International Nuclear Information System (INIS)
The 2-MW TRIGA MARK II research reactor at Centre National de l'Energie, des Sciences et des Techniques Nucleaires (CNESTEN) achieved initial criticality on May 2, 2007 with 71 fuel elements. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower and training and production of radioisotopes for their use in agriculture, industry and medicine. This work aims to study the time-dependent neutronics parameters of the TRIGA reactor for elaborating and planning of an in-core fuel management strategy to maximize the utilization of the TRIGA fluxes, using a new elaborated burnup computer code called 'BUCAL1'. The code can be used to aid in analysis, prediction, and optimization of fuel burnup performance in a nuclear reactor. It was developed to incorporate the neutron absorption tally/reaction information generated directly by MCNP5 code in the calculation of fissioned or neutron-transmuted isotopes for multi-fueled regions. The use of Monte Carlo method and punctual cross section data characterizing the MCNP code allows an accurate simulation of neutron life cycle in the reactor, and the integration of data on the entire energy spectrum, thus a more accurate estimation of results than deterministic code can do. Also, for the purpose of this study, a full-model of the TRIGA reactor was developed using the MCNP5 code. The validation of the MCNP model of the TRIGA reactor was made by benchmarking the reactivity experiments. (author)
Accuracy assessment of a new Monte Carlo based burnup computer code
International Nuclear Information System (INIS)
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.
International Nuclear Information System (INIS)
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)
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.)
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)
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
Validation of a new continuous Monte Carlo burnup code using a Mox fuel assembly
International Nuclear Information System (INIS)
The reactivity of nuclear fuel decreases with irradiation (or burnup) due to the transformation of heavy nuclides and the formation of fission products. Burnup credit studies aim at accounting for fuel irradiation in criticality studies of the nuclear fuel cycle (transport, storage, etc...). The principal objective of this study is to evaluate the potential capabilities of a newly developed burnup code called 'BUCAL1'. BUCAL1 differs in comparison with other burnup codes as it does not use the calculated neutron flux as input to other computer codes to generate the nuclide inventory for the next time step. Instead, BUCAL1 directly uses the neutron reaction tally information generated by MCNP for each nuclide of interest to determine the new nuclides inventory. This allows the full capabilities of MCNP to be incorporated into the calculation and a more accurate and robust analysis to be performed. Validation of BUCAL1 was processed by code-to-code comparisons using predictions of several codes from the NEA/OCED. Infinite multiplication factors (k∞) and important fission product and actinide concentrations were compared for a MOX core benchmark exercise. Results of calculations are analysed and discussed.
Geochemical computer codes. A review
International Nuclear Information System (INIS)
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)
Computer code abstract: NESTLE
International Nuclear Information System (INIS)
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
Computer codes in accelerator domain
International Nuclear Information System (INIS)
In this report a list of computer codes for calculations in accelerator physics is presented. The codes concern the design of accelerator shieldings, beam dynamics of synchrotrons and storage rings, the simulation of radiation fields in accelerators, the design of RF cavities, beam dynamics of microtrons, the optics of charged-particle beams, the design of accelerator components, the calculation of magnetic fields, the computation of thermal and mechanical processes in accelerator structures, the design of magnets, and the optimization of beam lines. Most of the codes are written in FORTRAN. (HSI) nge of computational results and pieces of software via E-mail. Also outstanding is the problem of a more efficient application of the known and tested forms of communication, e.g. selection and systematization of the data on the available program packages, Workshops of the interested users and unification of experts into working groups. (orig.)
Computer access security code system
Collins, Earl R., Jr. (Inventor)
1990-01-01
A security code system for controlling access to computer and computer-controlled entry situations comprises a plurality of subsets of alpha-numeric characters disposed in random order in matrices of at least two dimensions forming theoretical rectangles, cubes, etc., such that when access is desired, at least one pair of previously unused character subsets not found in the same row or column of the matrix is chosen at random and transmitted by the computer. The proper response to gain access is transmittal of subsets which complete the rectangle, and/or a parallelepiped whose opposite corners were defined by first groups of code. Once used, subsets are not used again to absolutely defeat unauthorized access by eavesdropping, and the like.
Computer code ONEBFP. Final report
International Nuclear Information System (INIS)
The final product of Task 10, the computer code ONEBFP, was delivered to the customer on 28 July 1995. ONEBFP is a newly written transport module specifically designed for use in the BOXIEMP code system. ONEBFP uses advanced transport methods and should provide better accuracy than the existing transport module in BOXIEMP. ONEBFP was designed for ease of maintenance and replaces an existing transport module consisting of over 100,000 lines of code. ONEBFP consists of 6632 lines of Fortran code, of which 3412 are comment lines used for documentation. Dynamic memory storage is used for all data arrays. Three methods are available; one is chosen at installation time. They are Fortran 90 automatic arrays, Fortran 90 Allocate/Deallocate, or the Allocate/Deallocate extension to Fortran 77 allowed in the Microsoft Fortran Powerstation product for IBM PC compatibles. ONEBFP requires two input files, a binary cross section input file and a text file of controls and other specifications. Two output files are produced, the first is a report file that regurgitates the input from the text input and a summary of how the calculation progressed and the second contains detailed pointwise and angular quantifies such as the flux that are the results of the calculation. This second file serves as the interface to the postprocessing modules of the BOXIEMP code
Computer code conversion using HISTORIAN
International Nuclear Information System (INIS)
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)
CHAINT computer code. Users guide
International Nuclear Information System (INIS)
CHAINT is a two-dimensional numerical model for the analysis of contaminant transport in a fractured porous medium. The physical processes accounted for include advection, dispersion, diffusion, retardation, radionuclide chain decay coupling, and mass injection. The computational scheme employed by CHAINT is based on a Galerkin finite-element method and block-diagonal frontal solution technique. Continuum portions of the medium may be modeled with two-dimensional isoparametric elements. Discrete features are modeled with one-dimensional elements that are embedded along the sides of the continuum elements. Principal input to this model consists of files from a predecessor MAGNUM-2D simulation of buoyancy driven fluid flow. Output from CHAINT includes a printed report of contaminant concentrations along with postprocessor graphics files. This report contains information relevant to general usage of the CHAINT code
Simplified computer codes for cask impact analysis
International Nuclear Information System (INIS)
In regard to the evaluation of the acceleration and deformation of casks, the simplified computer codes make analyses economical and decrease input and calculation time. The results obtained by the simplified computer codes have enough adequacy for their practical use. (J.P.N.)
Computer codes for RF cavity design
International Nuclear Information System (INIS)
In RF cavity design, numerical modeling is assuming an increasingly important role with the help of sophisticated computer codes and powerful yet affordable computers. A description of the cavity codes in use in the accelerator community has been given previously. The present paper will address the latest developments and discuss their applications to cavity tuning and matching problems. (Author) 8 refs., 10 figs
Computer codes for RF cavity design
International Nuclear Information System (INIS)
In RF cavity design, numerical modeling is assuming an increasingly important role with the help of sophisticated computer codes and powerful yet affordable computers. A description of the cavity codes in use in the accelerator community has been given previously. The present paper will address the latest developments and discuss their applications to cavity toning and matching problems
Implatation of MC2 computer code
International Nuclear Information System (INIS)
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.)
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)
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)
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)
Topological Code Architectures for Quantum Computation
Cesare, Christopher Anthony
This dissertation is concerned with quantum computation using many-body quantum systems encoded in topological codes. The interest in these topological systems has increased in recent years as devices in the lab begin to reach the fidelities required for performing arbitrarily long quantum algorithms. The most well-studied system, Kitaev's toric code, provides both a physical substrate for performing universal fault-tolerant quantum computations and a useful pedagogical tool for explaining the way other topological codes work. In this dissertation, I first review the necessary formalism for quantum information and quantum stabilizer codes, and then I introduce two families of topological codes: Kitaev's toric code and Bombin's color codes. I then present three chapters of original work. First, I explore the distinctness of encoding schemes in the color codes. Second, I introduce a model of quantum computation based on the toric code that uses adiabatic interpolations between static Hamiltonians with gaps constant in the system size. Lastly, I describe novel state distillation protocols that are naturally suited for topological architectures and show that they provide resource savings in terms of the number of required ancilla states when compared to more traditional approaches to quantum gate approximation.
Computer codes in particle transport physics
International Nuclear Information System (INIS)
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
Cluster Computing: A Mobile Code Approach
Patel, R. B.; Manpreet Singh
2006-01-01
Cluster computing harnesses the combined computing power of multiple processors in a parallel configuration. Cluster Computing environments built from commodity hardware have provided a cost-effective solution for many scientific and high-performance applications. In this paper we have presented design and implementation of a cluster based framework using mobile code. The cluster implementation involves the designing of a server named MCLUSTER which manages the configuring, resetting of clust...
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)
The Institute of Protection and Nuclear Safety (I.P.S.N.) in the frame of dosimetric impact estimation of releases from nuclear facilities in usual operation, evaluates the contamination transfer by the continental aquatic medium, then by the food chain up to man. To do it, I.P.S.N. has the AQUAREJ 2.3 computer code for the evaluation of dosimetric consequences of liquid radioactive effluents releases in the rivers. A comparison is made with the PC-CREAM computer code. (N.C.)
Development of probabilistic multimedia multipathway computer codes.
Energy Technology Data Exchange (ETDEWEB)
Yu, C.; LePoire, D.; Gnanapragasam, E.; Arnish, J.; Kamboj, S.; Biwer, B. M.; Cheng, J.-J.; Zielen, A. J.; Chen, S. Y.; Mo, T.; Abu-Eid, R.; Thaggard, M.; Sallo, A., III.; Peterson, H., Jr.; Williams, W. A.; Environmental Assessment; NRC; EM
2002-01-01
The deterministic multimedia dose/risk assessment codes RESRAD and RESRAD-BUILD have been widely used for many years for evaluation of sites contaminated with residual radioactive materials. The RESRAD code applies to the cleanup of sites (soils) and the RESRAD-BUILD code applies to the cleanup of buildings and structures. This work describes the procedure used to enhance the deterministic RESRAD and RESRAD-BUILD codes for probabilistic dose analysis. A six-step procedure was used in developing default parameter distributions and the probabilistic analysis modules. These six steps include (1) listing and categorizing parameters; (2) ranking parameters; (3) developing parameter distributions; (4) testing parameter distributions for probabilistic analysis; (5) developing probabilistic software modules; and (6) testing probabilistic modules and integrated codes. The procedures used can be applied to the development of other multimedia probabilistic codes. The probabilistic versions of RESRAD and RESRAD-BUILD codes provide tools for studying the uncertainty in dose assessment caused by uncertain input parameters. The parameter distribution data collected in this work can also be applied to other multimedia assessment tasks and multimedia computer codes.
COLD-SAT Dynamic Model Computer Code
Bollenbacher, G.; Adams, N. S.
1995-01-01
COLD-SAT Dynamic Model (CSDM) computer code implements six-degree-of-freedom, rigid-body mathematical model for simulation of spacecraft in orbit around Earth. Investigates flow dynamics and thermodynamics of subcritical cryogenic fluids in microgravity. Consists of three parts: translation model, rotation model, and slosh model. Written in FORTRAN 77.
Error-correcting codes in computer arithmetic.
Massey, J. L.; Garcia, O. N.
1972-01-01
Summary of the most important results so far obtained in the theory of coding for the correction and detection of errors in computer arithmetic. Attempts to satisfy the stringent reliability demands upon the arithmetic unit are considered, and special attention is given to attempts to incorporate redundancy into the numbers themselves which are being processed so that erroneous results can be detected and corrected.
LATTICE: an interactive lattice computer code
International Nuclear Information System (INIS)
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...
Cluster Computing: A Mobile Code Approach
Directory of Open Access Journals (Sweden)
R. B. Patel
2006-01-01
Full Text Available Cluster computing harnesses the combined computing power of multiple processors in a parallel configuration. Cluster Computing environments built from commodity hardware have provided a cost-effective solution for many scientific and high-performance applications. In this paper we have presented design and implementation of a cluster based framework using mobile code. The cluster implementation involves the designing of a server named MCLUSTER which manages the configuring, resetting of cluster. It allows a user to provide necessary information regarding the application to be executed via a graphical user interface (GUI. Framework handles- the generation of application mobile code and its distribution to appropriate client nodes, efficient handling of results so generated and communicated by a number of client nodes and recording of execution time of application. The client node receives and executes the mobile code that defines the distributed job submitted by MCLUSTER server and replies the results back. We have also the analyzed the performance of the developed system emphasizing the tradeoff between communication and computation overhead.
Computer codes used in particle accelerator design: First edition
International Nuclear Information System (INIS)
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
Present state of the SOURCES computer code
Energy Technology Data Exchange (ETDEWEB)
Shores, E. F. (Erik F.)
2002-01-01
In various stages of development for over two decades, the SOURCES computer code continues to calculate neutron production rates and spectra from four types of problems: homogeneous media, two-region interfaces, three-region interfaces and that of a monoenergetic alpha particle beam incident on a slab of target material. Graduate work at the University of Missouri - Rolla, in addition to user feedback from a tutorial course, provided the impetus for a variety of code improvements. Recently upgraded to version 4B, initial modifications to SOURCES focused on updates to the 'tape5' decay data library. Shortly thereafter, efforts focused on development of a graphical user interface for the code. This paper documents the Los Alamos SOURCES Tape1 Creator and Library Link (LASTCALL) and describes additional library modifications in more detail. Minor improvements and planned enhancements are discussed.
ABINIT: a computer code for matter
International Nuclear Information System (INIS)
The PAW (Projector Augmented Wave) method has been implemented in the ABINIT Code that computes electronic structures in atoms. This method relies on the simultaneous use of a set of auxiliary functions (in plane waves) and a sphere around each atom. This method allows the computation of systems including many atoms and gives the expression of energy, forces, stress... in terms of the auxiliary function only. We have generated atomic data for iron at very high pressure (over 200 GPa). We get a bcc-hcp transition around 10 GPa and the magnetic order disappears around 50 GPa. This method has been validated on a series of metals. The development of the PAW method has required a great effort for the massive parallelization of the ABINIT code. (A.C.)
Probabilistic structural analysis computer code (NESSUS)
Shiao, Michael C.
1988-01-01
Probabilistic structural analysis has been developed to analyze the effects of fluctuating loads, variable material properties, and uncertain analytical models especially for high performance structures such as SSME turbopump blades. The computer code NESSUS (Numerical Evaluation of Stochastic Structure Under Stress) was developed to serve as a primary computation tool for the characterization of the probabilistic structural response due to the stochastic environments by statistical description. The code consists of three major modules NESSUS/PRE, NESSUS/FEM, and NESSUS/FPI. NESSUS/PRE is a preprocessor which decomposes the spatially correlated random variables into a set of uncorrelated random variables using a modal analysis method. NESSUS/FEM is a finite element module which provides structural sensitivities to all the random variables considered. NESSUS/FPI is Fast Probability Integration method by which a cumulative distribution function or a probability density function is calculated.
Poisson/Superfish codes for personal computers
International Nuclear Information System (INIS)
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
Quality assurance of the computer code INDAR
International Nuclear Information System (INIS)
Detailed aquatic dispersion and radiation exposure models are required in order to assess the radiological impact of routine aquatic discharges from nuclear power stations in the United Kingdom. Such models have been developed and incorporated in the computer program INDAR. This report describes the quality assurance procedures adopted in producing and testing the first release of the code, which was complied in November 1988 and is currently stored in the production load module PROD.INDAR.V10. (author)
Validation Report for ISAAC Computer Code
International Nuclear Information System (INIS)
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.
New developments in the Saphire computer codes
International Nuclear Information System (INIS)
The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) refers to a suite of computer programs that were developed to create and analyze a probabilistic risk assessment (PRA) of a nuclear power plant. Many recent enhancements to this suite of codes have been made. This presentation will provide an overview of these features and capabilities. The presentation will include a discussion of the new GEM module. This module greatly reduces and simplifies the work necessary to use the SAPHIRE code in event assessment applications. An overview of the features provided in the new Windows version will also be provided. This version is a full Windows 32-bit implementation and offers many new and exciting features. [A separate computer demonstration was held to allow interested participants to get a preview of these features.] The new capabilities that have been added since version 5.0 will be covered. Some of these major new features include the ability to store an unlimited number of basic events, gates, systems, sequences, etc.; the addition of improved reporting capabilities to allow the user to generate and open-quotes scrollclose quotes through custom reports; the addition of multi-variable importance measures; and the simplification of the user interface. Although originally designed as a PRA Level 1 suite of codes, capabilities have recently been added to SAPHIRE to allow the user to apply the code in Level 2 analyses. These features will be discussed in detail during the presentation. The modifications and capabilities added to this version of SAPHIRE significantly extend the code in many important areas. Together, these extensions represent a major step forward in PC-based risk analysis tools. This presentation provides a current up-to-date status of these important PRA analysis tools
New developments in the Saphire computer codes
Energy Technology Data Exchange (ETDEWEB)
Russell, K.D.; Wood, S.T.; Kvarfordt, K.J. [Idaho Engineering Lab., Idaho Falls, ID (United States)] [and others
1996-03-01
The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) refers to a suite of computer programs that were developed to create and analyze a probabilistic risk assessment (PRA) of a nuclear power plant. Many recent enhancements to this suite of codes have been made. This presentation will provide an overview of these features and capabilities. The presentation will include a discussion of the new GEM module. This module greatly reduces and simplifies the work necessary to use the SAPHIRE code in event assessment applications. An overview of the features provided in the new Windows version will also be provided. This version is a full Windows 32-bit implementation and offers many new and exciting features. [A separate computer demonstration was held to allow interested participants to get a preview of these features.] The new capabilities that have been added since version 5.0 will be covered. Some of these major new features include the ability to store an unlimited number of basic events, gates, systems, sequences, etc.; the addition of improved reporting capabilities to allow the user to generate and {open_quotes}scroll{close_quotes} through custom reports; the addition of multi-variable importance measures; and the simplification of the user interface. Although originally designed as a PRA Level 1 suite of codes, capabilities have recently been added to SAPHIRE to allow the user to apply the code in Level 2 analyses. These features will be discussed in detail during the presentation. The modifications and capabilities added to this version of SAPHIRE significantly extend the code in many important areas. Together, these extensions represent a major step forward in PC-based risk analysis tools. This presentation provides a current up-to-date status of these important PRA analysis tools.
Parallel computing in the PRIZMA code
International Nuclear Information System (INIS)
PRIZMA code is a Monte-Carlo code dedicated to radiation transport. The paper describes parallelization and load balance algorithms implemented in the PRIZMA code. The parallelization algorithm described aims to maximally separate batch calculation and calculated results collection at minimal wastage. The algorithm implements a two-level procedure to collect calculated results. As often as specified, calculation results accumulated by processes of the same node are summed using shared memory. Current random numbers are also saved. Then we do inter-node summation of results on the process of rank 0 with the MPI-Reduce function. Random numbers are collected with the MPI-Gather function. All this is done using a separate thread on each node and hence does not require main computations to stop. After calculated results and random numbers have been collected, the threads inactively sleep till next saving session. So, calculated data collection and saving does not interfere with main computations and consumes almost no CPU time. Routine calculations demonstrate a rather high parallelization efficiency which is no lower than 99.99 per cent
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.
Neutron spectrum unfolding using computer code SAIPS
Karim, S
1999-01-01
The main objective of this project was to study the neutron energy spectrum at rabbit station-1 in Pakistan Research Reactor (PARR-I). To do so, multiple foils activation method was used to get the saturated activities. The computer code SAIPS was used to unfold the neutron spectra from the measured reaction rates. Of the three built in codes in SAIPS, only SANDI and WINDOWS were used. Contribution of thermal part of the spectra was observed to be higher than the fast one. It was found that the WINDOWS gave smooth spectra while SANDII spectra have violet oscillations in the resonance region. The uncertainties in the WINDOWS results are higher than those of SANDII. The results show reasonable agreement with the published results.
Computer code for quantitative ALARA evaluations
International Nuclear Information System (INIS)
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
Computer code validation by high temperature chemistry
International Nuclear Information System (INIS)
At least five of the computer codes utilized in analysis of severe fuel damage-type events are directly dependent upon or can be verified by high temperature chemistry. These codes are ORIGEN, CORSOR, CORCON, VICTORIA, and VANESA. With the exemption of CORCON and VANESA, it is necessary that verification experiments be performed on real irradiated fuel. For ORIGEN, the familiar knudsen effusion cell is the best choice and a small piece of known mass and known burn-up is selected and volatilized completely into the mass spectrometer. The mass spectrometer is used in the integral mode to integrate the entire signal from preselected radionuclides, and from this integrated signal the total mass of the respective nuclides can be determined. For CORSOR and VICTORIA, experiments with flowing high pressure hydrogen/steam must flow over the irradiated fuel and then enter the mass spectrometer. For these experiments, a high pressure-high temperature molecular beam inlet must be employed. Finally, in support of VANESA-CORCON, the very highest temperature and molten fuels must be contained and analyzed. Results from all types of experiments will be discussed and their applicability to present and future code development will also be covered
Using Binary Code Instrumentation in Computer Security
Marius POPA; Sergiu Marin CAPISIZU
2013-01-01
The paper approaches the low-level details of the code generated by compilers whose format permits outside actions. Binary code modifications are manually done when the internal format is known and understood, or automatically by certain tools developed to process the binary code. The binary code instrumentation goals may be various from security increasing and bug fixing to development of malicious software. The paper highlights the binary code instrumentation techniques by code injection to...
Using Binary Code Instrumentation in Computer Security
Directory of Open Access Journals (Sweden)
Marius POPA
2013-01-01
Full Text Available The paper approaches the low-level details of the code generated by compilers whose format permits outside actions. Binary code modifications are manually done when the internal format is known and understood, or automatically by certain tools developed to process the binary code. The binary code instrumentation goals may be various from security increasing and bug fixing to development of malicious software. The paper highlights the binary code instrumentation techniques by code injection to increase the security and reliability of a software application. Also, the paper offers examples for binary code formats understanding and how the binary code injection may be applied.
