechanical properties, or aging of actinides. During past ten years, the computer code MOLOCH has been developed at RFNC-VNIITF. It is a parallel code suitable for massive parallel computing. Modern programming techniques were used to make the code almost 100% efficient. Practically all instruments required for modelling were implemented in the code: a potential builder for different materials, simulation of physical processes in arbitrary 3D geometry, and calculated data processing. A set of tests was developed to analyse algorithms efficiency. It can be used to compare codes with different MD implementation between each other.
Ethical codes. Fig leaf argument, ballast or cultural element for radiation protection?
International Nuclear Information System (INIS)
Gellermann, Rainer
2014-01-01
The international association for radiation protection (IRPA) adopted in May 2004 a Code of Ethics in order to allow their members to hold an adequate professional level of ethical line of action. Based on this code of ethics the professional body of radiation protection (Fachverband fuer Strahlenschutz) has developed its own ethical code and adopted in 2005.
A dual origin of the Xist gene from a protein-coding gene and a set of transposable elements.
Directory of Open Access Journals (Sweden)
Eugeny A Elisaphenko
2008-06-01
Full Text Available X-chromosome inactivation, which occurs in female eutherian mammals is controlled by a complex X-linked locus termed the X-inactivation center (XIC. Previously it was proposed that genes of the XIC evolved, at least in part, as a result of pseudogenization of protein-coding genes. In this study we show that the key XIC gene Xist, which displays fragmentary homology to a protein-coding gene Lnx3, emerged de novo in early eutherians by integration of mobile elements which gave rise to simple tandem repeats. The Xist gene promoter region and four out of ten exons found in eutherians retain homology to exons of the Lnx3 gene. The remaining six Xist exons including those with simple tandem repeats detectable in their structure have similarity to different transposable elements. Integration of mobile elements into Xist accompanies the overall evolution of the gene and presumably continues in contemporary eutherian species. Additionally we showed that the combination of remnants of protein-coding sequences and mobile elements is not unique to the Xist gene and is found in other XIC genes producing non-coding nuclear RNA.
On Finite Element Computations of Contact Problems in Micropolar Elasticity
Eremeyev, Victor A.; Skrzat, Andrzej; Stachowicz, Feliks
2016-01-01
Within the linear micropolar elasticity we discuss the development of new finite element and its implementation in commercial software. Here we implement the developed 8-node hybrid isoparametric element into ABAQUS and perform solutions of contact problems. We consider the contact of polymeric stamp modelled within the micropolar elasticity with an elastic substrate. The peculiarities of modelling of contact problems with a user defined finite element in ABAQUS are discussed. The provided co...
International Nuclear Information System (INIS)
Kress, T.S.
1985-04-01
The determination of severe accident source terms must, by necessity it seems, rely heavily on the use of complex computer codes. Source term acceptability, therefore, rests on the assessed validity of such codes. Consequently, one element of NRC's recent efforts to reassess LWR severe accident source terms is to provide a review of the status of validation of the computer codes used in the reassessment. The results of this review is the subject of this document. The separate review documents compiled in this report were used as a resource along with the results of the BMI-2104 study by BCL and the QUEST study by SNL to arrive at a more-or-less independent appraisal of the status of source term modeling at this time
Energy Technology Data Exchange (ETDEWEB)
Kress, T. S. [comp.
1985-04-01
The determination of severe accident source terms must, by necessity it seems, rely heavily on the use of complex computer codes. Source term acceptability, therefore, rests on the assessed validity of such codes. Consequently, one element of NRC's recent efforts to reassess LWR severe accident source terms is to provide a review of the status of validation of the computer codes used in the reassessment. The results of this review is the subject of this document. The separate review documents compiled in this report were used as a resource along with the results of the BMI-2104 study by BCL and the QUEST study by SNL to arrive at a more-or-less independent appraisal of the status of source term modeling at this time.