Beam cavity interaction computer code for linacs
International Nuclear Information System (INIS)
A computer code BCI was used to calculate the energy loss of beam bunches in two types of cavities being studied in ORNL. The merit of the proposed on-axis coupled structure for CHEER is discussed in terms of beam energy loss. A calculation for a simple structure consisting of two accelerating cells and a coupling cell is performed. The amplitude modulation of the wake field is interpreted as the excitation of the zero and π coupled-resonator modes and coupling coefficients are derived. Beam loading of a cw linear accelerator is studied by calculating the bunch energy loss and cavity energy when a cavity is excited resonantly with a series of beam bunches. Calculations of non-resonant excitation are also presented
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. PMID:26601310
Computer codes for safety analysis of Indian PHWRs
International Nuclear Information System (INIS)
Computer codes for safety analysis of PHWRs have been developed in India over the years. Some of the codes that have been developed in NPC are discussed in this paper. Computer code THYNAC and ATMIKA have been developed in NPC for the analysis of LOCA scenario. Both the codes are based on UVET model using three equations and slip correlations. Computer code ATMIKA is an improved version of code THYNAC with regard to numerics and flexibility in modelling. Apart from thermal hydraulic model these codes also include point neutron kinetics model. Codes COOLTMP and RCOMP are used to estimate heat-up of primary coolant and core components respectively under off-normal shutdown conditions as may be existing during special maintenance job or postulated failure. Code validations have been performed either against experiments or the published results of experiments performed elsewhere, or through International benchmark exercises sponsored by IAEA. The paper discusses these codes, their validations and salient applications
ADORAVA - A computer code to sum random variables
International Nuclear Information System (INIS)
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.)
Development of an MCNP-tally based burnup code and validation through PWR benchmark exercises
International Nuclear Information System (INIS)
The aim of this study is to evaluate the capabilities of a newly developed burnup code called BUCAL1. The code provides the full capabilities of the Monte Carlo code MCNP5, through the use of the MCNP tally information. BUCAL1 uses the fourth order Runge Kutta method with the predictor-corrector approach as the integration method to determine the fuel composition at a desired burnup step. Validation of BUCAL1 was done by code vs. code comparison. Results of two different kinds of codes are employed. The first one is CASMO-4, a deterministic multi-group two-dimensional transport code. The second kind is MCODE and MOCUP, a link MCNP-ORIGEN codes. These codes use different burnup algorithms to solve the depletion equations system. Eigenvalue and isotope concentrations were compared for two PWR uranium and thorium benchmark exercises at cold (300 K) and hot (900 K) conditions, respectively. The eigenvalue comparison between BUCAL1 and the aforementioned two kinds of codes shows a good prediction of the systems'k-inf values during the entire burnup history, and the maximum difference is within 2%. The differences between the BUCAL1 isotope concentrations and the predictions of CASMO-4, MCODE and MOCUP are generally better, and only for a few sets of isotopes these differences exceed 10%.
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.
Computer codes validation for conditions of core voiding
International Nuclear Information System (INIS)
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)
Parallel-vector computation for CSI-design code
Nguyen, Duc T.
1990-01-01
Computational aspects of Control-Structure Interaction (CSI) DESIGN code is reviewed. Numerical intensive computation portions of CSI-DESIGN code were identified. Improvements in computational speed for the CSI-DESIGN code can be achieved by exploiting parallel and vector capabilities offered by modern computers, such as the Alliant, Convex, Cray-2, and Cray-YMP. Four options to generate the coefficient stiffness matrix and to solve the system of linear, simultaneous equations are currently available in the CSI-DESIGN code. A preprocessor to use RCM (Reverse Cuthill-Mackee) algorithm for bandwidth minimization was also developed for the CSI-DESIGN code. Preliminary results obtained by solving a small-scale, 97 node CSI finite element model (for eigensolution) have indicated that this new CSI-DESIGN code is 5 to 6 times faster (using 1 Alliant processor) than the old version of CSI-DESIGN code. This speed-up was achieved due to the RCM algorithm and the use of a new skyline solver. Efforts are underway to further improve the vector speed for CSI-DESIGN code, to evaluate its performance on a larger scale CSI model (such as phase zero CSI model) to make the code run efficiently on multiprocessor, parallel computer environment, and to make the code portable among different parallel computers available at NASA LaRC, such as Alliant, Convex, and Cray computers.
Calculations of angular momentum coupling coefficients on a computer code
International Nuclear Information System (INIS)
In this study, Clebsch-Gordan coefficients, 3j symbols, Racah coefficients, Wigner's 6j and 9j symbols were calculated by a prepared computer code of COEFF. The computer program COEFF is described which calculates angular momentum coupling coefficients and expresses them as quotient of two integers multiplied by the square root of the quotient of two integers. The program includes subroutines to encode an integer into its prime factors, to decode of prime factors back into an integer , to perform basic arithmetic operations on prime-coded numbers, as well as subroutines which calculate the coupling coefficients themselves. The computer code COEFF had been prepared to run on a VAX. In this study we rearranged the code to run on PC and tested it successfully. The obtained values in this study, were compared with the values of other computer programmes. A pretty good agreement is obtained between our prepared computer code and other computer programmes
Source Code Plagiarism in Computer Engineering Courses
Wolfgang Granzer; Friedrich Praus; Peter Balog
2013-01-01
In today’s university life, teachers are often confronted with plagiarism. A special form of plagiarism is source code plagiarism typically found in programming courses at universities and schools. Detecting or even preventing source code plagiarism is by no means a trivial task. Therefore, this paper explains and discusses different methods that can be used to prevent and detect source code plagiarism. The second part of this paper is focused on automatic tools that assist in detecting plagi...
Source Code Plagiarism in Computer Engineering Courses
Directory of Open Access Journals (Sweden)
Wolfgang Granzer
2013-08-01
Full Text Available In today’s university life, teachers are often confronted with plagiarism. A special form of plagiarism is source code plagiarism typically found in programming courses at universities and schools. Detecting or even preventing source code plagiarism is by no means a trivial task. Therefore, this paper explains and discusses different methods that can be used to prevent and detect source code plagiarism. The second part of this paper is focused on automatic tools that assist in detecting plagiarism. Finally, an approach is presented which can be used to detect source code plagiarism in PLC (programmable logic controller programs.
40 CFR 194.23 - Models and computer codes.
2010-07-01
... executing the computer codes, including hardware and software requirements, input and output formats with explanations of each input and output variable and parameter (e.g., parameter name and units); listings of input and output files from a sample computer run; and reports on code verification,...
The SEDA computer code and its utilization for Angra 1
International Nuclear Information System (INIS)
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)
ORNL ALICE: a statistical model computer code including fission competition
International Nuclear Information System (INIS)
A listing of the computer code ORNL ALICE is given. This code is a modified version of computer codes ALICE and OVERLAID ALICE. It allows for higher excitation energies and for a greater number of evaporated particles than the earlier versions. The angular momentum removal option was made more general and more internally consistent. Certain roundoff errors are avoided by keeping a strict accounting of partial probabilities. Several output options were added
Panel-Method Computer Code For Potential Flow
Ashby, Dale L.; Dudley, Michael R.; Iguchi, Steven K.
1992-01-01
Low-order panel method used to reduce computation time. Panel code PMARC (Panel Method Ames Research Center) numerically simulates flow field around or through complex three-dimensional bodies such as complete aircraft models or wind tunnel. Based on potential-flow theory. Facilitates addition of new features to code and tailoring of code to specific problems and computer-hardware constraints. Written in standard FORTRAN 77.
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......This paper investigated the possibility of intrinsic information conveying in network coding systems. The information is embedded into the coding vector by constructing the vector based on a set of predefined rules. This information can subsequently be retrieved by any receiver. The starting point...
Computer code for calculating reliability/availability of technical systems
International Nuclear Information System (INIS)
Three computer codes are reviewed, which can be applied to reliability analyses of technical systems. They are based on the fault tree and the laws of probability theory. The codes can be used for both non-repairable and repairable systems. The simulation code REMO 79 and the analytical code RELAV are based on the conception that a failure of system components is immediately detected and repaired. The model of the FUPRO2 code provides for failures to be detected and repaired only in periodic functional tests. Apart from code descriptions experience and far-reaching aspects resulting from modularization of the fault trees are summarized. (author)
Computer and compiler effects on code results: status report
International Nuclear Information System (INIS)
Within the framework of the international effort on the assessment of computer codes, which are designed to describe the overall reactor coolant system (RCS) thermalhydraulic response, core damage progression, and fission product release and transport during severe accidents, there has been a continuous debate as to whether the code results are influenced by different code users or by different computers or compilers. The first aspect, the 'Code User Effect', has been investigated already. In this paper the other aspects will be discussed and proposals are given how to make large system codes insensitive to different computers and compilers. Hardware errors and memory problems are not considered in this report. The codes investigated herein are integrated code systems (e. g. ESTER, MELCOR) and thermalhydraulic system codes with extensions for severe accident simulation (e. g. SCDAP/RELAP, ICARE/CATHARE, ATHLET-CD), and codes to simulate fission product transport (e. g. TRAPMELT, SOPHAEROS). Since all of these codes are programmed in Fortran 77, the discussion herein is based on this programming language although some remarks are made about Fortran 90. Some observations about different code results by using different computers are reported and possible reasons for this unexpected behaviour are listed. Then methods are discussed how to avoid portability problems
Computer codes for birds of North America
US Fish and Wildlife Service, Department of the Interior — Purpose of paper was to provide a more useful way to provide codes for all North American species, thus making the list useful for virtually all projects concerning...
Hanford Meteorological Station computer codes: Volume 8, The REVIEW computer code
Energy Technology Data Exchange (ETDEWEB)
Andrews, G.L.; Burk, K.W.
1988-08-01
The Hanford Meteorological Station (HMS) routinely collects meteorological data from sources on and off the Hanford Site. The data are averaged over both 15 minutes and 1 hour and are maintained in separate databases on the Digital Equipment Corporation (DEC) VAX 11/750 at the HMS. The databases are transferred to the Emergency Management System (EMS) DEC VAX 11/750 computer. The EMS is part of the Unified Dose Assessment Center, which is located on on the ground-level floor of the Federal building in Richland and operated by Pacific Northwest Laboratory. The computer program REVIEW is used to display meteorological data in graphical and alphanumeric form from either the 15-minute or hourly database. The code is available on the HMS and EMS computer. The REVIEW program helps maintain a high level of quality assurance on the instruments that collect the data and provides a convenient mechanism for analyzing meteorological data on a routine basis and during emergency response situations.
Low rank approximations for the DEPOSIT computer code
Litsarev, Mikhail; Oseledets, Ivan
2014-01-01
We present an efficient technique based on low-rank separated approximations for the computation of three-dimensional integrals in the computer code DEPOSIT that describes ion-atomic collision processes. Implementation of this technique decreases the total computational time by a factor of 1000. The general concept can be applied to more complicated models.
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.
High burnup models in computer code fair
International Nuclear Information System (INIS)
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
Continuous Materiality: Through a Hierarchy of Computational Codes
Directory of Open Access Journals (Sweden)
Jichen Zhu
2008-01-01
Full Text Available The legacy of Cartesian dualism inherent in linguistic theory deeply influences current views on the relation between natural language, computer code, and the physical world. However, the oversimplified distinction between mind and body falls short of capturing the complex interaction between the material and the immaterial. In this paper, we posit a hierarchy of codes to delineate a wide spectrum of continuous materiality. Our research suggests that diagrams in architecture provide a valuable analog for approaching computer code in emergent digital systems. After commenting on ways that Cartesian dualism continues to haunt discussions of code, we turn our attention to diagrams and design morphology. Finally we notice the implications a material understanding of code bears for further research on the relation between human cognition and digital code. Our discussion concludes by noticing several areas that we have projected for ongoing research.
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"…
Computer code qualification program for the Advanced CANDU Reactor
International Nuclear Information System (INIS)
Atomic Energy of Canada Ltd (AECL) has developed and implemented a Software Quality Assurance program (SQA) to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. This paper provides an overview of the computer programs used in Advanced CANDU Reactor (ACR) safety analysis, and assessment of their applicability in the safety analyses of the ACR design. An outline of the incremental validation program, and an overview of the experimental program in support of the code validation are also presented. An outline of the SQA program used to qualify these computer codes is also briefly presented. To provide context to the differences in the SQA with respect to current CANDUs, the paper also provides an overview of the ACR design features that have an impact on the computer code qualification. (author)
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.
Study of nuclear computer code maintenance and management system
International Nuclear Information System (INIS)
Software maintenance is one of the most important problems since late 1970's.We wish to develop a nuclear computer code system to maintenance and manage KAERI's nuclear software. As a part of this system, we have developed three code management programs for use on CYBER and PC systems. They are used in systematic management of computer code in KAERI. The first program is embodied on the CYBER system to rapidly provide information on nuclear codes to the users. The second and the third programs were embodied on the PC system for the code manager and for the management of data in korean language, respectively. In the requirement analysis, we defined each code, magnetic tape, manual and abstract information data. In the conceptual design, we designed retrieval, update, and output functions. In the implementation design, we described the technical considerations of database programs, utilities, and directions for the use of databases. As a result of this research, we compiled the status of nuclear computer codes which belonged KAERI until September, 1988. Thus, by using these three database programs, we could provide the nuclear computer code information to the users more rapidly. (Author)
Computer codes for level 1 probabilistic safety assessment
International Nuclear Information System (INIS)
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
Computer aided power flow software engineering and code generation
Energy Technology Data Exchange (ETDEWEB)
Bacher, R. [Swiss Federal Inst. of Tech., Zuerich (Switzerland)
1995-12-31
In this paper a software engineering concept is described which permits the automatic solution of a non-linear set of network equations. The power flow equation set can be seen as a defined subset of a network equation set. The automated solution process is the numerical Newton-Raphson solution process of the power flow equations where the key code parts are the numeric mismatch and the numeric Jacobian term computation. It is shown that both the Jacobian and the mismatch term source code can be automatically generated in a conventional language such as Fortran or C. Thereby one starts from a high level, symbolic language with automatic differentiation and code generation facilities. As a result of this software engineering process an efficient, very high quality Newton-Raphson solution code is generated which allows easier implementation of network equation model enhancements and easier code maintenance as compared to hand-coded Fortran or C code.
Computer aided power flow software engineering and code generation
Energy Technology Data Exchange (ETDEWEB)
Bacher, R. [Swiss Federal Inst. of Tech., Zuerich (Switzerland)
1996-02-01
In this paper a software engineering concept is described which permits the automatic solution of a non-linear set of network equations. The power flow equation set can be seen as a defined subset of a network equation set. The automated solution process is the numerical Newton-Raphson solution process of the power flow equations where the key code parts are the numeric mismatch and the numeric Jacobian term computation. It is shown that both the Jacobian and the mismatch term source code can be automatically generated in a conventional language such as Fortran or C. Thereby one starts from a high level, symbolic language with automatic differentiation and code generation facilities. As a result of this software engineering process an efficient, very high quality newton-Raphson solution code is generated which allows easier implementation of network equation model enhancements and easier code maintenance as compared to hand-coded Fortran or C code.
APC: A New Code for Atmospheric Polarization Computations
Korkin, Sergey V.; Lyapustin, Alexei I.; Rozanov, Vladimir V.
2014-01-01
A new polarized radiative transfer code Atmospheric Polarization Computations (APC) is described. The code is based on separation of the diffuse light field into anisotropic and smooth (regular) parts. The anisotropic part is computed analytically. The smooth regular part is computed numerically using the discrete ordinates method. Vertical stratification of the atmosphere, common types of bidirectional surface reflection and scattering by spherical particles or spheroids are included. A particular consideration is given to computation of the bidirectional polarization distribution function (BPDF) of the waved ocean surface.
Two-phase computer codes for zero-gravity applications
Energy Technology Data Exchange (ETDEWEB)
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.
Adaptation of HAMMER computer code to CYBER 170/750 computer
International Nuclear Information System (INIS)
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.)
A restructuring of CF package for MIDAS computer code
Energy Technology Data Exchange (ETDEWEB)
Park, S. H.; Kim, K. R.; Kim, D. H.; Cho, S. W. [KAERI, Taejon (Korea, Republic of)
2004-07-01
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.
A restructuring of CF package for MIDAS computer code
International Nuclear Information System (INIS)
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
Computer code ANISN multiplying media and shielding calculation II. Code description (input/output)
International Nuclear Information System (INIS)
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.)
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.
Computer code for intraply hybrid composite design
Chamis, C. C.; Sinclair, J. H.
1981-01-01
A computer program has been developed and is described herein for intraply hybrid composite design (INHYD). The program includes several composite micromechanics theories, intraply hybrid composite theories and a hygrothermomechanical theory. These theories provide INHYD with considerable flexibility and capability which the user can exercise through several available options. Key features and capabilities of INHYD are illustrated through selected samples.
Code system to compute radiation dose in human phantoms
International Nuclear Information System (INIS)
Monte Carlo photon transport code and a code using Monte Carlo integration of a point kernel have been revised to incorporate human phantom models for an adult female, juveniles of various ages, and a pregnant female at the end of the first trimester of pregnancy, in addition to the adult male used earlier. An analysis code has been developed for deriving recommended values of specific absorbed fractions of photon energy. The computer code system and calculational method are described, emphasizing recent improvements in methods
A program to validate computer codes for container impact analysis
International Nuclear Information System (INIS)
The detailed analysis of containers during impacts to assess either margins to failure or the consequences of different design strategies, requires the use of sophisticated computer codes to model the interactions of the various structural components. The combination of plastic deformation, impact and sliding at interfaces and dynamic loading effects provides a severe test of both the skill of the analyst and the robustness of the computer codes. A program of experiments has been under way at Winfrith since 1987 using extensively instrumented models to provide data for the validation of such codes. Three finite element codes, DYNA3D, HONDO-II and ABAQUS, were selected as suitable tools to cover the range of conditions expected in typical impacts. The impact orientation, velocity and instrumentation locations for the experiments are specified by pre-test calculations using these codes. Post-test analyses using the actual impact orientation and velocities are carried out as necessary if significant discrepancies are found
Computer vision cracks the leaf code.
Wilf, Peter; Zhang, Shengping; Chikkerur, Sharat; Little, Stefan A; Wing, Scott L; Serre, Thomas
2016-03-22
Understanding the extremely variable, complex shape and venation characters of angiosperm leaves is one of the most challenging problems in botany. Machine learning offers opportunities to analyze large numbers of specimens, to discover novel leaf features of angiosperm clades that may have phylogenetic significance, and to use those characters to classify unknowns. Previous computer vision approaches have primarily focused on leaf identification at the species level. It remains an open question whether learning and classification are possible among major evolutionary groups such as families and orders, which usually contain hundreds to thousands of species each and exhibit many times the foliar variation of individual species. Here, we tested whether a computer vision algorithm could use a database of 7,597 leaf images from 2,001 genera to learn features of botanical families and orders, then classify novel images. The images are of cleared leaves, specimens that are chemically bleached, then stained to reveal venation. Machine learning was used to learn a codebook of visual elements representing leaf shape and venation patterns. The resulting automated system learned to classify images into families and orders with a success rate many times greater than chance. Of direct botanical interest, the responses of diagnostic features can be visualized on leaf images as heat maps, which are likely to prompt recognition and evolutionary interpretation of a wealth of novel morphological characters. With assistance from computer vision, leaves are poised to make numerous new contributions to systematic and paleobotanical studies. PMID:26951664
Nuclear data to support computer code validation
International Nuclear Information System (INIS)
The rate of plutonium disposition will be a key parameter in determining the degree of success of the Fissile Materials Disposition Program. Estimates of the disposition rate are dependent on neutronics calculations. To ensure that these calculations are accurate, the codes and data should be validated against applicable experimental measurements. Further, before mixed-oxide (MOX) fuel can be fabricated and loaded into a reactor, the fuel vendors, fabricators, fuel transporters, reactor owners and operators, regulatory authorities, and the Department of Energy (DOE) must accept the validity of design calculations. This report presents sources of neutronics measurements that have potential application for validating reactor physics (predicting the power distribution in the reactor core), predicting the spent fuel isotopic content, predicting the decay heat generation rate, certifying criticality safety of fuel cycle facilities, and ensuring adequate radiation protection at the fuel cycle facilities and the reactor. The U.S. in-reactor experience with MOX fuel is first presented, followed by information related to other aspects of the MOX fuel performance information that is valuable to this program, but the data base remains largely proprietary. Thus, this information is not reported here. It is expected that the selected consortium will make the necessary arrangements to procure or have access to the requisite information
Theory Manual for ISAAC Computer Code
International Nuclear Information System (INIS)
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
A restructuring of RN2 package for MIDAS computer code
Energy Technology Data Exchange (ETDEWEB)
Park, S. H.; Kim, D. H. [KAERI, Taejon (Korea, Republic of)
2003-10-01
RN2 package, which is one of two fission product-related package in MELCOR, 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 data structure. To do this, data transferring methods of current MELCOR code are modified and adopted into the RN2 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. 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 RN2 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 validation has been done by comparing the results of the modified code with those from the existing code. As the trends are the 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.
A restructuring of RN1 package for MIDAS computer code
International Nuclear Information System (INIS)
RN1 package, which is one of two fission product-related packages in MELCOR, 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 RN1 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. 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 RN1 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
A restructuring of RN1 package for MIDAS computer code
Energy Technology Data Exchange (ETDEWEB)
Park, S. H.; Kim, D. H.; Kim, K. R. [KAERI, Taejon (Korea, Republic of)
2003-10-01
RN1 package, which is one of two fission product-related packages in MELCOR, 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 RN1 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. 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 RN1 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.
A restructuring of RN2 package for MIDAS computer code
International Nuclear Information System (INIS)
RN2 package, which is one of two fission product-related package in MELCOR, 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 data structure. To do this, data transferring methods of current MELCOR code are modified and adopted into the RN2 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. 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 RN2 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 validation has been done by comparing the results of the modified code with those from the existing code. As the trends are the 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
A restructuring of COR package for MIDAS computer code
International Nuclear Information System (INIS)
The COR package, which calculates the thermal response of the core and the lower plenum internal structures and models the relocation of the core and lower plenum structural materials, 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 a modernized data structure. To do this, the data transferring methods of the current MELCOR code are modified and adopted into the COR package. The data structure of the current MELCOR code using FORTRAN77 has a difficulty in grasping the meaning of the variables as well as a waste of memory. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type, which leads to an efficient memory treatment and an easy understanding of the code. Restructuring of the COR package addressed in this paper includes a module development, subroutine modification. 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 accelerated the code's domestication thanks to a direct understanding of each variable and an easy implementation of the modified or newly developed models. (author)
HUDU: The Hanford Unified Dose Utility computer code
Energy Technology Data Exchange (ETDEWEB)
Scherpelz, R.I.
1991-02-01
The Hanford Unified Dose Utility (HUDU) computer program was developed to provide rapid initial assessment of radiological emergency situations. The HUDU code uses a straight-line Gaussian atmospheric dispersion model to estimate the transport of radionuclides released from an accident site. For dose points on the plume centerline, it calculates internal doses due to inhalation and external doses due to exposure to the plume. The program incorporates a number of features unique to the Hanford Site (operated by the US Department of Energy), including a library of source terms derived from various facilities' safety analysis reports. The HUDU code was designed to run on an IBM-PC or compatible personal computer. The user interface was designed for fast and easy operation with minimal user training. The theoretical basis and mathematical models used in the HUDU computer code are described, as are the computer code itself and the data libraries used. Detailed instructions for operating the code are also included. Appendices to the report contain descriptions of the program modules, listings of HUDU's data library, and descriptions of the verification tests that were run as part of the code development. 14 refs., 19 figs., 2 tabs.
HUDU: The Hanford Unified Dose Utility computer code
International Nuclear Information System (INIS)
The Hanford Unified Dose Utility (HUDU) computer program was developed to provide rapid initial assessment of radiological emergency situations. The HUDU code uses a straight-line Gaussian atmospheric dispersion model to estimate the transport of radionuclides released from an accident site. For dose points on the plume centerline, it calculates internal doses due to inhalation and external doses due to exposure to the plume. The program incorporates a number of features unique to the Hanford Site (operated by the US Department of Energy), including a library of source terms derived from various facilities' safety analysis reports. The HUDU code was designed to run on an IBM-PC or compatible personal computer. The user interface was designed for fast and easy operation with minimal user training. The theoretical basis and mathematical models used in the HUDU computer code are described, as are the computer code itself and the data libraries used. Detailed instructions for operating the code are also included. Appendices to the report contain descriptions of the program modules, listings of HUDU's data library, and descriptions of the verification tests that were run as part of the code development. 14 refs., 19 figs., 2 tabs
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? &...
Preliminary blade design using integrated computer codes
Ryan, Arve
1988-12-01
Loads on the root of a horizontal axis wind turbine (HAWT) rotor blade were analyzed. A design solution for the root area is presented. The loads on the blades are given by different load cases that are specified. To get a clear picture of the influence of different parameters, the whole blade is designed from scratch. This is only a preliminary design study and the blade should not be looked upon as a construction reference. The use of computer programs for the design and optimization is extensive. After the external geometry is set and the aerodynamic loads calculated, parameters like design stresses and laminate thicknesses are run through the available programs, and a blade design optimized on basis of facts and estimates used is shown.
Low Computational Complexity Network Coding For Mobile Networks
DEFF Research Database (Denmark)
Heide, Janus
2012-01-01
Network Coding (NC) is a technique that can provide benefits in many types of networks, some examples from wireless networks are: In relay networks, either the physical or the data link layer, to reduce the number of transmissions. In reliable multicast, to reduce the amount of signaling and enable...... 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......-flow coding technique. One of the key challenges of this technique is its inherent computational complexity which can lead to high computational load and energy consumption in particular on the mobile platforms that are the target platform in this work. To increase the coding throughput several...
A restructuring of TF package for MIDAS computer code
Energy Technology Data Exchange (ETDEWEB)
Park, S. H.; Song, Y. M.; Kim, D. H. [KAERI, Taejon (Korea, Republic of)
2002-10-01
TF package which defines some interpolation and extrapolation condition through user defined table has been restructured in MIDAS computer code. To do this, data transferring methods of current MELCOR code are modified and adopted into TF package. The data structure of the current MELCOR code using FORTRAN77 causes a difficult grasping of 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 use the user-defined data type, which lead to an efficient memory treatment and an easy understanding of the code. Restructuring of TF package addressed in this paper does module development and subroutine modification, and treats MELGEN which is making restart file as well as MELCOR which is processing calculation. The validation has been done by comparing the results of the modified code with those from the existing code, and it is confirmed that the results are the same. It hints that the similar 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.
A restructuring of DCH package for MIDAS computer code
Energy Technology Data Exchange (ETDEWEB)
Park, S. H.; Kim, K. R.; Kim, D. H.; Cho, S. W. [KAERI, Taejon (Korea, Republic of)
2004-07-01
DCH package, which is one of thermal-hydraulic packages in MELCOR, 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. 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. 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 DCH package addressed in this paper includes module development, subroutine modification, and treats MELGEN, which generates an initial data file, as well as MELCOR, which is processing a calculation. The results of the modified code are verified against 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.
A restructuring of DCH package for MIDAS computer code
International Nuclear Information System (INIS)
DCH package, which is one of thermal-hydraulic packages in MELCOR, 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. 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. 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 DCH package addressed in this paper includes module development, subroutine modification, and treats MELGEN, which generates an initial data file, as well as MELCOR, which is processing a calculation. The results of the modified code are verified against 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
A restructuring of TF package for MIDAS computer code
International Nuclear Information System (INIS)
TF package which defines some interpolation and extrapolation condition through user defined table has been restructured in MIDAS computer code. To do this, data transferring methods of current MELCOR code are modified and adopted into TF package. The data structure of the current MELCOR code using FORTRAN77 causes a difficult grasping of 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 use the user-defined data type, which lead to an efficient memory treatment and an easy understanding of the code. Restructuring of TF package addressed in this paper does module development and subroutine modification, and treats MELGEN which is making restart file as well as MELCOR which is processing calculation. The validation has been done by comparing the results of the modified code with those from the existing code, and it is confirmed that the results are the same. It hints that the similar 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
Continuous Materiality: Through a Hierarchy of Computational Codes
Jichen Zhu; Kenneth J. Knoespe
2008-01-01
The legacy of Cartesian dualism inherent in linguistic theory deeply influences current views on the relation between natural language, computer code, and the physical world. However, the oversimplified distinction between mind and body falls short of capturing the complex interaction between the material and the immaterial. In this paper, we posit a hierarchy of codes to delineate a wide spectrum of continuous materiality. Our research suggests that diagrams in architecture provide a valuabl...
Sample test cases using the environmental computer code NECTAR
International Nuclear Information System (INIS)
This note demonstrates a few of the many different ways in which the environmental computer code NECTAR may be used. Four sample test cases are presented and described to show how NECTAR input data are structured. Edited output is also presented to illustrate the format of the results. Two test cases demonstrate how NECTAR may be used to study radio-isotopes not explicitly included in the code. (U.K.)
RADTRAN: a computer code to analyze transportation of radioactive material
International Nuclear Information System (INIS)
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
Problems of the thermal-hydraulic computer codes perfection
International Nuclear Information System (INIS)
The analysis of the current state of research and development in the field of thermal-hydraulic computer codes. The experience of the creation of domestic and foreign versions of the most advanced versions of code improved estimate. Considerable attention is paid to the problems of calculation of the critical heat fluxes in the channels of nuclear reactors. Considered problematic issues to ensure the reliability of thermal-hydraulic steam-generating channels in a thermoacoustic oscillation
COMPBRN III: a computer code for modeling compartment fires
International Nuclear Information System (INIS)
The computer code COMPBRN III deterministically models the behavior of compartment fires. This code is an improvement of the original COMPBRN codes. It employs a different air entrainment model and numerical scheme to estimate properties of the ceiling hot gas layer model. Moreover, COMPBRN III incorporates a number of improvements in shape factor calculations and error checking, which distinguish it from the COMPBRN II code. This report presents the ceiling hot gas layer model employed by COMPBRN III as well as several other modifications. Information necessary to run COMPBRN III, including descriptions of required input and resulting output, are also presented. Simulation of experiments and a sample problem are included to demonstrate the usage of the code. 37 figs., 46 refs
CRACKEL: a computer code for CFR fuel management calculations
International Nuclear Information System (INIS)
The CRACKLE computer code is designed to perform rapid fuel management surveys of CFR systems. The code calculates overall features such as reactivity, power distributions and breeding gain, and also calculates for each sub-assembly plutonium content and power output. A number of alternative options are built into the code, in order to permit different fuel management strategies to be calculated, and to perform more detailed calculations when necessary. A brief description is given of the methods of calculation, and the input facilities of CRACKLE, with examples. (author)
User`s manual for the NEFTRAN II computer code
Energy Technology Data Exchange (ETDEWEB)
Olague, N.E.; Campbell, J.E.; Leigh, C.D. [Sandia National Labs., Albuquerque, NM (USA); Longsine, D.E. [INTERA, Inc., Austin, TX (USA)
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.
User's manual for the NEFTRAN II computer code
International Nuclear Information System (INIS)
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
Important Cautions in Shielding Computation by using the FLUKA Code
Energy Technology Data Exchange (ETDEWEB)
Heo, Seung Uk; Oh, Sunju; Song, Yong keun; Kum, Oyeon [Korea Institute of Radiological and Medical Sciences, Seoul (Korea, Republic of); Nam, Sang Hee [Inje Univ.., Kimhae (Korea, Republic of)
2014-10-15
It is important to consider how efficiently the facility is designed against the shielding and activation problems in order to minimize the radiation exposure. In order to evaluate shielding capability of facility, although the simple calculation with approximate methods were popular until recently, owe to the development of information technology and the advances of computational mathematics, the Monte Carlo codes such as the MCNP, FLUKA, GEANT, and PHITS which can provide accurate answer are popularly used now. Advantage of Monte Carlo code is to perform the correct calculation but takes a long time for computing. More importantly, the exact and precise input data for Monte Carlo codes is essential in order to obtain accurate results. Thus, in this paper, important cautions are presented for shielding computation with the FLUKA code because the ignorance of such important cautions makes big troubles. The absorbed doses and errors show similar tendency in the comparison of groups, 1 and 2. Specifically, the results confirmed the more similar tendency in the high convergence areas. In group 3, although the comparison with groups, 1 and 2, shows the similar absorbed dose in the detectors with high convergences, the results themselves are unreliable because the errors are 99.9%. Thus, we need more careful attention to the average value and error in using the FLUKA code. Simply, it is better for us to have other benchmark tools such as MCNPX. However, it is recommended that the best computing method with the FLUKA code is the same as the computing of group 2, the usual multiprocessing with semi-automatic data handling. As shown in group 3, higher number of the cycle is a better method than the higher history to get more reliable result or to reduce errors. However, these values should be carefully evaluated.
Decontamination planning based on computer simulation code CDE
International Nuclear Information System (INIS)
Decontamination planning based on a computer simulation code CDE is discussed in this paper. Large amount of radionuclides had been discharged to environment in the accident of the Tokyo Electronic Power Company Fukushima Dai-ichi Nuclear Power Plant. CDE has been developed to support planning the decontamination. From the present study, it is validated that the computer simulation is very useful to predict the effect of the scenario before actions, and to plan the decontamination. (J.P.N.)
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)
Benchmarking Severe Accident Computer Codes for Heavy Water Reactor Applications
International Nuclear Information System (INIS)
Requests for severe accident investigations and assurance of mitigation measures have increased for operating nuclear power plants and the design of advanced nuclear power plants. Severe accident analysis investigations necessitate the analysis of the very complex physical phenomena that occur sequentially during various stages of accident progression. Computer codes are essential tools for understanding how the reactor and its containment might respond under severe accident conditions. The IAEA organizes coordinated research projects (CRPs) to facilitate technology development through international collaboration among Member States. The CRP on Benchmarking Severe Accident Computer Codes for HWR Applications was planned on the advice and with the support of the IAEA Nuclear Energy Department's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). This publication summarizes the results from the CRP participants. The CRP promoted international collaboration among Member States to improve the phenomenological understanding of severe core damage accidents and the capability to analyse them. The CRP scope included the identification and selection of a severe accident sequence, selection of appropriate geometrical and boundary conditions, conduct of benchmark analyses, comparison of the results of all code outputs, evaluation of the capabilities of computer codes to predict important severe accident phenomena, and the proposal of necessary code improvements and/or new experiments to reduce uncertainties. Seven institutes from five countries with HWRs participated in this CRP
A review of computer codes MODTURC-CLAS and PHOENICS
International Nuclear Information System (INIS)
This report provides a review of computer codes MODTURC-CLAS and PHOENICS, as applied to simulating the moderator flow inside the calandria of a CANDU nuclear reactor. It is concluded that the mathematical formulations of the codes account for the dominant physics of the moderator flows. However, weaknesses in these formulations include the pressure loss model for the calandria tube effects; the turbulence models which currently do not account for buoyancy effects, streamline curvature effects and low Reynolds number effects; and the resolution of the computational grids used: Two-dimensional simulations are in relatively good qualitative agreement with experimental data, although some quantitative differences warrant further investigation. It is recommended that additional verification of both two-and three-dimensional simulations be carried out with both codes. The problems identified with PHOENICS should, however, be corrected prior to testing it on other moderator flow situations. (author) 26 refs., 3 tabs., 16 figs
Mathematical models and computer code ELESIM used for CANDU reactors
International Nuclear Information System (INIS)
Candu reactors are used in many countries all over the world for power generation. This is because the reactors use natural uranium fuel, with simple design, which permits local manufacturing of the reactor components, in addition to safety in operation. The operation of Candu reactors is accompanied by highly sensitive automatic control loops, which in turn are accompanied by using a set of computer codes to simulate the components of the reactor. One of those codes is ELESIM, which is a computer program for simulating the behaviour of fuel element under the normal operating conditions. In this report, the most important phenomena modelled in ELESIM are manipulated in accordance to their dependence on each other. When necessary the mathematical model used in each item is given, while the equations used in the code is represented in appendix. 6 FIG
Methods and computer codes for nuclear systems calculations
Indian Academy of Sciences (India)
B P Kochurov; A P Knyazev; A Yu Kwaretzkheli
2007-02-01
Some numerical methods for reactor cell, sub-critical systems and 3D models of nuclear reactors are presented. The methods are developed for steady states and space–time calculations. Computer code TRIFON solves space-energy problem in (, ) systems of finite height and calculates heterogeneous few-group matrix parameters of reactor cells. These parameters are used as input data in the computer code SHERHAN solving the 3D heterogeneous reactor equation for steady states and 3D space–time neutron processes simulation. Modification of TRIFON was developed for the simulation of space–time processes in sub-critical systems with external sources. An option of SHERHAN code for the system with external sources is under development.
Plagiarism Detection Algorithm for Source Code in Computer Science Education
Liu, Xin; Xu, Chan; Ouyang, Boyu
2015-01-01
Nowadays, computer programming is getting more necessary in the course of program design in college education. However, the trick of plagiarizing plus a little modification exists among some students' home works. It's not easy for teachers to judge if there's plagiarizing in source code or not. Traditional detection algorithms cannot fit this…
Computer code for double beta decay QRPA based calculations
International Nuclear Information System (INIS)
The computer code developed by our group some years ago for the evaluation of nuclear matrix elements, within the QRPA and PQRPA nuclear structure models, involved in neutrino-nucleus reactions, muon capture and β± processes, is extended to include also the nuclear double beta decay
User's manual for the ORIGEN2 computer code
International Nuclear Information System (INIS)
This report describes how to use a revised version of the ORIGEN computer code, designated ORIGEN2. Included are a description of the input data, input deck organization, and sample input and output. ORIGEN2 can be obtained from the Radiation Shielding Information Center at ORNL
Method for quantitative assessment of nuclear safety computer codes
International Nuclear Information System (INIS)
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
Multilevel Coding Schemes for Compute-and-Forward
Hern, Brett
2010-01-01
We investigate techniques for designing modulation/coding schemes for the wireless two-way relaying channel. The relay is assumed to have perfect channel state information, but the transmitters are assumed to have no channel state information. We consider physical layer network coding based on multilevel coding techniques. Our multilevel coding framework is inspired by the compute-and-forward relaying protocol. Indeed, we show that the framework developed here naturally facilitates decoding of linear combinations of codewords for forwarding by the relay node. We develop our framework with general modulation formats in mind, but numerical results are presented for the case where each node transmits using the QPSK constellation with gray labeling. We focus our discussion on the rates at which the relay may reliably decode linear combinations of codewords transmitted from the end nodes.
Validation of Numerical Codes to Compute Tsunami Runup And Inundation
Velioğlu, Deniz; Cevdet Yalçıner, Ahmet; Kian, Rozita; Zaytsev, Andrey
2015-04-01
FLOW 3D and NAMI DANCE are two numerical codes which can be applied to analysis of flow and motion of long waves. Flow 3D simulates linear and nonlinear propagating surface waves as well as irregular waves including long waves. NAMI DANCE uses finite difference computational method to solve nonlinear shallow water equations (NSWE) in long wave problems, specifically tsunamis. Both codes can be applied to tsunami simulations and visualization of long waves. Both codes are capable of solving flooding problems. However, FLOW 3D is designed mainly to solve flooding problem from land and NAMI DANCE is designed to solve flooding problem from the sea. These numerical codes are applied to some benchmark problems for validation and verification. One useful benchmark problem is the runup of solitary waves which is investigated analytically and experimentally by Synolakis (1987). Since 1970s, solitary waves have commonly been used to model tsunamis especially in experimental and numerical studies. In this respect, a benchmark problem on runup of solitary waves is a relevant choice to assess the capability and validity of the numerical codes on amplification of tsunamis. In this study both codes have been tested, compared and validated by applying to the analytical benchmark problem of solitary wave runup on a sloping beach. Comparison of the results showed that both codes are in good agreement with the analytical and experimental results and thus can be proposed to be used in inundation of long waves and tsunami hazard analysis.
Verification and uncertainty analysis of fuel codes using distributed computing
International Nuclear Information System (INIS)
Of late, nuclear safety analysis computer codes have been held to increasingly high standards of quality assurance. As well, best estimate with uncertainty analysis is taking a more prominent role, displacing to some extent the idea of a limit consequence analysis. In turn, these activities have placed ever-increasing burdens on available computing resources. A recent project at Ontario Hydro has been the development of the capability of using the workstations on our Windows NT LAN as a distributed batch queue. The application developed is called SheepDog. This paper reports on the challenges and opportunities met in this project, as well as the experience gained in applying this method to verification and uncertainty analysis of fuel codes. SheepDog has been applied to performing uncertainty analysis, in a basically CSAU like method, of fuel behaviour during postulated accident scenarios at a nuclear power station. For each scenario, several hundred cases were selected according to a Latin Hypercube scheme, and used to construct a response surface surrogate for the codes. Residual disparities between code predictions and response surfaces led to the suspicion that there were discontinuities in the predictions of the analysis codes. This led to the development of 'stress testing' procedures. This refers to two procedures: coarsely scanning through several input parameters in combination, and finely scanning individual input parameters. For either procedure, the number of code runs required is several hundred. In order to be able to perform stress testing in a reasonable time, SheepDog was applied. The results are examined for such considerations as continuity, smoothness, and physical reasonableness of trends and interactions. In several cases, this analysis uncovered previously unknown errors in analysis codes, and allowed pinpointing the part of the codes that needed to be modified. The challenges involved include the following: the usual choices of development
Computed radiography simulation using the Monte Carlo code MCNPX
International Nuclear Information System (INIS)
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)
Numerical computation of molecular integrals via optimized (vectorized) FORTRAN code
International Nuclear Information System (INIS)
The calculation of molecular properties based on quantum mechanics is an area of fundamental research whose horizons have always been determined by the power of state-of-the-art computers. A computational bottleneck is the numerical calculation of the required molecular integrals to sufficient precision. Herein, we present a method for the rapid numerical evaluation of molecular integrals using optimized FORTRAN code generated by Maple. The method is based on the exploitation of common intermediates and the optimization can be adjusted to both serial and vectorized computations. (orig.)
Verification of fuel performance simulation codes: application of distributed computing
International Nuclear Information System (INIS)
The Canadian Nuclear Industry as a whole is attempting to find accurate, efficient and cost-effective methods to either test or demonstrate that complex simulation codes are behaving as they were intended to. One method that has been used at Ontario Power Generation is 'stress testing': running the code or code suite a sufficient number of times with small variations in input data, and tracking the resulting change in calculated results. This technique has only recently become practical for large simulation codes with the increasing power of desktop computers and the ability to dispatch batch jobs across a network; the method used for the work performed to date is described in detail in. The stress testing methodology is a balance between safety analysis needs and software engineering principles. It is not in general possible to completely cover the range of all input parameters and their interactions in a reasonable amount of time. Hence the number of input parameters to test was restricted to those that have the largest impact on the key output results (those outputs that are actually used directly or indirectly as safety criteria). Of particular value in identifying code deficiencies was 'parameter scanning', where one input parameter was varied using a fine increment over a range encompassing the normal variation. The key output results were then plotted against the varying input parameter, and the results studied to ensure that the predicted trends were correct and no non-physical discontinuities were present. This paper reports the results from two such stress testing exercises as examples of applications of the stress testing method. One application is on the ELESIM/ELOCA code suite to show how stress testing can identify problems in code testing and verification. The other application is from the FACTAR code suite focusing on how the stress testing method can be applied to a large code suite. (author)
Fault-tolerant quantum computing with color codes
Landahl, Andrew J; Rice, Patrick R
2011-01-01
We present and analyze protocols for fault-tolerant quantum computing using color codes. We present circuit-level schemes for extracting the error syndrome of these codes fault-tolerantly. We further present an integer-program-based decoding algorithm for identifying the most likely error given the syndrome. We simulated our syndrome extraction and decoding algorithms against three physically-motivated noise models using Monte Carlo methods, and used the simulations to estimate the corresponding accuracy thresholds for fault-tolerant quantum error correction. We also used a self-avoiding walk analysis to lower-bound the accuracy threshold for two of these noise models. We present and analyze two architectures for fault-tolerantly computing with these codes: one with 2D arrays of qubits are stacked atop each other and one in a single 2D substrate. Our analysis demonstrates that color codes perform slightly better than Kitaev's surface codes when circuit details are ignored. When these details are considered, w...
A DOE Computer Code Toolbox: Issues and Opportunities
International Nuclear Information System (INIS)
The initial activities of a Department of Energy (DOE) Safety Analysis Software Group to establish a Safety Analysis Toolbox of computer models are discussed. The toolbox shall be a DOE Complex repository of verified and validated computer models that are configuration-controlled and made available for specific accident analysis applications. The toolbox concept was recommended by the Defense Nuclear Facilities Safety Board staff as a mechanism to partially address Software Quality Assurance issues. Toolbox candidate codes have been identified through review of a DOE Survey of Software practices and processes, and through consideration of earlier findings of the Accident Phenomenology and Consequence Evaluation program sponsored by the DOE National Nuclear Security Agency/Office of Defense Programs. Planning is described to collect these high-use codes, apply tailored SQA specific to the individual codes, and implement the software toolbox concept. While issues exist such as resource allocation and the interface among code developers, code users, and toolbox maintainers, significant benefits can be achieved through a centralized toolbox and subsequent standardized applications
Some numerical results with the COMMIX-2 computer code
International Nuclear Information System (INIS)
The computer code COMMIX-2 has been developed for analyzing and designing thermal-hydraulic aspects of nuclear reactor components. The code employs a two-fluid model for solving transient, three-dimensional two-phase (or single phase) nonhomogeneous and nonequilibrium flow conditions. The report presents numerical results of four problems selected to demonstrate the capabilities of COMMIX-2: (1) transient single-phase flow with heat source; (2) two-phase flow in a vertical tube, where the surface heat flux is sufficiently high that a single-phase liquid emerges as a mixture of liquid and vapor; (3) separation of vapor and liquid; and (4) a high-pressure jet impinging on a vertical plate. The third and fourth problems were selected to demonstrate, respectively, that the code can handle computational difficulties usually encountered in problems with sharp interfaces, and the important role of interfacial mass and momentum exhange. The numerical results obtained by COMMIX-2 code are very encouraging. It has not only demonstrated the computational capability but has also exhibited the ability of modeling complex phenomena of the jet impingement problems with very simple interfacial drag and evaporation models
LMFBR models for the ORIGEN2 computer code
International Nuclear Information System (INIS)
Reactor physics calculations have led to the development of nine liquid-metal fast breeder reactor (LMFBR) models for the ORIGEN2 computer code. Four of the models are based on the U-Pu fuel cycle, two are based on the Th-U-Pu fuel cycle, and three are based on the Th-238U fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST are given
LMFBR models for the ORIGEN2 computer code
Energy Technology Data Exchange (ETDEWEB)
Croff, A.G.; McAdoo, J.W.; Bjerke, M.A.
1981-10-01
Reactor physics calculations have led to the development of nine liquid-metal fast breeder reactor (LMFBR) models for the ORIGEN2 computer code. Four of the models are based on the U-Pu fuel cycle, two are based on the Th-U-Pu fuel cycle, and three are based on the Th-/sup 238/U fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST are given.
Computer codes for evaluation of control room habitability (HABIT)
International Nuclear Information System (INIS)
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
Computer codes for evaluation of control room habitability (HABIT)
Energy Technology Data Exchange (ETDEWEB)
Stage, S.A. [Pacific Northwest Lab., Richland, WA (United States)
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.
The 3D-SEEP computer code user's manual
International Nuclear Information System (INIS)
This report describes the 3D-SEEP computer code and presents the direction to use the code effectively. 3D-SEEP calculates the saturated-unsaturated time dependent or steady state flow of groundwater in permeable geologic media for the safety evaluation of nuclear waste disposal. 3D-SEEP is based on the 3-dimensional Galerkin finite element method. This allows the modeling of complex geometrical shapes and complicated patterns of geologic media. The flow is modeled by single phase flow governed by Darcy's law, and the simplified double porosity model is introduced to consider fractured media. This code can handle non-uniform flow regions having irregular boundaries and arbitrary degree of local anisotropy. (author)
PWRDYN: a computer code for PWR plant dynamic analysis
International Nuclear Information System (INIS)
This report describes analytical models and calculated results of a PWR plant dynamic analysis code PWRDYN. The code has been developed in order to analyze and evaluate transient responses for small disturbance such as operating mode change and control system characteristic analysis. The features included in PWRDYN are 1) One loop approximation of primary loops, 2) Praimary coolant is always subcooled, 3) At the secondary side of steam generator is used one dimensional model and natural circulation is calculated assuming constant by positive driving head. 4) Main control systems are incorporated. In the transient responses caused by small perturbation, the calculated results by PWRDYN are in good agreement with the RETRAN calculations. Furthermore, computing time is very short so as about one seventh of real time, hence the code is convenient and useful for dynamic analysis of PWR plants. (author)
Evaluation of the FRAPTRAN -1.3 Computer Code
Energy Technology Data Exchange (ETDEWEB)
Manngaard, Tero [Quantum Technologies AB, Uppsala Science Park, SE-751 83 Uppsala (Sweden)
2007-03-15
The FRAPTRAN-1.3 computer code has been evaluated regarding its applicability, modelling 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 reactor power and coolant transients, such as overpower transients, reactivity initiated accidents (RIA), boiling-water reactor power oscillations without scram, and loss of coolant accidents (LOCA). Its experimental database covers boiling- and pressurized water reactor fuel rods with UO{sub 2} fuel up to rod burnups around 64 MWd/kgU. In FRAPTRAN-1.3, the fundamental equations for heat transfer and structural analysis are solved in one-dimensional (in the radial direction) and transient (time-dependent) form, and interaction between axial segments of the rod is confined to calculations of coolant axial flow, rod internal gas pressure and optionally axial flow of fission gases. The clad-to-coolant heat transfer conditions can either be specified as pre-calculated data or can be determined by a coolant channel model in the code. The code provides different clad rupture models depending on cladding temperature and amount of cladding plastic hoop strain. For LOCA analysis, a model calculating local clad shape (ballooning) and associated local stresses is available to predict clad burst. A strain based failure model is present for cladding rupture driven by pellet-cladding mechanical interaction. Two models exist for computation of high-temperature clad oxidation under LOCA (i) the Baker-Just model for licensing calculations and (ii) the Cathcart-Pawel model for best-estimate calculations. The code appears to be fairly easy to use, however, the applicability of the current version as a self-standing analysis tool for LOCA and RIA analyses depends highly on the numerical robustness of the coolant channel model for generation of clad-to-coolant heat transfer boundary conditions. The main
Evaluation of the FRAPTRAN -1.3 Computer Code
International Nuclear Information System (INIS)
The FRAPTRAN-1.3 computer code has been evaluated regarding its applicability, modelling 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 reactor power and coolant transients, such as overpower transients, reactivity initiated accidents (RIA), boiling-water reactor power oscillations without scram, and loss of coolant accidents (LOCA). Its experimental database covers boiling- and pressurized water reactor fuel rods with UO2 fuel up to rod burnups around 64 MWd/kgU. In FRAPTRAN-1.3, the fundamental equations for heat transfer and structural analysis are solved in one-dimensional (in the radial direction) and transient (time-dependent) form, and interaction between axial segments of the rod is confined to calculations of coolant axial flow, rod internal gas pressure and optionally axial flow of fission gases. The clad-to-coolant heat transfer conditions can either be specified as pre-calculated data or can be determined by a coolant channel model in the code. The code provides different clad rupture models depending on cladding temperature and amount of cladding plastic hoop strain. For LOCA analysis, a model calculating local clad shape (ballooning) and associated local stresses is available to predict clad burst. A strain based failure model is present for cladding rupture driven by pellet-cladding mechanical interaction. Two models exist for computation of high-temperature clad oxidation under LOCA (i) the Baker-Just model for licensing calculations and (ii) the Cathcart-Pawel model for best-estimate calculations. The code appears to be fairly easy to use, however, the applicability of the current version as a self-standing analysis tool for LOCA and RIA analyses depends highly on the numerical robustness of the coolant channel model for generation of clad-to-coolant heat transfer boundary conditions. The main documentation
SCRIMP. A thermal-hydraulic subchannel analysis computer code
International Nuclear Information System (INIS)
SCRIMP is a FORTRAN IV computer code for calculating pressure drop, flow rates, heat transfer rates and temperatures in heat exchangers such as fuel elements of typical gas cooled nuclear reactors, under steady state conditions. The subchannel analysis computer code SCRIMP is an improved version of the SCEPTIC program. The most important modification is the introduction of a new subroutine FASTCAL for the friction factor and heat transfer coefficient calculations. The different boundary conditions of the subchannels such as geometry changes, quality of surfaces, heat flux variation and unheated wall are considered in each particular case by using this subroutine. Due to its great flexibility, particularly with respect to geometrical arrangement, and the relatively short calculation time, SCRIMP is a very useful tool to analyze a variety of thermohydraulic problems. (Auth.)
Benchmarking of computer codes and approaches for modeling exposure scenarios
International Nuclear Information System (INIS)
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
Optimization of Russian roulette parameters for the KENO computer code
Energy Technology Data Exchange (ETDEWEB)
Hoffman, T.J.
1982-10-01
Proper specification of the (statistical) weight standards for Monte Carlo calculations can lead to a substantial reduction in computer time. Frequently these weights are set intuitively. When optimization is performed, it is usually based on a simplified model (to enable mathematical analysis) and involves minimization of the sample variance. In this report, weight standards are optimized through consideration of the actual implementation of Russian roulette in the KENO computer code. The goal is minimization of computer time rather than minimization of sample variance. Verification of the development and assumptions is obtained from Monte Carlo simulations. The results indicate that the current default weight standards are appropriate for most problems in which thermal neutron transport is not a major consumer of computer time. For thermal systems, the optimization technique described in this report should be used.
Verification of structural analysis computer codes in nuclear engineering
International Nuclear Information System (INIS)
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)
Nuclear shell-model code for massive parallel computation, "KSHELL"
Shimizu, Noritaka
2013-01-01
A new code for nuclear shell-model calculations, "KSHELL", is developed. It aims at carrying out both massively parallel computation and single-node computation in the same manner. We solve the Schr\\"{o}dinger's equation in the $M$-scheme shell-model model space, utilizing Thick-Restart Lanczos method. During the Lanczos iteration, the whole Hamiltonian matrix elements are generated "on-the-fly" in every matrix-vector multiplication. The vectors of the Lanczos method are distributed and store...
War of ontology worlds: mathematics, computer code, or Esperanto?
Andrey Rzhetsky; 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 an...
Bragg optics computer codes for neutron scattering instrument design
Energy Technology Data Exchange (ETDEWEB)
Popovici, M.; Yelon, W.B.; Berliner, R.R. [Missouri Univ. Research Reactor, Columbia, MO (United States); Stoica, A.D. [Institute of Physics and Technology of Materials, Bucharest (Romania)
1997-09-01
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)
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
Parallel computing by Monte Carlo codes MVP/GMVP
International Nuclear Information System (INIS)
General-purpose Monte Carlo codes MVP/GMVP are well-vectorized and thus enable us to perform high-speed Monte Carlo calculations. In order to achieve more speedups, we parallelized the codes on the different types of parallel computing platforms or by using a standard parallelization library MPI. The platforms used for benchmark calculations are a distributed-memory vector-parallel computer Fujitsu VPP500, a distributed-memory massively parallel computer Intel paragon and a distributed-memory scalar-parallel computer Hitachi SR2201, IBM SP2. As mentioned generally, linear speedup could be obtained for large-scale problems but parallelization efficiency decreased as the batch size per a processing element(PE) was smaller. It was also found that the statistical uncertainty for assembly powers was less than 0.1% by the PWR full-core calculation with more than 10 million histories and it took about 1.5 hours by massively parallel computing. (author)
Computer codes for the analysis of flask impact problems
International Nuclear Information System (INIS)
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)
A study on the nuclear computer code maintenance and management system
International Nuclear Information System (INIS)
According to current software development and quality assurance trends. It is necessary to develop computer code management system for nuclear programs. For this reason, the project started in 1987. Main objectives of the project are to establish a nuclear computer code management system, to secure software reliability, and to develop nuclear computer code packages. Contents of performing the project in this year were to operate and maintain computer code information system of KAERI computer codes, to develop application tool, AUTO-i, for solving the 1st and 2nd moments of inertia on polygon or circle, and to research nuclear computer code conversion between different machines. For better supporting the nuclear code availability and reliability, assistance from users who are using codes is required. Lastly, for easy reference about the codes information, we presented list of code names and information on the codes which were introduced or developed during this year. (Author)
A computer code for computing the beam profiles in the NBI beam line 'BEMPROF'
International Nuclear Information System (INIS)
A computer code was developed which can compute the beam profiles and the percentage heat loadings on the various components in the NBI beam line such as the beam target, the beam limiters and the calorimeter. The geometrical injection efficiency of NBI and the heat input pattern on the counter surface of the injection port of the torus can also be computed. The major feature of this code is that the effects of the beamlet intensity distribution, the beamlet deflection, the beam screening by the upstream limiters and also the plasma density distribution and the divergence angle distribution over the beam extraction area can be taken into account. (author)
ABINIT: a computer code for matter; Abinit: un code au service de la matiere
Energy Technology Data Exchange (ETDEWEB)
Amadon, B.; Bottin, F.; Bouchet, J.; Dewaele, A.; Jollet, F.; Jomard, G.; Loubeyre, P.; Mazevet, S.; Recoules, V.; Torrent, M.; Zerah, G. [CEA Bruyeres-le-Chatel, 91 (France)
2008-07-01
The PAW (Projector Augmented Wave) method has been implemented in the ABINIT Code that computes electronic structures in atoms. This method relies on the simultaneous use of a set of auxiliary functions (in plane waves) and a sphere around each atom. This method allows the computation of systems including many atoms and gives the expression of energy, forces, stress... in terms of the auxiliary function only. We have generated atomic data for iron at very high pressure (over 200 GPa). We get a bcc-hcp transition around 10 GPa and the magnetic order disappears around 50 GPa. This method has been validated on a series of metals. The development of the PAW method has required a great effort for the massive parallelization of the ABINIT code. (A.C.)
International Nuclear Information System (INIS)
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
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.
Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis
International Nuclear Information System (INIS)
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
WSRC approach to validation of criticality safety computer codes
Energy Technology Data Exchange (ETDEWEB)
Finch, D.R.; Mincey, J.F.
1991-12-31
Recent hardware and operating system changes at Westinghouse Savannah River Site (WSRC) have necessitated review of the validation for JOSHUA criticality safety computer codes. As part of the planning for this effort, a policy for validation of JOSHUA and other criticality safety codes has been developed. This policy will be illustrated with the steps being taken at WSRC. The objective in validating a specific computational method is to reliably correlate its calculated neutron multiplication factor (K{sub eff}) with known values over a well-defined set of neutronic conditions. Said another way, such correlations should be: (1) repeatable; (2) demonstrated with defined confidence; and (3) identify the range of neutronic conditions (area of applicability) for which the correlations are valid. The general approach to validation of computational methods at WSRC must encompass a large number of diverse types of fissile material processes in different operations. Special problems are presented in validating computational methods when very few experiments are available (such as for enriched uranium systems with principal second isotope {sup 236}U). To cover all process conditions at WSRC, a broad validation approach has been used. Broad validation is based upon calculation of many experiments to span all possible ranges of reflection, nuclide concentrations, moderation ratios, etc. Narrow validation, in comparison, relies on calculations of a few experiments very near anticipated worst-case process conditions. The methods and problems of broad validation are discussed.
WSRC approach to validation of criticality safety computer codes
Energy Technology Data Exchange (ETDEWEB)
Finch, D.R.; Mincey, J.F.
1991-01-01
Recent hardware and operating system changes at Westinghouse Savannah River Site (WSRC) have necessitated review of the validation for JOSHUA criticality safety computer codes. As part of the planning for this effort, a policy for validation of JOSHUA and other criticality safety codes has been developed. This policy will be illustrated with the steps being taken at WSRC. The objective in validating a specific computational method is to reliably correlate its calculated neutron multiplication factor (K{sub eff}) with known values over a well-defined set of neutronic conditions. Said another way, such correlations should be: (1) repeatable; (2) demonstrated with defined confidence; and (3) identify the range of neutronic conditions (area of applicability) for which the correlations are valid. The general approach to validation of computational methods at WSRC must encompass a large number of diverse types of fissile material processes in different operations. Special problems are presented in validating computational methods when very few experiments are available (such as for enriched uranium systems with principal second isotope {sup 236}U). To cover all process conditions at WSRC, a broad validation approach has been used. Broad validation is based upon calculation of many experiments to span all possible ranges of reflection, nuclide concentrations, moderation ratios, etc. Narrow validation, in comparison, relies on calculations of a few experiments very near anticipated worst-case process conditions. The methods and problems of broad validation are discussed.
Improved Flow Modeling in Transient Reactor Safety Analysis Computer Codes
International Nuclear Information System (INIS)
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)
MAGNUM-2D computer code: user's guide
Energy Technology Data Exchange (ETDEWEB)
England, R.L.; Kline, N.W.; Ekblad, K.J.; Baca, R.G.
1985-01-01
Information relevant to the general use of the MAGNUM-2D computer code is presented. This computer code was developed for the purpose of modeling (i.e., simulating) the thermal and hydraulic conditions in the vicinity of a waste package emplaced in a deep geologic repository. The MAGNUM-2D computer computes (1) the temperature field surrounding the waste package as a function of the heat generation rate of the nuclear waste and thermal properties of the basalt and (2) the hydraulic head distribution and associated groundwater flow fields as a function of the temperature gradients and hydraulic properties of the basalt. MAGNUM-2D is a two-dimensional numerical model for transient or steady-state analysis of coupled heat transfer and groundwater flow in a fractured porous medium. The governing equations consist of a set of coupled, quasi-linear partial differential equations that are solved using a Galerkin finite-element technique. A Newton-Raphson algorithm is embedded in the Galerkin functional to formulate the problem in terms of the incremental changes in the dependent variables. Both triangular and quadrilateral finite elements are used to represent the continuum portions of the spatial domain. Line elements may be used to represent discrete conduits. 18 refs., 4 figs., 1 tab.
WSRC approach to validation of criticality safety computer codes
International Nuclear Information System (INIS)
Recent hardware and operating system changes at Westinghouse Savannah River Site (WSRC) have necessitated review of the validation for JOSHUA criticality safety computer codes. As part of the planning for this effort, a policy for validation of JOSHUA and other criticality safety codes has been developed. This policy will be illustrated with the steps being taken at WSRC. The objective in validating a specific computational method is to reliably correlate its calculated neutron multiplication factor (Keff) with known values over a well-defined set of neutronic conditions. Said another way, such correlations should be: (1) repeatable; (2) demonstrated with defined confidence; and (3) identify the range of neutronic conditions (area of applicability) for which the correlations are valid. The general approach to validation of computational methods at WSRC must encompass a large number of diverse types of fissile material processes in different operations. Special problems are presented in validating computational methods when very few experiments are available (such as for enriched uranium systems with principal second isotope 236U). To cover all process conditions at WSRC, a broad validation approach has been used. Broad validation is based upon calculation of many experiments to span all possible ranges of reflection, nuclide concentrations, moderation ratios, etc. Narrow validation, in comparison, relies on calculations of a few experiments very near anticipated worst-case process conditions. The methods and problems of broad validation are discussed
Probabilistic evaluations for CANTUP computer code analysis improvement
International Nuclear Information System (INIS)
Structural analysis with finite element method is today an usual way to evaluate and predict the behavior of structural assemblies subject to hard conditions in order to ensure their safety and reliability during their operation. A CANDU 600 fuel channel is an example of an assembly working in hard conditions, in which, except the corrosive and thermal aggression, long time irradiation, with implicit consequences on material properties evolution, interferes. That leads inevitably to material time-dependent properties scattering, their dynamic evolution being subject to a great degree of uncertainness. These are the reasons for developing, in association with deterministic evaluations with computer codes, the probabilistic and statistical methods in order to predict the structural component response. This work initiates the possibility to extend the deterministic thermomechanical evaluation on fuel channel components to probabilistic structural mechanics approach starting with deterministic analysis performed with CANTUP computer code which is a code developed to predict the long term mechanical behavior of the pressure tube - calandria tube assembly. To this purpose the structure of deterministic calculus CANTUP computer code has been reviewed. The code has been adapted from LAHEY 77 platform to Microsoft Developer Studio - Fortran Power Station platform. In order to perform probabilistic evaluations, it was added a part to the deterministic code which, using a subroutine from IMSL library from Microsoft Developer Studio - Fortran Power Station platform, generates pseudo-random values of a specified value. It was simulated a normal distribution around the deterministic value and 5% standard deviation for Young modulus material property in order to verify the statistical calculus of the creep behavior. The tube deflection and effective stresses were the properties subject to probabilistic evaluation. All the values of these properties obtained for all the values for
Interactive computer code for dynamic and soil structure interaction analysis
Energy Technology Data Exchange (ETDEWEB)
Mulliken, J.S.
1995-12-01
A new interactive computer code is presented in this paper for dynamic and soil-structure interaction (SSI) analyses. The computer program FETA (Finite Element Transient Analysis) is a self contained interactive graphics environment for IBM-PC`s that is used for the development of structural and soil models as well as post-processing dynamic analysis output. Full 3-D isometric views of the soil-structure system, animation of displacements, frequency and time domain responses at nodes, and response spectra are all graphically available simply by pointing and clicking with a mouse. FETA`s finite element solver performs 2-D and 3-D frequency and time domain soil-structure interaction analyses. The solver can be directly accessed from the graphical interface on a PC, or run on a number of other computer platforms.
Approaches for computing uncertainties in predictions of complex-codes
International Nuclear Information System (INIS)
Uncertainty analysis aims at characterizing the errors associated with experiments and predictions of computer codes, in contradistinction with sensitivity analysis, which aims at determining the rate of change (i.e., derivative) in the predictions of codes when one or more (typically uncertain) input parameters varies within its range of interest. This paper reviews the salient features of three independent approaches for estimating uncertainties associated with predictions of complex system codes. The first approach reviewed in this paper as the prototype for propagation of code input errors is the so-called 'GRS method', which includes the so-called 'CSAU method' (Code Scaling, Applicability and Uncertainty) and the majority of methods adopted by the nuclear industry. Although the entire set of the actual number of input parameters for a typical NPP (Nuclear Power Plant) input deck, ranging up to about 105 input parameters, could theoretically be considered as uncertainty sources by these methods, only a 'manageable' number (of the order of several tens) is actually taken into account in practice. Ranges of variations, together with suitable PDF (Probability Density Function) are then assigned for each of the uncertain input parameter actually considered in the analysis. The number of computations using the code under investigation needed for obtaining the desired confidence in the results can be determined theoretically (it is of the order of 100). Subsequently, an additional number of computations (ca. 100) with the code are performed to propagate the uncertainties inside the code, from inputs to outputs (results). The second approach reviewed in this paper is the propagation of code output errors, as representatively illustrated by the UMAE-CIAU (Uncertainty Method based upon Accuracy Extrapolation 'embedded' into the Code with capability of Internal Assessment of Uncertainty). Note that this class of methods includes only a few applications from industry
Recommended documentation plan for the FLAG and CHEMFLUB computer codes
Energy Technology Data Exchange (ETDEWEB)
None
1983-09-02
Reviews have been conducted on both FLAG and CHEMFLUB's documentation and computer codes. The documentation of both models is: (1) incomplete, (2) confusing, (3) not helpful to the reader, (4) filled with extraneous information and (5) lack claimed versatility in analyzing coal gasifier systems. The documentation is such that the computer coding itself must be used as a reference to complete the documentation. Once the codes are set up they are relatively easy to run. We have exercised both of them. Most of our efforts thus far have been concentrated on FLAG because of its importance and complexity. FLAG in its present form can not be expected to yield meaningful data applicable to coal gasifier systems. The reasons for this are twofold. First, the model is incorrect in describing some aspects of fluid particle behavior in coal gasifier systems. Second, the numerical formulation/solution methodology is incorrectly implemented and introduces spurious numerical effects, thereby obscuring the physics of the model. In brief, this means that resulting calculations are not correctly related to the physics. CHEMFLUB, while less extensively exercised, shows that it should be no surprise that CHEMFLUB is best utilized as a tool for generating first approximations. We have concluded from these reviews that we cannot perform meaningful comparisons as required under tasks 3.3, 3.4, and 3.5 without first reconstructing and correcting when necessary the physical/numerical models. A plan is presented for accomplishing this reconstruction/modification.
Internal pump reactor stability: Qualification of TRACG computer code
International Nuclear Information System (INIS)
Stability tests of the Forsmark Unit 1 reactor were conducted in January 1989 during power and flow conditions which put the plant at the onset of limit cycle instabilities. Studies of the LPRM recordings indicate the predominant mode to be global ''in phase'' oscillation. (The core made a few attempts to break away from the in phase pattern, the out of phase mode lasted only for two oscillation periods.) Sufficient data was taken to quantify the decay ratio and measure the effect of both small and large, power and flow changes. The axial power distribution also varied during the tests due to xenon transients. A computer model of the Forsmark Unit 1 has been constructed for use with TRACG, the GE Nuclear Energy version of the TRAC-BD1 computer code. TRACG has been thoroughly validated for application in loss of coolant, anticipated transients and stability analysis. The stability qualification data base has previously consisted of jet pump BWRs. The additional flow restriction of internal recirculation pumps generally causes such reactors to have more limiting stability. Therefore separate qualification of the TRACG code is desirable, although not strictly necessary. TRACG is shown to be a valuable tool in determining the range of stable operating conditions for BWRs. The TRACG calculated decay ratio and measured decay ratio are reported for five tests. At the limit cycle condition the TRACG decay ratio matches within 15% of the measurement, with the code calculating a higher than actual decay ratio
Computer codes and methods for simulating accelerator driven systems
International Nuclear Information System (INIS)
A large set of computer codes and associated data libraries have been developed by nuclear research and industry over the past half century. A large number of them are in the public domain and can be obtained under agreed conditions from different Information Centres. The areas covered comprise: basic nuclear data and models, reactor spectra and cell calculations, static and dynamic reactor analysis, criticality, radiation shielding, dosimetry and material damage, fuel behaviour, safety and hazard analysis, heat conduction and fluid flow in reactor systems, spent fuel and waste management (handling, transportation, and storage), economics of fuel cycles, impact on the environment of nuclear activities etc. These codes and models have been developed mostly for critical systems used for research or power generation and other technological applications. Many of them have not been designed for accelerator driven systems (ADS), but with competent use, they can be used for studying such systems or can form the basis for adapting existing methods to the specific needs of ADS's. The present paper describes the types of methods, codes and associated data available and their role in the applications. It provides Web addresses for facilitating searches for such tools. Some indications are given on the effect of non appropriate or 'blind' use of existing tools to ADS. Reference is made to available experimental data that can be used for validating the methods use. Finally, some international activities linked to the different computational aspects are described briefly. (author)
Heat pipe design handbook, part 2. [digital computer code specifications
Skrabek, E. A.
1972-01-01
The utilization of a digital computer code for heat pipe analysis and design (HPAD) is described which calculates the steady state hydrodynamic heat transport capability of a heat pipe with a particular wick configuration, the working fluid being a function of wick cross-sectional area. Heat load, orientation, operating temperature, and heat pipe geometry are specified. Both one 'g' and zero 'g' environments are considered, and, at the user's option, the code will also perform a weight analysis and will calculate heat pipe temperature drops. The central porous slab, circumferential porous wick, arterial wick, annular wick, and axial rectangular grooves are the wick configurations which HPAD has the capability of analyzing. For Vol. 1, see N74-22569.
FLICA III M - reactors or test loops thermohydraulic computer code
International Nuclear Information System (INIS)
The FLICA III M code issued from the FLICA III code, of which it is the present stage of the development. This program calculates the flow and their transfer in steady and transient state in complex geometry described by subchannels. It is particularly used for the thermal-hydraulic analysis of reactors and experimental loops with heating rod bundles. A new solution method for the hydraulic problem is developed. It gives short computer times and allows the calculation of large subchannels sets. This makes possible the detailed calculations of hot subchannels jointly with these of rod bundles set in powered reactor cores. The equations solved take into account all the significant terms of the fundamental thermal-hydraulic equations and present models for turbulence and two-phase flows. The solution method couples together all the physical variables and makes possible the detailed description of complex flows
Compilation of the abstracts of nuclear computer codes available at CPD/IPEN
International Nuclear Information System (INIS)
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.)
CARP: a computer code and albedo data library for use by BREESE, the MORSE albedo package
International Nuclear Information System (INIS)
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
Computer code application programme of TAEA for thermal hydraulic research
International Nuclear Information System (INIS)
Evaluation of thermal-hydraulic conditions, fuel behavior, and reactor kinetic during various operating and postulated accident conditions results in conclusions that support decision-making process, the review of license application, and the resolution of other technical issues related to nuclear safety. Also these activities increase the understanding and involvement into new technical developments. Thermal-hydraulic research activities at TAEA focus on the application of computer codes that simulate the behavior of the reactor system. The computer codes are used to analyzed loss of coolant accidents, and system transients in light water nuclear reactors and to assess the consequences if imbalance occurs and to determine the effectiveness of mitigating actions. TAEA has used nuclear reactor system codes (RELAP5/Mod3.2 and higher versions, PARCS) and nuclear plant visual analyzer codes (NPA, SNAP, and XNGR5) obtained by in the framework of the CAMP Agreement signed between TAEA and the United States Nuclear Regulatory Commission (US NRC). TAEA performs and documents the code assessments including improvements and error corrections. Moreover, research activities concerning the passive cooling application and simulations of advanced nuclear power plant have been carried out by both experimental and theoretical means. For example, the experiment test facility, which was designed to investigate the effect of noncondensable gases on condensation, was conducted in cooperation with the Mechanical Engineering Department of the Middle East Technical University in Ankara (Turkey) and was finished. The text matrix obtained from this research was also submitted to US NRC data bank. Application of RELAP5 code f system transients include International Standard Problem (ISP) studies (ISP 33, 38, 42, 45, 46), accident analysis for different reactors and special topics in nuclear heat transfer problems (mid-loop operation). Research studies of severe accidents assess the detailed
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.
Multicode comparison of selected source-term computer codes
International Nuclear Information System (INIS)
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
Computer codes in nuclear safety, radiation transport and dosimetry
International Nuclear Information System (INIS)
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
pyro: A teaching code for computational astrophysical hydrodynamics
Zingale, Michael
2013-01-01
We describe pyro: a simple, freely-available code to aid students in learning the computational hydrodynamics methods widely used in astrophysics. pyro is written with simplicity and learning in mind and intended to allow students to experiment with various methods popular in the field, including those for advection, compressible and incompressible hydrodynamics, multigrid, and diffusion in a finite-volume framework. We show some of the test problems from pyro, describe its design philosophy, and suggest extensions for students to build their understanding of these methods.
Nyx: A MASSIVELY PARALLEL AMR CODE FOR COMPUTATIONAL COSMOLOGY
International Nuclear Information System (INIS)
We present a new N-body and gas dynamics code, called Nyx, for large-scale cosmological simulations. Nyx follows the temporal evolution of a system of discrete dark matter particles gravitationally coupled to an inviscid ideal fluid in an expanding universe. The gas is advanced in an Eulerian framework with block-structured adaptive mesh refinement; a particle-mesh scheme using the same grid hierarchy is used to solve for self-gravity and advance the particles. Computational results demonstrating the validation of Nyx on standard cosmological test problems, and the scaling behavior of Nyx to 50,000 cores, are presented.
Fault-tolerance for MPI Codes on Computational Clusters
Hagen, Knut Imar
2007-01-01
This thesis focuses on fault-tolerance for MPI codes on computational clusters. When an application runs on a very large cluster with thousands of processors, there is likely that a process crashes due to a hardware or software failure. Fault-tolerance is the ability of a system to respond gracefully to an unexpected hardware or software failure. A test application which is meant to run for several weeks on several nodes is used in this thesis. The application is a seismic MPI application, w...
Computer code for thermal hydraulic analysis of light water reactors
International Nuclear Information System (INIS)
A computer programme (THAL) has been developed to perform thermal hydraulic analysis of a single channel in a light water moderated core. In this code the hydrodynamic and thermodynamic equations describing one-dimensional axial flow have been discretized and solved explicitly stepwise up the coolant channel for an arbitrary power profile. THAL has been developed for use on small computers and it is capable of predicting the coolant, clad and fuel temperature profiles, steam quality, void fraction, pressure drop, critical heat flux and DNB ratio throughout the core. A boiling water reactor and a pressurized water reactor have been analyzed as test cases. The results obtained through the use of THAL compare favourably with those given in the design reports of these reactor systems. (author)
Computer code for shielding calculations of x-rays rooms
International Nuclear Information System (INIS)
The building an effective barrier against ionizing radiation present in radiographic rooms requires consideration of many variables. The methodology used for thickness specification of primary and secondary, barrier of a traditional radiographic room, considers the following factors: Use Factor, Occupational Factor, distance between the source and the wall, Workload, Kerma in the air and distance between the patient and the source. With these data it was possible to develop a computer code, which aims to identify and use variables in functions obtained through graphics regressions provided by NCRP-147 (Structural Shielding Design for Medical X-Ray Imaging Facilities) report, for shielding calculation of room walls, and the walls of the dark room and adjacent areas. With the implemented methodology, it was made a code validation by comparison of results with a study case provided by the report. The obtained values for thickness comprise different materials such as concrete, lead and glass. After validation it was made a case study of an arbitrary radiographic room.The development of the code resulted in a user-friendly tool for planning radiographic rooms to comply with the limits established by CNEN-NN-3:01 published in september/2011. (authors)
Comparison of computer code calculations with FEBA test data
International Nuclear Information System (INIS)
The FEBA forced feed reflood experiments included base line tests with unblocked geometry. The experiments consisted of separate effect tests on a full-length 5x5 rod bundle. Experimental cladding temperatures and heat transfer coefficients of FEBA test No. 216 are compared with the analytical data postcalculated utilizing the SSYST-3 computer code. The comparison indicates a satisfactory matching of the peak cladding temperatures, quench times and heat transfer coefficients for nearly all axial positions. This agreement was made possible by the use of an artificially adjusted value of the empirical code input parameter in the heat transfer for the dispersed flow regime. A limited comparison of test data and calculations using the RELAP4/MOD6 transient analysis code are also included. In this case the input data for the water entrainment fraction and the liquid weighting factor in the heat transfer for the dispersed flow regime were adjusted to match the experimental data. On the other hand, no fitting of the input parameters was made for the COBRA-TF calculations which are included in the data comparison. (orig.)
A computer code for seismic qualification of nuclear service valves
International Nuclear Information System (INIS)
The computer code, CERTIVALVE, has been developed for detailed frequency, seismic stress, and deformation analysis of nuclear service valves. It is an expedient means of analyzing, designing, and qualifying nuclear service valves for dynamic loads. The program is designed to be applicable to virtually all types of ASME Class 1, 2, and 3 valves, including gate, globe, disk, ball, safety-relief, diaphragm, and butterfly valves. It can evaluate valves of nearly all major manufacturers both foreign and domestic. CERTIVALVE computes all natural frequencies of any valve up to 40 Hertz (or higher if the user desires). The program constructs a multi-degree-offreedom (MDOF) finite element model which is used in the eigenvalue (natural frequency) solution. CERTIVALVE also computes the maximum allowable acceleration capacity of any valve on the basis of user-supplied allowable stress limits and the geometric and material properties of the valve components. The worst spatial orientation of the valve is assumed. The maximum allowable resultant is reported in terms of a multiple of weight (that is, in values of g) for each valve component. The valve components evaluated include the non-pressure retaining yoke structure for prismatic, nonprismatic, and curved sections; the yoke-bonnet junction for bolted, clamped and bossed connection details; the bonnet and body sections for various geometric cross sections; and the body-bonnet junction for both bolted-gasket and threaded connections. Finally, the program also performs an operability deformation analysis to check that stem binding is precluded. The evaluation procedures are performed in accordance with included codes and standards. The theoretical development of the program is provided and examples of the program options are presented
A computer code to simulate X-ray imaging techniques
International Nuclear Information System (INIS)
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
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.
International Nuclear Information System (INIS)
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.)
Interface design of VSOP'94 computer code for safety analysis
International Nuclear Information System (INIS)
Today, most software applications, also in the nuclear field, come with a graphical user interface. VSOP'94 (Very Superior Old Program), was designed to simplify the process of performing reactor simulation. VSOP is a integrated code system to simulate the life history of a nuclear reactor that is devoted in education and research. One advantage of VSOP program is its ability to calculate the neutron spectrum estimation, fuel cycle, 2-D diffusion, resonance integral, estimation of reactors fuel costs, and integrated thermal hydraulics. VSOP also can be used to comparative studies and simulation of reactor safety. However, existing VSOP is a conventional program, which was developed using Fortran 65 and have several problems in using it, for example, it is only operated on Dec Alpha mainframe platforms and provide text-based output, difficult to use, especially in data preparation and interpretation of results. We develop a GUI-VSOP, which is an interface program to facilitate the preparation of data, run the VSOP code and read the results in a more user friendly way and useable on the Personal 'Computer (PC). Modifications include the development of interfaces on preprocessing, processing and postprocessing. GUI-based interface for preprocessing aims to provide a convenience way in preparing data. Processing interface is intended to provide convenience in configuring input files and libraries and do compiling VSOP code. Postprocessing interface designed to visualized the VSOP output in table and graphic forms. GUI-VSOP expected to be useful to simplify and speed up the process and analysis of safety aspects
A computer code for analysis of severe accidents in LWRs
International Nuclear Information System (INIS)
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.)
Application of the RESRAD computer code to VAMP scenario S
International Nuclear Information System (INIS)
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
Computer code verification for power cycling behaviour of fuel pins
International Nuclear Information System (INIS)
A major goal of Indian our fuel development program is to develop the theoretical and analytical approach for the behaviour of power reactor fuels and wherever possible, to compare the results of the irradiation of fuel characterised rods with the theoretical predictions. Except for 2 BWRs, the rest of the thermal power reactors in India are PWRs. The fuel in the PWRs has behaved extremely well, but on the other hand, the failure rate of fuel in our BWRs is rather high. As the analytical approach to both types of fuel is different we commissioned 2 independent computer programs PECITIS-II and COMTA for predicting the behaviour of PHWR fuel and BWR fuel respectively. Because of relatively higher failure rate of BWR fuel, we are continuously putting a considerable effort in improving the code COMTA and are trying to irradiate representative and well characterised fuel rods in the pressurised water loop of the research reactor, CIRUS. Thermal and mechanical behaviour, of fuel pins with free standing clad, is analysed by the code COMTA, on an integral basis
Development Of The Computer Code For Comparative Neutron Activation Analysis
International Nuclear Information System (INIS)
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
Development of computer code for expansion stage in vapor explosion
International Nuclear Information System (INIS)
A computer code for numerical analysis of two-dimensional compressible flow during expansion stage in vapor explosion has been developed for safety assessment of severe accidents of light water reactors. The field equation is based on a homogeneous compressible model considering distributions of water, steam and core debris. The solution technique is finite volume method using non-staggered mesh scheme with the second order accuracy. In the initial condition, high temperature core debris spread in mixing area of water pool. Before vapor explosion occurred, core debris was covered with entrained air and heat transfer from the core debris to water was not so large. Contact between core debris and water took place somewhere in the mixing region, and pressure pulse appeared due to rapid evaporation. The pressure pulse propagated in the mixing region and contact between core debris and water began in the pressure pulse front. Transferred heat from core debris to water during τ after debris-water contact was assumed to be expressed as: Q(τ)=Qtotal{ 1-exp(-τ/τ0)} Qtotal is the total contribution from stored heat in core debris and τ0 is a time constant of evaporation. ALPHA experiments of melt-coolant interaction performed by JAERI were analyzed with the present computer code. The calculated pressure history in the mixing region was compared with the data. The predicted peak and width of pressure pulse were comparable with the data when the contributed fraction of stored heat was 0.5 and the evaporation time constant τ0 equaled to 2.5 ms. (author)
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.
HYDRA, A finite element computational fluid dynamics code: User manual
Energy Technology Data Exchange (ETDEWEB)
Christon, M.A.
1995-06-01
HYDRA is a finite element code which has been developed specifically to attack the class of transient, incompressible, viscous, computational fluid dynamics problems which are predominant in the world which surrounds us. The goal for HYDRA has been to achieve high performance across a spectrum of supercomputer architectures without sacrificing any of the aspects of the finite element method which make it so flexible and permit application to a broad class of problems. As supercomputer algorithms evolve, the continuing development of HYDRA will strive to achieve optimal mappings of the most advanced flow solution algorithms onto supercomputer architectures. HYDRA has drawn upon the many years of finite element expertise constituted by DYNA3D and NIKE3D Certain key architectural ideas from both DYNA3D and NIKE3D have been adopted and further improved to fit the advanced dynamic memory management and data structures implemented in HYDRA. The philosophy for HYDRA is to focus on mapping flow algorithms to computer architectures to try and achieve a high level of performance, rather than just performing a port.
Final technical position on documentation of computer codes for high-level waste management
International Nuclear Information System (INIS)
Guidance is given for the content of documentation of computer codes which are used in support of a license application for high-level waste disposal. The guidelines cover theoretical basis, programming, and instructions for use of the code
RADTRAN II: revised computer code to analyze transportation of radioactive material
International Nuclear Information System (INIS)
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
International Nuclear Information System (INIS)
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
An efficient methodology for modeling complex computer codes with Gaussian processes
Marrel, Amandine; Iooss, Bertrand; Van Dorpe, Francois; Volkova, Elena
2008-01-01
International audience Complex computer codes are often too time expensive to be directly used to perform uncertainty propagation studies, global sensitivity analysis or to solve optimization problems. A well known and widely used method to circumvent this inconvenience consists in replacing the complex computer code by a reduced model, called a metamodel, or a response surface that represents the computer code and requires acceptable calculation time. One particular class of metamodels is...
The TESS [Tandem Experiment Simulation Studies] computer code user's manual
International Nuclear Information System (INIS)
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
International Nuclear Information System (INIS)
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)
Assessment of the computer code COBRA/CFTL
International Nuclear Information System (INIS)
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
International Nuclear Information System (INIS)
This report is a user's manual for MLSOIL (Multiple Layer SOIL model) and DFSOIL (Dose Factors for MLSOIL) and a documentation of the computational methods used in those two computer codes. MLSOIL calculates an effective ground surface concentration to be used in computations of external doses. This effective ground surface concentration is equal to (the computed dose in air from the concentration in the soil layers)/(the dose factor for computing dose in air from a plane). MLSOIL implements a five compartment linear-transfer model to calculate the concentrations of radionuclides in the soil following deposition on the ground surface from the atmosphere. The model considers leaching through the soil as well as radioactive decay and buildup. The element-specific transfer coefficients used in this model are a function of the k/sub d/ and environmental parameters. DFSOIL calculates the dose in air per unit concentration at 1 m above the ground from each of the five soil layers used in MLSOIL and the dose per unit concentration from an infinite plane source. MLSOIL and DFSOIL have been written to be part of the Computerized Radiological Risk Investigation System (CRRIS) which is designed for assessments of the health effects of airborne releases of radionuclides. 31 references, 3 figures, 4 tables
IMIE computer codes: 10 years in the internal dosimetry
International Nuclear Information System (INIS)
Full text: The IMIE ('Individual Monitoring for Internal Exposure') computer codes are the family of interactive tools for interpretation of the bioassay data, individual dose assessments, tracking the history of exposure and documentation of the process of assessment. All members of the 'IMIE family' use the unified graphical interface, the common library of tabulated 'bioassay/dose response functions' and the Paradox database for storage of individual bioassay histories. During 10 years of the IMIE evolution the extensive experience in the IMIE application have been accumulated. Advances in the code development have been made thanks to IAEA and 3rd European Intercomparison Exercises, numerous users' requests and, especially, as a result of interaction in the framework of the IDEAS project aimed to the development of the International Guideline on Interpretation of Bioassay Data as well as to the organization of the 4th European Intercomparison Exercise. The main distinguished feature of the IMIE ideology is the automated numerical analysis of all requested by the user exposure scenarios with the succeeding interactive identification of events of intake(s) and assessment of associated doses. The visualization of data interpretation is based on the interaction with the user by means of interactive plots and tables. The IMIE exploits also the 'dose per unit measured bioassay value' (DPUM) concept, the useful tool for planning of bioassay programs as well as for interpretation of bioassay data. The numerical deconvolution algorithms and the massive library of tabulated 'bioassay/dose response functions' (time-dependent organ activity and excretion rates per unit deposited activity in one of lungs regions or per unit ingested/injected activity; associated committed equivalent doses to organs and the effective dose) are employed for processing of an arbitrary pattern of intake and complex exposure conditions. Nowadays distinguishing features of IMIE are: simultaneous
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.
T-Matrix: Codes for Computing Electromagnetic Scattering by Nonspherical and Aggregated Particles
Waterman, Peter; Mishchenko, Michael I.; Travis, Larry D.; Mackowski, Daniel W.
2015-11-01
The T-Matrix package includes codes to compute electromagnetic scattering by homogeneous, rotationally symmetric nonspherical particles in fixed and random orientations, randomly oriented two-sphere clusters with touching or separated components, and multi-sphere clusters in fixed and random orientations. All codes are written in Fortran-77. LAPACK-based, extended-precision, Gauss-elimination- and NAG-based, and superposition codes are available, as are double-precision superposition, parallelized double-precision, double-precision Lorenz-Mie codes, and codes for the computation of the coefficients for the generalized Chebyshev shape.
International Nuclear Information System (INIS)
In recent years, the use of computer codes to study the response of primary containment of large, liquid-metal fast breeder reactors (LMFBR) under postulated accident conditions has been adopted by most fast reactor projects. This paper gives a brief survey of the computational methods and codes available for LMFBR containment analysis. The various numerical methods commonly used in the computer codes are compared. It provides the reactor engineers up-to-date information on the development of structural dynamics in LMFBR containment analysis. It can also be used as a basis for the selection of the numerical method in the future code development. (Auth.)
Numerical computation of phase distribution in two fluid flow using the two-dimensional TOFFEA code
International Nuclear Information System (INIS)
A new iterative approach has been developed for multidimensional computational analysis of two fluid flow. It has been implemented and tested in a two-dimensional computer code. Parametric surveys are described to illustrate that this code rationally predicts separation of two fluid flows under gravitational and centrifugal influences. Comparisons are made between behaviour computed by the code, and results reported in experimental studies of air and water flowing in elbows and pipes. Plans for extending the code to three dimensions are discussed, as are methods for incorporating an improved model of turbulence
3-D field computation: The near-triumph of commerical codes
Energy Technology Data Exchange (ETDEWEB)
Turner, L.R.
1995-07-01
In recent years, more and more of those who design and analyze magnets and other devices are using commercial codes rather than developing their own. This paper considers the commercial codes and the features available with them. Other recent trends with 3-D field computation include parallel computation and visualization methods such as virtual reality systems.
How in-house computer codes help utilities get the best out of their fuel
International Nuclear Information System (INIS)
Utilities are increasingly realizing the importance of having a full in-house capability in the field of in-core fuel management. This requires competent staff as well as accurate, reliable and easy-to-use computer codes. The development of advanced computer codes is thus an important task for the utilities and the supporting industry. (author)
Modification in the CITATION computer code: change of microscopic cross section by zone
International Nuclear Information System (INIS)
The modifications done in the Citation computer code in order to compute the accumulated burnup after each burnup step for each reactor zone and to allow the use of update microscopic cross sections for each zone according to the accumulated burnup are presented. Some input data was introduced in the code. The modifications reported here were checked and some comparisons were made with results obtained by running the code with and without these modifications. (Author)
TRANS4: a computer code calculation of solid fuel penetration of a concrete barrier
International Nuclear Information System (INIS)
The computer code, TRANS4, models the melting and penetration of a solid barrier by a solid disc of fuel following a core disruptive accident. This computer code has been used to model fuel debris penetration of basalt, limestone concrete, basaltic concrete, and magnetite concrete. Sensitivity studies were performed to assess the importance of various properties on the rate of penetration. Comparisons were made with results from the GROWS II code
Compendium of computer codes for the safety analysis of fast breeder reactors
International Nuclear Information System (INIS)
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
Regulatory requirements to the thermal-hydraulic and thermal-mechanical computer codes
International Nuclear Information System (INIS)
The paper presents an overview of the regulatory requirements to the thermal-hydraulic and thermal-mechanical computer codes, which are used for safety assessment of the fuel design and the fuel utilization. Some requirements to the model development, verification and validation of the codes and analysis of code uncertainties are also define. Questions concerning Quality Assurance during development and implementation of the codes as well as preparation of a detailed verification and validation plan are briefly discussed
Cooperation of experts' opinion, experiment and computer code development
International Nuclear Information System (INIS)
The connection between code development, code assessment and confidence in the analysis of transients will be discussed. In this manner, the major sources of errors in the codes and errors in applications of the codes will be shown. Standard problem results emphasize that, in order to have confidence in licensing statements, the codes must be physically realistic and the code user must be qualified and experienced. We will discuss why there is disagreement between the licensing authority and vendor concerning assessment of the fullfillment of safety goal requirements. The answer to the question lies in the different confidence levels of the assessment of transient analysis. It is expected that a decrease in the disagreement will result from an increased confidence level. Strong efforts will be made to increase this confidence level through improvements in the codes, experiments and related organizational strcutures. Because of the low probability for loss-of-coolant-accidents in the nuclear industry, assessment must rely on analytical techniques and experimental investigations. (orig./HP)
Fuel behavior modeling using the MARS computer code
International Nuclear Information System (INIS)
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)
Computer code ANISN multiplying media and shielding calculation 2. Code description (input/output)
International Nuclear Information System (INIS)
The new code CCC-0514-ANISN/PC is described, as well as a ''GENERAL DESCRIPTION OF ANISN/PC code''. In addition to the ANISN/PC code, the transmittal package includes an interactive input generation programme called APE (ANISN Processor and Evaluator), which facilitates the work of the user in giving input. Also, a 21 group photon cross section master library FLUNGP.LIB in ISOTX format, which can be edited by an executable file LMOD.EXE, is included in the package. The input and output subroutines are reviewed. 6 refs, 1 fig., 1 tab
Research on parallel computing of MCNP code based on MPI
International Nuclear Information System (INIS)
This paper introduces the method that develops a parallel computing platform with ordinary PCs and the mpi.nt.1.2.5 software based on MPI (Message Interface Passing) standard specification on Windows operating system. The parallel computing of MCNP on this platform is realized, and the parallel computing performance of MCNP is analyzed in this paper. (authors)
Computer codes for simulating atomic-displacement cascades in solids subject to irradiation
International Nuclear Information System (INIS)
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)
International Nuclear Information System (INIS)
The CITHAN computer code was developed at IPEN (Instituto de Pesquisas Energeticas e Nucleares) to link the HAMMER computer code with a fuel depletion routine and to provide neutron cross sections to be read with the appropriate format of the CITATION code. The problem arised due to the efforts to addapt the new version denomined HAMMER-TECHION with the routine refered. The HAMMER-TECHION computer code was elaborated by Haifa Institute, Israel within a project with EPRI. This version is at CNEN to be used in multigroup constant generation for neutron diffusion calculation in the scope of the new methodology to be adopted by CNEN. The theoretical formulation of CITHAM computer code, tests and modificatins are described. (Author)
Qualification of the new version of HAMMER computer code
International Nuclear Information System (INIS)
(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 238U captures; D25 - ratio of epicadmium-to-subcadmium 235U fission; D28 - ratio of 238U fissions to 235U fissions; C - ratio of 238U captures to 235U fissions; RC02 - ratio of epicadmium-to-subcadmium 232Th 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)
Licensing applications and topical reviews of the RETRAN computer code
International Nuclear Information System (INIS)
On September 4, 1984, the US Nuclear Regulatory Commission (NRC) approved, contingent upon completion of correction of errors discovered in the review process, use of RETRAN-02/MOD002 for certain utility applications. Because this generic code review had not included review of plant specific applications, the approval was of the code only, and applicants are required to justify use of the code and its models on a plant and transient specific basis. During and following completion of the generic review, a number of utilities have submitted analyses requesting that the NRC approve the submittals for purposes ranging from simply demonstrating that the applicant was able to use the code, to detailed transient and accident analysis in the licensing arena. The NRC guidance for applicants consists primarily of Generic Letter 83-11 in which an applicant is broadly instructed to perform its own code verification and demonstrate its own technical competence. This broad guidance has led to a wide range of technical detail in submittals and therefore to a correspondingly broad spectrum of NRC approvals. This paper discusses the status of those applications, reviews the breadth of requests and, finally discusses issues that have arisen with respect to these applications
International Nuclear Information System (INIS)
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
Fuel management methodology upgrade of Thai Research Reactor (TRR-1/M1) using SRAC computer code
International Nuclear Information System (INIS)
This paper presents the effort to upgrade the fuel management methodology of Thai Research Reactor-1/Modification 1 (TRR-1/M1) which is currently under responsibility of Thailand Institute of Nuclear Technology (TINT). The more advanced SRAC computer code is being introduced to replace the TRIGAP computer code for the fuel management calculation of TRR-1/M1. With the new methodology, the hexagonal lattices of TRR-1/M1 can be modeled without approximating the lattices into cylindrical rings as performed by the TRIGAP computer code. In addition, the SRAC computer code is able to provide pin-wise results such as normalized power distribution which is unable to obtain by the TRIGAP computer code. Also, the paper compares the core excess reactivity of core loading 1 and core loading 2 calculated by SRAC computer code with the measurement data from the operation log book. The comparison shows good agreement between the calculated values and measured values. With the promising result, the SRAC computer code is expected to be employed as the usual fuel management methodology for TRR-1/M1 in the near future. (author)
Development of a graphical interface computer code for reactor fuel reloading optimization
International Nuclear Information System (INIS)
This report represents the results of the project performed in 2007. The aim of this project is to develop a graphical interface computer code that allows refueling engineers to design fuel reloading patterns for research reactor using simulated graphical model of reactor core. Besides, this code can perform refueling optimization calculations based on genetic algorithms as well as simulated annealing. The computer code was verified based on a sample problem, which relies on operational and experimental data of Dalat research reactor. This code can play a significant role in in-core fuel management practice at nuclear research reactor centers and in training. (author)
International Nuclear Information System (INIS)
Code verification and validation are the key elements during any computer program development. Canadian regulatory framework provides requirements and guidelines for conducting code validation activities that are consistent with other jurisdictions. It also ensures that verification and validation process in support of the design and licensing of CANDU power reactors is compliant with regulatory expectations. The paper presents a new Canadian regulatory perspective for key software QA activities: code verification and validation. Also, the regulatory approved practices for planning, conducting and documentation produced to support verification and validation of computer codes are also discussed. (author)
Study and application of Dot 3.5 computer code in radiation shielding problems
International Nuclear Information System (INIS)
The application of nuclear transportation code S sub(N), Dot 3.5, to radiation shielding problems is revised. Aiming to study the better available option (convergence scheme, calculation mode), of DOT 3.5 computer code to be applied in radiation shielding problems, a standard model from 'Argonne Code Center' was selected and a combination of several calculation options to evaluate the accuracy of the results and the computational time was used, for then to select the more efficient option. To illustrate the versatility and efficacy in the application of the code for tipical shielding problems, the streaming neutrons calculation along a sodium coolant channel is ilustrated. (E.G.)
International Nuclear Information System (INIS)
A 70-group, 37-isotope library of multigroup constants for fast reactor nuclear design calculations is described. Nuclear cross sections, transfer matrices, and self-shielding factors were generated with NJOY code and an auxiliary program RGENDF using evaluated data from ENDF/B-IV. The output is being issued in a format suitable for EXPANDA code. Comparisons with JFS-2 library, as well as, test resuls for 14 CSEWG benchmark critical assemblies are presented. (Author)
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.
Efficient Quantification of Uncertainties in Complex Computer Code Results Project
National Aeronautics and Space Administration — Propagation of parameter uncertainties through large computer models can be very resource intensive. Frameworks and tools for uncertainty quantification are...
Hartenstein, Richard G., Jr.
1985-08-01
Computer codes have been developed to analyze antennas on aircraft and in the presence of scatterers. The purpose of this study is to use these codes to develop accurate computer models of various aircraft and antenna systems. The antenna systems analyzed are a P-3B L-Band antenna, an A-7E UHF relay pod antenna, and traffic advisory antenna system installed on a Bell Long Ranger helicopter. Computer results are compared to measured ones with good agreement. These codes can be used in the design stage of an antenna system to determine the optimum antenna location and save valuable time and costly flight hours.
Calculations of reactor-accident consequences, Version 2. CRAC2: computer code user's guide
International Nuclear Information System (INIS)
The CRAC2 computer code is a revision of the Calculation of Reactor Accident Consequences computer code, CRAC, developed for the Reactor Safety Study. The CRAC2 computer code incorporates significant modeling improvements in the areas of weather sequence sampling and emergency response, and refinements to the plume rise, atmospheric dispersion, and wet deposition models. New output capabilities have also been added. This guide is to facilitate the informed and intelligent use of CRAC2. It includes descriptions of the input data, the output results, the file structures, control information, and five sample problems
Validation of thermohydraulic codes by comparison of experimental results with computer simulations
International Nuclear Information System (INIS)
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.)
Development of the computer code system for the analyses of PWR core
International Nuclear Information System (INIS)
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)
Code for prompt numerical computation of the leading order GPD evolution
Vinnikov, A. V.
2006-01-01
This paper describes the design and work of a set of computer routines capable for numerical computation of generalized parton distributions (GPDs) evolution at the leading order. The main intention of this work is to present a fast-working computer code making possible fitting of GPDs parameters to the data on hard electron-nucleon scattering.
International Nuclear Information System (INIS)
Components of reactor systems and related equipment are identified in which multidimensional computational thermal hydraulics can be used to advantage to assess and improve design. Models of single- and two-phase flow are reviewed, and the governing equations for multidimensional analysis are discussed. Suitable computational algorithms are introduced, and sample results from the application of particular multidimensional computer codes are given
Application of two-dimensional burnup computer codes to the operation of nuclear power plants
International Nuclear Information System (INIS)
The needs for three-dimensional computer calculations of the power density distribution in WWER type reactors are outlined. In most cases, however, two-dimensional calculations provide sufficiently exact results and result in a decrease in computer costs. The application, performance and computer codes of two-dimensional calculations are dealt with. (author)
Simulations of PWR spray systems by ASTEC computer code
International Nuclear Information System (INIS)
In this paper, sequence of loss of feedwater of steam generators on a PWR 900 MWe is performed by application of integral code ASTEC. The influence of the spray on evolution and source term of severe accident in the containment and in the environment is mainly studied. The results are helpful for the investigation of mitigative measures of severe accident. (authors)
Use of NESTLE computer code for NPP transition process analysis
International Nuclear Information System (INIS)
A newly created WWER-440 reactor model with use NESTLE code is discussed. Results of 'fast' and 'slow' transition processes based on it are presented. This model was developed for Rovno NPP reactor and it can be used also for WWER-1000 reactor in Zaporozhe NPP
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.
Selection of a computer code for Hanford low-level waste engineered-system performance assessment
Energy Technology Data Exchange (ETDEWEB)
McGrail, B.P.; Mahoney, L.A.
1995-10-01
Planned performance assessments for the proposed disposal of low-level waste (LLW) glass produced from remediation of wastes stored in underground tanks at Hanford, Washington will require calculations of radionuclide release rates from the subsurface disposal facility. These calculations will be done with the aid of computer codes. Currently available computer codes were ranked in terms of the feature sets implemented in the code that match a set of physical, chemical, numerical, and functional capabilities needed to assess release rates from the engineered system. The needed capabilities were identified from an analysis of the important physical and chemical process expected to affect LLW glass corrosion and the mobility of radionuclides. The highest ranked computer code was found to be the ARES-CT code developed at PNL for the US Department of Energy for evaluation of and land disposal sites.
International Nuclear Information System (INIS)
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
International Nuclear Information System (INIS)
In the years 1990, the knowledge gained in the area of interest and advances in computer technology provided the basis for developing the new generation computer codes within Russia. The development of computer codes intended for system modeling of NPP reactor dynamics was motivated by the need in reconstruction and life extension of ageing NPPs; construction of Russian-design NPPs abroad; development of the new generation NPPs with passive safety features. Solution to these problems involves, a depth numerical analysis of transients, design-basis and beyond design-basis accidents at NPPs as an integral part of nuclear safety research required by both domestic and international standards. NITI researchers started developing the KORSAR code in 1996. It was planned to develop three base versions of the code. Code developers faced three problems presented and discussed in this paper. (author)
Selection of a computer code for Hanford low-level waste engineered-system performance assessment
International Nuclear Information System (INIS)
Planned performance assessments for the proposed disposal of low-level waste (LLW) glass produced from remediation of wastes stored in underground tanks at Hanford, Washington will require calculations of radionuclide release rates from the subsurface disposal facility. These calculations will be done with the aid of computer codes. Currently available computer codes were ranked in terms of the feature sets implemented in the code that match a set of physical, chemical, numerical, and functional capabilities needed to assess release rates from the engineered system. The needed capabilities were identified from an analysis of the important physical and chemical process expected to affect LLW glass corrosion and the mobility of radionuclides. The highest ranked computer code was found to be the ARES-CT code developed at PNL for the US Department of Energy for evaluation of and land disposal sites
LIMBO computer code for analyzing coolant-voiding dynamics in LMFBR safety tests
International Nuclear Information System (INIS)
The LIMBO (liquid metal boiling) code for the analysis of two-phase flow phenomena in an LMFBR reactor coolant channel is presented. The code uses a nonequilibrium, annular, two-phase flow model, which allows for slip between the phases. Furthermore, the model is intended to be valid for both quasi-steady boiling and rapid coolant voiding of the channel. The code was developed primarily for the prediction of, and the posttest analysis of, coolant-voiding behavior in the SLSF P-series in-pile safety test experiments. The program was conceived to be simple, efficient, and easy to use. It is particularly suited for parametric studies requiring many computer runs and for the evaluation of the effects of model or correlation changes that require modification of the computer program. The LIMBO code, of course, lacks the sophistication and model detail of the reactor safety codes, such as SAS, and is therefore intended to compliment these safety codes
A new 3-D integral code for computation of accelerator magnets
International Nuclear Information System (INIS)
For computing accelerator magnets, integral codes have several advantages over finite element codes; far-field boundaries are treated automatically, and computed fields in the bore region satisfy Maxwell's equations exactly. A new integral code employing the 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
International Nuclear Information System (INIS)
In this paper two new numerical methods for computing two phase flow transients are presented. One methods uses a semi-implicit technique and the other one applies a fully implicit technique. A computer module: TRITON, has been developped using these two methods. The physical modeling of this code is the two fluid flow model with six partial differential equations, the various transfer terms between wall and fluid or between liquid and steam are given by correlations implemented into the POSEIDON Nuclear Safety Analysis Computer code system. (orig.)
International Nuclear Information System (INIS)
In this paper two new numerical methods for computing two phase flow transients are presented. One method uses a semi-implicit technique and the other one applies a fully implicit technique. A computer module Triton has been developped using these two methods. The physical modeling of this code is the two fluid flow model with six partial differential equations, the various transfer terms between wall and fluid or between liquid and steam are given by correlations implemented into the Poseidon Nuclear Safety Analysis Computer code system
Computer codes used during upgrading activities at MINT TRIGA reactor
Energy Technology Data Exchange (ETDEWEB)
Mohammad Suhaimi Kassim; Adnan Bokhari; Mohd. Idris Taib [Malaysian Institute for Nuclear Technology Research, Kajang (Malaysia)
1999-10-01
MINT TRIGA Reactor is a 1-MW swimming pool nuclear research reactor commissioned in 1982. In 1993, a project was initiated to upgrade the thermal power to 2 MW. The IAEA assistance was sought to assist the various activities relevant to an upgrading exercise. For neutronics calculations, the IAEA has provided expert assistance to introduce the WIMS code, TRIGAP, and EXTERMINATOR2. For thermal-hydraulics calculations, PARET and RELAP5 were introduced. Shielding codes include ANISN and MERCURE. However, in the middle of 1997, MINT has decided to change the scope of the project to safety upgrading of the MINT Reactor. This paper describes some of the activities carried out during the upgrading process. (author)
DCHAIN 2: a computer code for calculation of transmutation of nuclides
International Nuclear Information System (INIS)
DCHAIN2 is a one-point depletion code which solves the coupled equation of radioactive growth and decay for a large number of nuclides by the Bateman method. A library of nuclear data for 1170 fission products has been prepared for providing input data to this code. The Bateman method surpasses the matrix exponential method in computational accuracies and in saving computer storage for the code. However, most existing computer codes based on the Bateman method have shown serious drawbacks in treating cyclic chains and more than a few specific types of decay chains. The present code has surmounted the above drawbacks by improving the code FP-S, and has the following characteristics: (1) The code can treat any type of transmutation through decays or neutron induced reactions. Multiple decays and reactions are allowed for a nuclide. (2) Unknown decay energy in the nuclear data library can be estimated. (3) The code constructs the decay scheme of each nuclide in the code and breaks it up into linear chains. Nuclide names, decay types and branching ratios of mother nuclides are necessary as the input data for each nuclide. Order of nuclides in the library is arbitrary because each nuclide is destinguished by its nuclide name. (4) The code can treat cyclic chains by an approximation. A library of the nuclear data has been prepared for 1170 fission products, including the data for half-lives, decay schemes, neutron absorption cross sections, fission yields, and disintegration energies. While DCHAIN2 is used to compute the compositions, radioactivity and decay heat of fission products, the gamma-ray spectrum of fission products can be computed also by a separate code FPGAM using the composition obtained from DCHAIN2. (J.P.N.)
GATE: computation code for medical imagery, radiotherapy and dosimetry
International Nuclear Information System (INIS)
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
A key equation and the computation of error values for codes from order domains
Little, John B.
2003-01-01
We study the computation of error values in the decoding of codes constructed from order domains. Our approach is based on a sort of analog of the key equation for decoding Reed-Solomon and BCH codes. We identify a key equation for all codes from order domains which have finitely-generated value semigroups; the field of fractions of the order domain may have arbitrary transcendence degree, however. We provide a natural interpretation of the construction using the theory of Macaulay's inverse ...
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 user's guide to GENEX, SDR, and related computer codes
International Nuclear Information System (INIS)
This series of codes will be of use in a variety of fields connected with reactor physics, examples of which are: (a) In evaluation of nuclear data in which the RESP-GENEX part of the system would be used to examine and produce a cross-section set based on the theories and experiments of the nuclear physicists. The approximations in GENEX must however be kept in mind, the chief one being the diagonal expansion approximation of the inverse level matrix originally due to Bethe which precludes a correct representation of strong interference effects (the Lynn effect). (b) In the calculation of Doppler effects or other resonance effects such as establishing equivalence relationships, approximate resonance treatments, etc. A given set of tapes generated by GENEX (or by some other means into the GENEX format) would be used to run the SDH code. The SDR code produces cross-sections and reaction rates over any group structure within its working range. In situations with complex geometries the spatial representation of SDR is liable to be inadequate and in these circumstances it is recommended that the reaction rates are not used directly but instead the cross-sections are used in a more accurate spatial calculation to produce revised reaction rates. (c) Finally the system may be used for a variety of special investigations such as an analysis of the variance of the Doppler coefficient in fast reactors or the accurate assessment of ideal integral measurements, (for instance the Aldermaston sphere experiment
International Nuclear Information System (INIS)
This report presents the NRC staff with a tool for assessing the potential effects of accidental releases of radioactive materials and toxic substances on habitability of nuclear facility control rooms. The tool is a computer code that estimates concentrations at nuclear facility control room air intakes given information about the release and the environmental conditions. The name of the computer code is EXTRAN. EXTRAN combines procedures for estimating the amount of airborne material, a Gaussian puff dispersion model, and the most recent algorithms for estimating diffusion coefficients in building wakes. It is a modular computer code, written in FORTRAN-77, that runs on personal computers. It uses a math coprocessor, if present, but does not require one. Code output may be directed to a printer or disk files. 25 refs., 8 figs., 4 tabs
Structural dynamics in LMFBR containment analysis: a brief survey of computational methods and codes
Energy Technology Data Exchange (ETDEWEB)
Chang, Y.W.; Gvildys, J.
1977-01-01
In recent years, the use of computer codes to study the response of primary containment of large, liquid-metal fast breeder reactors (LMFBR) under postulated accident conditions has been adopted by most fast reactor projects. Since the first introduction of REXCO-H containment code in 1969, a number of containment codes have evolved and been reported in the literature. The paper briefly summarizes the various numerical methods commonly used in containment analysis in computer programs. They are compared on the basis of truncation errors resulting in the numerical approximation, the method of integration, the resolution of the computed results, and the ease of programming in computer codes. The aim of the paper is to provide enough information to an analyst so that he can suitably define his choice of method, and hence his choice of programs.
Simulation of CANON experience with RELAP4/MOD5 computer code
International Nuclear Information System (INIS)
The results obtained by RELAP4/MOD5 computer code for simulating the CANON experience, which studies the depressurization phase of a loss of primary coolant accident in PWR type reactors, are presented. (M.C.K.)
A Computer Code For Evaluation of Design Parameters of Concrete Piercing Earth Shock Missile Warhead
Roy, P. K.; K. Ramarao
1985-01-01
A simple and reliable computer code has been devised for evaluating various design parameters, and predicting the penetration performance of concrete piercing earth shock missile-warhead and will be useful to the designers of earth penetrating weapon system.
Computer code for nuclear reactor core thermal reliability calculation
International Nuclear Information System (INIS)
RASTENAR program was described for computing heat-engineering reliability of cores in nuclear reactors operating under stationary conditions. The following factors of heat-engineering reliability were found to be computable: rated critical margin; limiting critical margin; probability of initiation of critical heat removal in channel (inferior conditions of heat transfer); probability that no channel would be subject to critical heat removal; and reactor power reserve coefficient. The probability that no channel in the core would experience critical heat removal when boiling during operation of the reactor at fixed power level was taken for the principal quantitative criterion. The structure and limitations of the program were described together with the computation algorithm. The program was written for an M-220 computer
LWR-WIMS, a computer code for light water reactor lattice calculations
International Nuclear Information System (INIS)
LMR-WIMS is a comprehensive scheme of computation for studying the reactor physics aspects and burnup behaviour of typical lattices of light water reactors. This report describes the physics methods that have been incorporated in the code, and the modifications that have been made since the code was issued in 1972. (U.K.)
Simulation of the blowdown experiencies using the TRAC-PD2 computer code
International Nuclear Information System (INIS)
The experiments CANON and EDWARD'S PIPE were intented to simulate the blowdown phase of a typical PWR loss-of-coolant accident by depressurizing horizontal tubes filled with water at different pressures and temperatures. In this work the computer code TRAC-PD2 was employed to model those experiments. The code results are in good agreement with the experimental data. (Author)
PAD: a one-dimensional, coupled neutronic-thermodynamic-hydrodynamic computer code
International Nuclear Information System (INIS)
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
Study and application of Morse computer code in radiation shielding problems
International Nuclear Information System (INIS)
The uses of Morse computer code by the group of radiation transport and Shielding of IPEN (Instituto de Pesquisas Energeticas e Nucleares) are shown. Three sample problems publicated in the literature were solved, aiming to compare the results obtained. A small introduction on the logical construction and potencialites of the code is presented. (E.G.)
Performance analysis of VVER-type fuel rods with the STOFFEL-1 computer code
International Nuclear Information System (INIS)
The main features of the fuel rod performance modelling computer code STOFFEL-1 are described. Submodels of the code are briefly characterized, and some results of comparisons between model predictions and experiments are presented. Examples of modelling calculations are given for some thermo-mechanical values of VVER-1000 fuel rods. (author)
HADES. A computer code for fast neutron cross section from the Optical Model
International Nuclear Information System (INIS)
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
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…
Improvement of the computing speed of the FBR fuel pin bundle deformation analysis code 'BAMBOO'
International Nuclear Information System (INIS)
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)
Distribution of absorbed dose in human eye simulated by SRNA-2KG computer code
International Nuclear Information System (INIS)
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)
A FORTRAN computer code for calculating flows in multiple-blade-element cascades
Mcfarland, E. R.
1985-01-01
A solution technique has been developed for solving the multiple-blade-element, surface-of-revolution, blade-to-blade flow problem in turbomachinery. The calculation solves approximate flow equations which include the effects of compressibility, radius change, blade-row rotation, and variable stream sheet thickness. An integral equation solution (i.e., panel method) is used to solve the equations. A description of the computer code and computer code input is given in this report.
Off-site dose calculation computer code based on ICRP-60(II) - liquid radioactive effluents -
International Nuclear Information System (INIS)
The development of computer code for calculating off-site doses(K-DOSE60) was based on ICRP-60 and the dose calculationi equations of Reg. Guide 1.109. In this paper, the methodology to compute dose for liquid effluents was described. To examine reliability of the K-DOSE60 code the results obtained from K-DOSE60 were compared with analytic solutions. For liquid effluents. The results by K-DOSE60 are in agreement with analytic solution
A restructuring of the CF/EDF packages for the MIDAS computer code
International Nuclear Information System (INIS)
The CF and EDF packages, which allow the user to define the functions of variables in a database and the usage of an external data file, 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 restructure the code, the data transferring methods of the current MELCOR code are modified and then partially adopted into the CF/EDF packages. The data structure of the current MELCOR code using FORTRAN77 has a difficulty in grasping the meaning of the variables as pointers are used to define their addresses. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type without pointers leading to an efficient memory treatment and an easy understanding of the code. Restructuring of the CF/EDF packages addressed in this paper includes a module development and subroutine modification. The verification has been done by comparing the results of the modified code with those of the existing code and the trends are almost the same to each other. Therefore the similar approach could be extended to the entire code package for code restructuring. It is expected that the code restructuring will accelerate the code's domestication thanks to a direct understanding of each variable and an easy implementation of the modified or newly developed models. (author)
A restructuring of the CF/EDF packages for the MIDAS computer code
Energy Technology Data Exchange (ETDEWEB)
Park, S.H.; Kim, K.R.; Kim, D.H. [Korea Atomic Energy Research Inst., Yuseong, Daejon (Korea, Republic of)]. E-mail: shpark2@kaeri.re.kr; krkim@kaeri.re.kr; dhkim8@kaeri.re.kr
2004-07-01
The CF and EDF packages, which allow the user to define the functions of variables in a database and the usage of an external data file, 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 restructure the code, the data transferring methods of the current MELCOR code are modified and then partially adopted into the CF/EDF packages. The data structure of the current MELCOR code using FORTRAN77 has a difficulty in grasping the meaning of the variables as pointers are used to define their addresses. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type without pointers leading to an efficient memory treatment and an easy understanding of the code. Restructuring of the CF/EDF packages addressed in this paper includes a module development and subroutine modification. The verification has been done by comparing the results of the modified code with those of the existing code and the trends are almost the same to each other. Therefore the similar approach could be extended to the entire code package for code restructuring. It is expected that the code restructuring will accelerate the code's domestication thanks to a direct understanding of each variable and an easy implementation of the modified or newly developed models. (author)
Development of a one-dimensional computer code describing the convection in arbitrary flow networks
International Nuclear Information System (INIS)
The subject of this paper is in the first place the computer code LOOPY and the application of this code to the mathematical investigation of the flow behavior of the HHT demonstration plant at afterheat removal operation and on failure of all afterheat removal circulators. Moreover, comparative calculations assuming failure of the afterheat removal were performed for the THTR in order to verify the code. The present code may be used to solve the problems mentioned above. In combination with the two-dimensional flow and heat code THERMIX and other code units a modular code system may be established by means of which the flow and temperature behavior in HTR plants may be described with justifiable computing effort. Within such a code system the one-dimensional code takes over the duty of describing the loops adjacent to the core. The results obtained using the one-dimensional code must, however, be judged critically with respect to natural convection where in part only small mass flows will be formed. (orig.)
Plutonium explosive dispersal modeling using the MACCS2 computer code
International Nuclear Information System (INIS)
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
The computer code SEURBNUK/EURDYN. Pt. 2
International Nuclear Information System (INIS)
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
Further development of the computer codes COCOSYS and ASTEC
International Nuclear Information System (INIS)
In this project the long-term code development of GRS's simulation tools COCOSYS and ASTEC sponsored by the BMWi, for phenomena within the containment of light water reactors (LWR) has been continued. In correspondence to the top-level objective several important models in COCOSYS and ASTEC for phenomena which are currently under investigation have been updated and improved according to the progress of international research and development (R and D). A modification of the COCOSYS zone model has been elaborated and implemented in COCOSYS; this allows now the simulation of a complete flooding of containment rooms by introducing a new type of generic flow connection between two compartments. Through this flow connection a combined flow of water gas in parallel is possible. This work will be continued in a possible follow-up development project in order to consider in more detail the actual conditions of technical flow connections like doors, flaps etc. In co-operation with ITWM Kaiserslautern the coupling of COCOSYS with the pool model CoPool, which allows the detailed calculation of 3D convection patterns and temperature distribution in a deep water pool and which is under development at ITWM, has been realized. Furthermore, a model for metal fibre filter systems has been developed for COCOSYS for detailed investigations on the efficiency of filtered containment venting systems. This model has been successfully tested on the basis of two experiments from the ACE-A-series in which appropriate filter types had been used. Apart from this, the models for several other source term related phenomena have been improved: Consideration of the impact of steam concentration on the decomposition of ozone; the latter contributes to the formation of iodine aerosols; modeling of the radiolytic formation of nitric acid, which takes effect on the pH of the sump and thus on the iodine chemistry, within the atmosphere of the containment; consideration of the impact of humidity on
Status Report on Hydrogen Management and Related Computer Codes
International Nuclear Information System (INIS)
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
Verification of the network flow and transport/distributed velocity (NWFT/DVM) computer code
International Nuclear Information System (INIS)
The Network Flow and Transport/Distributed Velocity Method (NWFT/DVM) computer code was developed primarily to fulfill a need for a computationally efficient ground-water flow and contaminant transport capability for use in risk analyses where, quite frequently, large numbers of calculations are required. It is a semi-analytic, quasi-two-dimensional network code that simulates ground-water flow and the transport of dissolved species (radionuclides) in a saturated porous medium. The development of this code was carried out under a program funded by the US Nuclear Regulatory Commission (NRC) to develop a methodology for assessing the risk from disposal of radioactive wastes in deep geologic formations (FIN: A-1192 and A-1266). In support to the methodology development program, the NRC has funded a separate Maintenance of Computer Programs Project (FIN: A-1166) to ensure that the codes developed under A-1192 or A-1266 remain consistent with current operating systems, are as error-free as possible, and have up-to-date documentations for reference by the NRC staff. Part of this effort would include verification and validation tests to assure that a code correctly performs the operations specified and/or is representing the processes or system for which it is intended. This document contains four verification problems for the NWFT/DVM computer code. Two of these problems are analytical verifications of NWFT/DVM where results are compared to analytical solutions. The other two are code-to-code verifications where results from NWFT/DVM are compared to those of another computer code. In all cases NWFT/DVM showed good agreement with both the analytical solutions and the results from the other code
International Nuclear Information System (INIS)
The KAPROS module TRANSX provides self-shielded group constants from SIGMN-data-blocks on external data files or in KAPROS data blocks for the computer codes TWODANT, DIAMANT2, TRITAC and DEGEN/DEGRAT, which cannot handle SIGMN-blocks themselves. This report describes the procedures for the transfer of the cross sections. Some sample problems illustrate the use of TRANSX for each of the corresponding applications. (orig.)
Sharing of computer codes and data for accelerator shield modelling
International Nuclear Information System (INIS)
The Radiation Shielding Information Center (RSIC) and the NEA Data Bank (DB) acquire, verify and distribute computer programs and data sets which are needed by the communities working in nuclear research and applications. Programs and Data are shared through cooperative arrangements at the international level in order to avoid uneconomical duplication of efforts. These activities respond to needs emerging from national programmes and expressed by the users. This paper addresses explicitly the field of accelerator shield modelling and the available cross section data and computer programs required for the purpose. It suggests that international cooperation between the centres and participants in this field should be strengthened. Relevant computer programs are being benchmarked against experiments and the Centers are promoting and activity for collecting them in a computerized data base for easy access. (authors). 1 ref
Development and qualification of AECL computer codes for ACR safety analysis applications
International Nuclear Information System (INIS)
Over the past decade, AECL has developed and rigorously implemented a software Quality Assurance Program to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. This paper provides an overview of the computer programs used in Advanced CANDU Reactor (ACR) safety analysis, including their purpose and linkages, and assessments of their applicability in the safety analyses of the ACR design. The paper also reviews the key elements of the Software Quality Assurance program applied in development and validation of these computer programs for ACR application including the computer code change control process and the documentation produced to support the qualification of the computer codes. (author)
TPASS: a gamma-ray spectrum analysis and isotope identification computer code
International Nuclear Information System (INIS)
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
Development of a system of computer codes for severe accident analyses and its applications
International Nuclear Information System (INIS)
The objectives of this study is to develop a system of computer codes for postulated severe accident analyses in Nuclear Power Plants. This system of codes is necessary to conduct individual plant examination for domestic nuclear power plants. As a result of this study, one can conduct severe accident assessments more easily, and can extract the plant-specific vulnerabilities for severe accidents and at the same time the ideas for enhancing overall accident resistance. The scope and contents of this study are as follows : development of a system of computer codes for severe accident analyses, development of severe accident management strategy
Development of a system of computer codes for severe accident analyses and its applications
Energy Technology Data Exchange (ETDEWEB)
Chang, Soon Hong; Cheon, Moon Heon; Cho, Nam jin; No, Hui Cheon; Chang, Hyeon Seop; Moon, Sang Kee; Park, Seok Jeong; Chung, Jee Hwan [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)
1991-12-15
The objectives of this study is to develop a system of computer codes for postulated severe accident analyses in Nuclear Power Plants. This system of codes is necessary to conduct individual plant examination for domestic nuclear power plants. As a result of this study, one can conduct severe accident assessments more easily, and can extract the plant-specific vulnerabilities for severe accidents and at the same time the ideas for enhancing overall accident resistance. The scope and contents of this study are as follows : development of a system of computer codes for severe accident analyses, development of severe accident management strategy.
Validation of computer codes used in the safety analysis of Canadian research reactors
International Nuclear Information System (INIS)
AECL has embarked on a validation program for the suite of computer codes that it uses in performing the safety analyses for its research reactors. Current focus is on codes used for the analysis of the two MAPLE reactors under construction at Chalk River but the program will be extended to include additional codes that will be used for the Irradiation Research Facility. The program structure is similar to that used for the validation of codes used in the safety analyses for CANDU power reactors. (author)
FRAPCON-1: a computer code for the steady state analysis of oxide fuel rods
Energy Technology Data Exchange (ETDEWEB)
Berna, G. A.; Bohn, M. P.; Coleman, D. R.; Lanning, D. D.
1978-08-01
FRAPCON is a FORTRAN IV computer code which predicts the steady state long-term burnup response of a light water reactor fuel rod. The coupled effects of fuel and cladding deformation, temperature, and internal gas pressure on the behavior of the fuel rod are considered in determining fuel rod response. The cladding deformation model includes multi-axial, elasto-plastic analysis and considers both primary and secondary creep. The fuel temperature model considers the effects of fuel cracking and relocation in determining the fuel temperature distribution. Burnup dependent fission gas generation and release is included in calculating fuel rod internal pressure. An integral fuel rod failure subcode determines failure and failure modes based on the operating conditions at each timestep. The material property subcode, MATPRO, provides gas, fuel and cladding properties to the computational subcodes in FRAPCON. No material properties need to be supplied by the code user. FRAPCON is a completely modular code with each major computational subcode isolated within the code and coupled to the main code by subroutine calls and data transfer through argument lists. FRAPCON is soft-coupled to the transient fuel rod code, FRAP-T, to provide initial conditions to initiate analysis of such off-normal transients as a loss-of-coolant accident. The code is presently programmed and running on a CDC 7600 computer.
International Nuclear Information System (INIS)
A fast running computer code SHETEMP has been developed for analysis of reactivity initiated accidents under constant core cooling conditions such as coolant temperature and heat transfer coefficient on fuel rods. This code can predict core power and fuel temperature behaviours. A control rod movement can be taken into account in power control system. The objective of the code is to provide fast running capability with easy handling of the code required for audit and design calculations where a large number of calculations are performed for parameter surveys during short time period. The fast running capability of the code was realized by neglection of fluid flow calculation. The computer code SHETEMP was made up by extracting and conglomerating routines for reactor kinetics and heat conduction in the transient reactor thermal-hydraulic analysis code ALARM-P1, and by combining newly developed routines for reactor power control system. As ALARM-P1, SHETEMP solves point reactor kinetics equations by the modified Runge-Kutta method and one-dimensional transient heat conduction equations for slab and cylindrical geometries by the Crank-Nicholson methods. The model for reactor power control system takes into account effects of PID regulator and control rod drive mechanism. In order to check errors in programming of the code, calculated results by SHETEMP were compared with analytic solution. Based on the comparisons, the appropriateness of the programming was verified. Also, through a sample calculation for typical modelling, it was concluded that the code could satisfy the fast running capability required for audit and design calculations. This report will be described as a code manual of SHETEMP. It contains descriptions on a sample problem, code structure, input data specifications and usage of the code, in addition to analytical models and results of code verification calculations. (author)
Proceedings of the conference on computer codes and the linear accelerator community
International Nuclear Information System (INIS)
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
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.
Condensation model for the FLUENT computer code. Rev. 0
International Nuclear Information System (INIS)
Developed within the project 'Use of a fluidoelastic 3D code for analysis of the function of VVER-440/V-213 containments with a bubble condenser and a spray system as pressure limiters', the report is structured as follows: (1) Condensation model for FLUENT 6 - use of the Species model (Physical properties of the mixture); (2) Specific condensation model (Cells of walls, User defined function (UDF) for condensation in FLUENT); (3) Test problem for condensation (Particular sample of the test problem; Results of the test problem); (4) Notes on multi-phase models in FLUENT. Proposals for refining (UDF for condensation in Volume of Fluid/mixture multi-phase models); (5) Testing and modifying the network for the EREC facility with the new FLUENT 6.1 version and a new version of the condensation model. (P.A.)
Evaluation of the use of six diagnostic X-ray spectra computer codes.
Meyer, P; Buffard, E; Mertz, L; Kennel, C; Constantinesco, A; Siffert, P
2004-03-01
A knowledge of photon energy spectra emitted from X-ray tubes in radiology is crucial for many research domains in the medical field. Since spectrometry is difficult because of high photon fluence rates, a convenient solution is to use computational models. This paper describes the use of six computer codes based on semiempirical or empirical models. The use of the codes was assessed, notably by comparing theoretical half value layers and air kerma with measurements on five different X-ray tubes used in a research hospital. It was found that three out of the six computer codes give relative spectra very close to those produced by X-ray units equipped with constant potential generators: the mean difference between measured and modelled half value layer was less than 3% with a standard deviation of 3.6% whatever the tube and the applied voltage. Absolute output is less accurate: for four computer codes, the mean difference between the measured and modelled air kerma was between 18% and 36%, with a standard deviation of 9% whatever the tube (except for the single phase generator) and the applied voltage. One of the codes gives a good output and beam quality for X-ray units equipped with 100% ripple voltage generators. The use of computational codes as described in this paper provides a means of modelling relative diagnostic X-ray spectra, the usefulness of the tube output data depending on the accuracy required by the end user. PMID:15020364
Computer code for general analysis of radon risks (GARR)
International Nuclear Information System (INIS)
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
Spent fuel management fee methodology and computer code user's manual
International Nuclear Information System (INIS)
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
Compute-and-Forward: Harnessing Interference through Structured Codes
Nazer, Bobak; Gastpar, Michael C.
2009-01-01
Abstract—Interference is usually viewed as an obstacle to communication in wireless networks. This paper proposes a new strategy, compute-and-forward, that exploits interference to obtain significantly higher rates between users in a network. The key idea is that relays should decode linear functions of transmitted messages according to their observed channel coefficients rather than ignoring the interference as noise. After decoding these linear equations, the relays simply send them t...
Characterization of the MCNPX computer code in micro processed architectures
International Nuclear Information System (INIS)
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)
A finite element option for the MARC transport/ diffusion theory computer code
International Nuclear Information System (INIS)
The MARC multigroup transport/diffusion theory computer code has been extended to include a finite element option. The facility is available in two-dimensional geometry and has a novel feature in allowing high order polynomial approximations to the flux using an automated computer procedure. (U.K.)
H2OPROP - computer code for determination of light water properties
International Nuclear Information System (INIS)
An accurate determination of light water properties plays a significant role in various thermal hydraulic analyses. Determination of thermophysical properties of light water has evolved over a number of years. Reactor Safety Division has participated in IAEA coordinated research project and arrived at formulations for light water based on IAPS formulations. However, safety analysis codes like RELAP IV/MOD6 and COHRA needs a few more formulations and over extended range and hence additional verifications. In view of this a computer code H2OPROP has been developed. Formulations developed in this code are general, so that any desired thermodynamic property can be evaluated in future, which is not presented in this report. This code has been compared with various reference data for all properties over the entire range. It has also been used to generate properties that are required as inputs to different safety analysis codes. This report describes various formulation used in the code, its uses and its inputs. (author)
Benchmark testing and independent verification of the VS2DT computer code
International Nuclear Information System (INIS)
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
The WECHSL Code: A computer program for the interaction of a core melt with concrete
International Nuclear Information System (INIS)
The WECHSL Code is a mechanistic computer code developed for the analysis of the interaction of molten LWR reactor materials with concrete. The code in its present state performs calculations from the time of the initial contact of a hot molten pool slumping into a concrete cavity to the time of freezing of an entire layer of the pool. The modelling of the freezing phase, however, needs some further elaboration along with experimental confirmation. The code is capable to treat both the limited type of simulation experiments with melt masses between 100 and 600 kg as well as hypothetical core melt down accidents with actual, full scale reactor dimensions. The submodels used in the code and their physical background are described in detail. Furthermore, the results of a test sample along with instructions for the use of the code are given. (orig.)
D20PROP- computer code for determination of heavy water properties
International Nuclear Information System (INIS)
An accurate determination of heavy water properties plays a significant role in various thermal hydraulic analyses. Determination of thermophysical properties of heavy water has evolved over a number of years. Reactor Safety Division has participated in IAEA coordinated research project and arrived at formulations for heavy water based on IAPS formulations. However, safety analysis codes like RELAP IV/MOD6 and COHRA needs few more formulations and over extended range and hence additional verifications. In view of this a computer code D2OPROP has been developed. Formulations developed in this code are general, so that any desired thermodynamic property can be evaluated in future, which is not presented in this report. This code has been compared with various reference data for all properties over the entire range. It has also been used to generate properties that are required as inputs to different safety analysis codes. This report describes various formulation used in the code, its uses and its inputs. (author)
Development of MCNPX-ESUT computer code for simulation of neutron/gamma pulse height distribution
Energy Technology Data Exchange (ETDEWEB)
Abolfazl Hosseini, Seyed, E-mail: sahosseini@sharif.edu [Department of Energy Engineering, Sharif University of Technology, Tehran 8639-11365 (Iran, Islamic Republic of); Vosoughi, Naser [Department of Energy Engineering, Sharif University of Technology, Tehran 8639-11365 (Iran, Islamic Republic of); Zangian, Mehdi [Nuclear Engineering Department, Shahid Beheshti University, G.C., PO Box 1983963113, Tehran (Iran, Islamic Republic of)
2015-05-11
In this paper, the development of the MCNPX-ESUT (MCNPX-Energy Engineering of Sharif University of Technology) computer code for simulation of neutron/gamma pulse height distribution is reported. Since liquid organic scintillators like NE-213 are well suited and routinely used for spectrometry in mixed neutron/gamma fields, this type of detectors is selected for simulation in the present study. The proposed algorithm for simulation includes four main steps. The first step is the modeling of the neutron/gamma particle transport and their interactions with the materials in the environment and detector volume. In the second step, the number of scintillation photons due to charged particles such as electrons, alphas, protons and carbon nuclei in the scintillator material is calculated. In the third step, the transport of scintillation photons in the scintillator and lightguide is simulated. Finally, the resolution corresponding to the experiment is considered in the last step of the simulation. Unlike the similar computer codes like SCINFUL, NRESP7 and PHRESP, the developed computer code is applicable to both neutron and gamma sources. Hence, the discrimination of neutron and gamma in the mixed fields may be performed using the MCNPX-ESUT computer code. The main feature of MCNPX-ESUT computer code is that the neutron/gamma pulse height simulation may be performed without needing any sort of post processing. In the present study, the pulse height distributions due to a monoenergetic neutron/gamma source in NE-213 detector using MCNPX-ESUT computer code is simulated. The simulated neutron pulse height distributions are validated through comparing with experimental data (Gohil et al. Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 664 (2012) 304–309.) and the results obtained from similar computer codes like SCINFUL, NRESP7 and Geant4. The simulated gamma pulse height distribution for a {sup
The extensive international use of commercial computational fluid dynamics (CFD) codes
International Nuclear Information System (INIS)
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)
High pressure humidification columns: Design equations, algorithm, and computer code
Energy Technology Data Exchange (ETDEWEB)
Enick, R.M. [Pittsburgh Univ., PA (United States). Dept. of Chemical and Petroleum Engineering; Klara, S.M. [USDOE Pittsburgh Energy Technology Center, PA (United States); Marano, J.J. [Burns and Roe Services Corp., Pittsburgh, PA (United States)
1994-07-01
This report describes the detailed development of a computer model to simulate the humidification of an air stream in contact with a water stream in a countercurrent, packed tower, humidification column. The computer model has been developed as a user model for the Advanced System for Process Engineering (ASPEN) simulator. This was done to utilize the powerful ASPEN flash algorithms as well as to provide ease of use when using ASPEN to model systems containing humidification columns. The model can easily be modified for stand-alone use by incorporating any standard algorithm for performing flash calculations. The model was primarily developed to analyze Humid Air Turbine (HAT) power cycles; however, it can be used for any application that involves a humidifier or saturator. The solution is based on a multiple stage model of a packed column which incorporates mass and energy, balances, mass transfer and heat transfer rate expressions, the Lewis relation and a thermodynamic equilibrium model for the air-water system. The inlet air properties, inlet water properties and a measure of the mass transfer and heat transfer which occur in the column are the only required input parameters to the model. Several example problems are provided to illustrate the algorithm`s ability to generate the temperature of the water, flow rate of the water, temperature of the air, flow rate of the air and humidity of the air as a function of height in the column. The algorithm can be used to model any high-pressure air humidification column operating at pressures up to 50 atm. This discussion includes descriptions of various humidification processes, detailed derivations of the relevant expressions, and methods of incorporating these equations into a computer model for a humidification column.
Development of a computer code for elasto-plastic axisymmetric thin shells
International Nuclear Information System (INIS)
A computer code (AXSHEL2) utilizing the finite element method is developed to analyze elastoplastic thin shells of revolution under axisymmetric loading. The solution is based on a doubly curved axisymmetric thin shell element formulated using Novozhilov's theory. The element is well suited for all ranges of latitude angles of rotational shells. The element accommodates rigid body translation which improves element convergence behaviour. The tangent stiffness method in conjunction with incremental flow theory of plasticity is used in the solution technique. The computer program is set up to analyze thin shells of elastic - perfectly plastic and isotropic hardening materials using Von Mises yeild criteria. In addition, the program can analyze elastic or elasto-plastic buckling problems. Several numerical examples are given to illustrate accuracy, convergence and suitability of the computer code for practical applications. The computer code provides an efficient 'in-house' capability for solving linear and non-linear structural problems associated with nuclear power plant equipment
Verification of thermal-hydraulic computer codes against standard problems for WWER reflooding
International Nuclear Information System (INIS)
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 modification and application of RAMS computer code. Final report
International Nuclear Information System (INIS)
The Regional Atmospheric Modeling System (RAMS) has been utilized in its most updated form, version 3a, to simulate a case night from the Atmospheric Studies in COmplex Terrain (ASCOT) experimental program. ASCOT held a wintertime observational campaign during February, 1991 to observe the often strong drainage flows which form on the Great Plains and in the canyons embedded within the slope from the Continental Divide to the Great Plains. A high resolution (500 m grid spacing) simulation of the 4-5 February 1991 case night using the more advanced turbulence closure now available in RAMS 3a allowed greater analysis of the physical processes governing the drainage flows. It is found that shear interaction above and within the drainage flow are important, and are overpredicted with the new scheme at small grid spacing (< ∼1000 m). The implication is that contaminants trapped in nighttime stable flows such as these, will be mixed too strongly in the vertical reducing predicted ground concentrations. The HYPACT code has been added to the capability at LANL, although due to the reduced scope of work, no simulations with HYPACT were performed
The Uncertainty Test for the MAAP Computer Code
Energy Technology Data Exchange (ETDEWEB)
Park, S. H.; Song, Y. M.; Park, S. Y.; Ahn, K. I.; Kim, K. R.; Lee, Y. J. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2008-10-15
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.
SAVER: digital computer code to determine pressure drops
International Nuclear Information System (INIS)
The SAVER code calculates the steady-state flow and pressure drop in all paths of an arbitrary flow network. SAVER calculates elevation and momentum recoverable pressure drops and the types of unrecoverable pressure drops encountered most often. The pressure drop in each path is the sum of the elevation, momentum, and unrecoverable pressure drops. The momentum and unrecoverable pressure drops are expressed as R x W2, where R is the resistance to flow and W is the flow rate. The elevation pressure drop is not a function of flow rate. The elevation pressure drop and the resistances to flow are calculated using the B and W standard methods. A set of flow and pressure drop equations is written using these values. First, N-1 flow balance equations are written, where N is the number of flow path junctions in the network. Then, all the unique pressure drop balance equations are written, considering that the pressure drop across each path through the network must equal the pressure drop imposed by the boundary conditions. These equations are then solved for the flow rate in each path. (auth)
A restructuring of the MELCOR fission product packages for the MIDAS computer code
International Nuclear Information System (INIS)
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)
Development of a model and computer code to describe solar grade silicon production processes
Gould, R. K.; Srivastava, R.
1979-01-01
Two computer codes were developed for describing flow reactors in which high purity, solar grade silicon is produced via reduction of gaseous silicon halides. 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. It can be used to described 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, deposition 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.
Validation of physics and thermalhydraulic computer codes for advanced Candu reactor applications
International Nuclear Information System (INIS)
Atomic Energy of Canada Ltd. (AECL) is developing an Advanced Candu Reactor (ACR) that is an evolutionary advancement of the currently operating Candu 6 reactors. The ACR is being designed to produce electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The ACR retains the modular Candu concept of horizontal fuel channels surrounded by a heavy water moderator. However, ACR uses slightly enriched uranium fuel compared to the natural uranium used in Candu 6. This achieves the twin goals of improved economics (via large reductions in the heavy water moderator volume and replacement of the heavy water coolant with light water coolant) and improved safety. AECL has developed and implemented a software quality assurance program to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. Since the basic design of the ACR is equivalent to that of the Candu 6, most of the key phenomena associated with the safety analyses of ACR are common, and the Candu industry standard tool-set of safety analysis codes can be applied to the analysis of the ACR. A systematic assessment of computer code applicability addressing the unique features of the ACR design was performed covering the important aspects of the computer code structure, models, constitutive correlations, and validation database. Arising from this assessment, limited additional requirements for code modifications and extensions to the validation databases have been identified. This paper provides an outline of the AECL software quality assurance program process for the validation of computer codes used to perform physics and thermal-hydraulics safety analyses of the ACR. It describes the additional validation work that has been identified for these codes and the planned, and ongoing, experimental programs to extend the code validation as required to address specific ACR design
Algorithms and computer codes for atomic and molecular quantum scattering theory
International Nuclear Information System (INIS)
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
Institute of Scientific and Technical Information of China (English)
无
2010-01-01
The eukaryotic genome contains varying numbers of non-coding RNA(ncRNA) genes."Computational RNomics" takes a multidisciplinary approach,like information science,to resolve the structure and function of ncRNAs.Here,we review the main issues in "Computational RNomics" of data storage and management,ncRNA gene identification and characterization,ncRNA target identification and functional prediction,and we summarize the main methods and current content of "computational RNomics".
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
ISP-46 analysis with RELAP5/SCDAPSIM computer code
International Nuclear Information System (INIS)
The thermal-hydraulic and severe accidents analysis code RELAP5/SCDAPSIM was used in the calculation of the Phebus FPT1 in-pile experiment. This experiment, carried out on 26 July 1996 in the Phebus facility, Cadarache, France, was chosen as the basis for the OECD International Standard Problem (ISP-46) exercise. Investigation of severe accidents phenomena like fuel degradation and hydrogen production was the objective of the ISP and of the presented analysis. The ISP was an open exercise, that is, all the relevant experimental results were available to the participants from the start. The FPT1 test bundle included 18 PWR fuel rods previously irradiated to a mean burnup of 23.4 GWd/tU, two instrumented fresh fuel rods and one silver-indium-cadmium control rod. The bundle was housed in an insulating shroud and introduced into the Phebus driver core which supplied the nuclear power. The fuel degradation phase of the test lasted about 5 hours during which the bundle was cooled by steam at pressure of about 2 bar with the mass flow rate varying between 0.5 g/s and 2.2 g/s, while the bundle nuclear power was being progressively increased from zero up to 36.5 kW. RELAP5/SCDAPSIM modelling of the Phebus facility and the main results, such as the temperature response of all rods and shroud, the oxidation and resulting hydrogen production, will be discussed and presented in this paper. The analysis of fuel rods degradation and problems related to SCDAPSIM underprediction of the amount of relocated fuel and cladding will also be covered. (author)
Development of computing code system for level 3 PSA
Energy Technology Data Exchange (ETDEWEB)
Jeong, Jong Tae; Yu, Dong Han; Kim, Seung Hwan
1997-07-01
Among the various research areas of the level 3 PSA, the effect of terrain on the transport of radioactive material was investigated through wind tunnel experiment. These results will give a physical insight in the development of a new dispersion model. Because there are some discrepancies between the results from Gaussian plume model and those from field test, the effect of terrain on the atmospheric dispersion was investigated by using CTDMPLUS code. Through this study we find that the model which can treat terrain effect is essential in the atmospheric dispersion of radioactive materials and the CTDMPLUS model can be used as a useful tool. And it is suggested that modification of a model and experimental study should be made through the continuous effort. The health effect assessment near the Yonggwang site by using IPE (Individual plant examination) results and its site data was performed. The health effect assessment is an important part of consequence analysis of a nuclear power plant site. The MACCS was used in the assessment. Based on the calculation of CCDF for each risk measure, it is shown that CCDF has a slow slope and thus wide probability distribution in cases of early fatality, early injury, total early fatality risk, and total weighted early fatality risk. And in cases of cancer fatality and population dose within 48km and 80km, the CCDF curve have a steep slope and thus narrow probability distribution. The establishment of methodologies for necessary models for consequence analysis resulting form a server accident in the nuclear power plant was made and a program for consequence analysis was developed. The models include atmospheric transport and diffusion, calculation of exposure doses for various pathways, and assessment of health effects and associated risks. Finally, the economic impact resulting form an accident in a nuclear power plant was investigated. In this study, estimation models for each cost terms that considered in economic
Multiplexing Genetic and Nucleosome Positioning Codes: A Computational Approach.
Eslami-Mossallam, Behrouz; Schram, Raoul D; Tompitak, Marco; van Noort, John; Schiessel, Helmut
2016-01-01
Eukaryotic DNA is strongly bent inside fundamental packaging units: the nucleosomes. It is known that their positions are strongly influenced by the mechanical properties of the underlying DNA sequence. Here we discuss the possibility that these mechanical properties and the concomitant nucleosome positions are not just a side product of the given DNA sequence, e.g. that of the genes, but that a mechanical evolution of DNA molecules might have taken place. We first demonstrate the possibility of multiplexing classical and mechanical genetic information using a computational nucleosome model. In a second step we give evidence for genome-wide multiplexing in Saccharomyces cerevisiae and Schizosacharomyces pombe. This suggests that the exact positions of nucleosomes play crucial roles in chromatin function. PMID:27272176
Universal holonomic quantum computing with cat-codes
Albert, Victor V.; Shu, Chi; Krastanov, Stefan; Shen, Chao; Liu, Ren-Bao; Yang, Zhen-Biao; Schoelkopf, Robert J.; Mirrahimi, Mazyar; Devoret, Michel H.; Jiang, Liang
2016-05-01
Universal computation of a quantum system consisting of superpositions of well-separated coherent states of multiple harmonic oscillators can be achieved by three families of adiabatic holonomic gates. The first gate consists of moving a coherent state around a closed path in phase space, resulting in a relative Berry phase between that state and the other states. The second gate consists of ``colliding'' two coherent states of the same oscillator, resulting in coherent population transfer between them. The third gate is an effective controlled-phase gate on coherent states of two different oscillators. Such gates should be realizable via reservoir engineering of systems which support tunable nonlinearities, such as trapped ions and circuit QED.
Computer-aided software understanding systems to enhance confidence of scientific codes
International Nuclear Information System (INIS)
A unique characteristic of nuclear waste disposal is the very long time span over which the combined engineered and natural containment system must remain effective: hundreds of thousands of years. Since there is no precedent in human history for such an endeavour, simulation with the use of computers is the only means we have of forecasting possible future outcomes quantitatively. The need for reliable models and software to make such forecasts so far into the future is obvious. One of the critical elements necessary to ensure reliability is the degree of reviewability of the computer program. Among others, there are two very important reasons for this. Firstly, if there is to be any chance at all of validating the conceptual models as implemented by the computer code, peer reviewers must be able to see and understand what the program is doing. It is all but impossible to achieve this understanding by just looking at the code due to possible unfamiliarity with the language and often due as well to the length and complexity of the code. Secondly, a thorough understanding of the code is also necessary to carry out code maintenance activities which include among others, error detection, error correction and code modification for purposes of enhancing its performance, functionality or to adapt it to a changed environment. The emerging concepts of computer-aided software understanding and reverse engineering can answer precisely these needs. This paper will discuss the role they can play in enhancing the confidence one has on computer codes and several examples will be provided. Finally a brief discussion of combining state-of-art forward engineering systems with reverse engineering systems will show how powerfully they can contribute to the overall quality assurance of a computer program. (13 refs., 7 figs.)
Skála, J.; Baruffa, F.; Büchner, J.; Rampp, M.
2015-08-01
Context. The numerical simulation of turbulence and flows in almost ideal astrophysical plasmas with large Reynolds numbers motivates the implementation of magnetohydrodynamical (MHD) computer codes with low resistivity. They need to be computationally efficient and scale well with large numbers of CPU cores, allow obtaining a high grid resolution over large simulation domains, and be easily and modularly extensible, for instance, to new initial and boundary conditions. Aims: Our aims are the implementation, optimization, and verification of a computationally efficient, highly scalable, and easily extensible low-dissipative MHD simulation code for the numerical investigation of the dynamics of astrophysical plasmas with large Reynolds numbers in three dimensions (3D). Methods: The new GOEMHD3 code discretizes the ideal part of the MHD equations using a fast and efficient leap-frog scheme that is second-order accurate in space and time and whose initial and boundary conditions can easily be modified. For the investigation of diffusive and dissipative processes the corresponding terms are discretized by a DuFort-Frankel scheme. To always fulfill the Courant-Friedrichs-Lewy stability criterion, the time step of the code is adapted dynamically. Numerically induced local oscillations are suppressed by explicit, externally controlled diffusion terms. Non-equidistant grids are implemented, which enhance the spatial resolution, where needed. GOEMHD3 is parallelized based on the hybrid MPI-OpenMP programing paradigm, adopting a standard two-dimensional domain-decomposition approach. Results: The ideal part of the equation solver is verified by performing numerical tests of the evolution of the well-understood Kelvin-Helmholtz instability and of Orszag-Tang vortices. The accuracy of solving the (resistive) induction equation is tested by simulating the decay of a cylindrical current column. Furthermore, we show that the computational performance of the code scales very
KC-A Kinectic computer code for investigation of parametric plasma instabilities
International Nuclear Information System (INIS)
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)
International Nuclear Information System (INIS)
The objective of this paper is to present a compilation of computer codes for the assessment of accidental or routine releases of radioactivity to the environment from nuclear power facilities. The capabilities of 83 computer codes in the areas of environmental transport and radiation dosimetry are summarized in tabular form. This preliminary analysis clearly indicates that the initial efforts in assessment methodology development have concentrated on atmospheric dispersion, external dosimetry, and internal dosimetry via inhalation. The incorporation of terrestrial and aquatic food chain pathways has been a more recent development and reflects the current requirements of environmental legislation and the needs of regulatory agencies. The characteristics of the conceptual models employed by these codes are reviewed. The appendixes include abstracts of the codes and indexes by author, key words, publication description, and title
Vectorization of nuclear codes on FACOM 230-75 APU computer
International Nuclear Information System (INIS)
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)
Energy Technology Data Exchange (ETDEWEB)
Hoffman, F. O.; Miller, C. W.; Shaeffer, D. L.; Garten, Jr., C. T.; Shor, R. W.; Ensminger, J. T.
1977-04-01
The objective of this paper is to present a compilation of computer codes for the assessment of accidental or routine releases of radioactivity to the environment from nuclear power facilities. The capabilities of 83 computer codes in the areas of environmental transport and radiation dosimetry are summarized in tabular form. This preliminary analysis clearly indicates that the initial efforts in assessment methodology development have concentrated on atmospheric dispersion, external dosimetry, and internal dosimetry via inhalation. The incorporation of terrestrial and aquatic food chain pathways has been a more recent development and reflects the current requirements of environmental legislation and the needs of regulatory agencies. The characteristics of the conceptual models employed by these codes are reviewed. The appendixes include abstracts of the codes and indexes by author, key words, publication description, and title.
Theory of the space-dependent fuel management computer code ''UAFCC''
International Nuclear Information System (INIS)
This report displays the theory of the spatial burnup computer code ''UAFCC'' which has been constructed as a part of an integrated reactor calculation scheme proposed at the Reactors Department of the ARE Atomic Energy Authority. The ''UAFCC'' is a single energy-one-dimensional diffusion burnup FORTRAN computer code for well moderated, multiregion, cylindrical thermal reactors. The effect of reactivity variation with burnup is introduced in the steady state diffusion equation by a fictitious neutron source. The infinite multiplication factor, the total migration area, and the power density per unit thermal flux are calculated from the point model burnup code ''UABUC'' fitted to polynomials of suitable degree in the flux-time, and then used as an input data to the ''UAFCC'' code. The proposed burnup spatial model has been used to study the different stratogemes of the incore fuel management schemes. The conclusions of this study will be presented in a future publication. (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.
Compendium of computer codes for the researcher in magnetic fusion energy
International Nuclear Information System (INIS)
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
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.
A Compact Code for Simulations of Quantum Error Correction in Classical Computers
International Nuclear Information System (INIS)
This study considers implementations of error correction in a simulation language on a classical computer. Error correction will be necessarily in quantum computing and quantum information. We will give some examples of the implementations of some error correction codes. These implementations will be made in a more general quantum simulation language on a classical computer in the language Mathematica. The intention of this research is to develop a programming language that is able to make simulations of all quantum algorithms and error corrections in the same framework. The program code implemented on a classical computer will provide a connection between the mathematical formulation of quantum mechanics and computational methods. This gives us a clear uncomplicated language for the implementations of algorithms.
HIFI: a computer code for projectile fragmentation accompanied by incomplete fusion
International Nuclear Information System (INIS)
A brief summary of a model proposed to describe projectile fragmentation accompanied by incomplete fusion and the instructions for the use of the computer code HIFI are given. The code HIFI calculates single inclusive spectra, coincident spectra and excitation functions resulting from particle-induced reactions. It is a multipurpose program which can calculate any type of coincident spectra as long as the reaction is assumed to take place in two steps
SAMDIST: A computer code for calculating statistical distributions for R-matrix resonance parameters
Energy Technology Data Exchange (ETDEWEB)
Leal, L.C.; Larson, N.M.
1995-09-01
The SAMDIST computer code has been developed to calculate distribution of resonance parameters of the Reich-Moore R-matrix type. The program assumes the parameters are in the format compatible with that of the multilevel R-matrix code SAMMY. SAMDIST calculates the energy-level spacing distribution, the resonance width distribution, and the long-range correlation of the energy levels. Results of these calculations are presented in both graphic and tabular forms.
COSMOCR: A Numerical Code for Cosmic Ray Studies in Computational Cosmology
Miniati, Francesco
2001-01-01
We present COSMOCR, a numerical code for the investigation of cosmic ray related studies in computational cosmology. The code follows the diffusive shock acceleration, the mechanical and radiative energy losses and the spatial transport of the supra-thermal particles in cosmic environment. Primary cosmic ray electrons and ions are injected at shocks according to the thermal leakage prescription. Secondary electrons are continuously injected as a results of p-p inelastic collisions of primary ...
Definition of RBMK-1500 reactor spent nuclear fuel characteristics using different computer codes
International Nuclear Information System (INIS)
Concentrations of actinides were calculated in dependence of the burn-up by means of computer codes APOLLO1 and SAS2H (SCALE), applying five concentric zones model for RBMK-1500 fuel assembly. Using SAS2H code the faster burn out of 235U and slower generation of 239Pu and 241Pu isotopes was revealed. It is possible due to different initial data such as cross sections and fission products yields libraries. (author)
Suitability of the present computer codes for the future nuclear power plants
International Nuclear Information System (INIS)
The report presents evaluations of calculation results and modelling abilities of three thermal hydraulic computer codes for analysis of gravity driven emergency core cooling (ECC) systems in advanced reactor concepts. The work is based on PACTEL (PArallel Channel TEst Loop) experiments modelled with RELAP5/mod3.1, APROS(V2.11) and CATHARE 2 V1.3 code. (8 refs.)
Description of computer code PRINS, Program for Interpreting Gamma Spectra, developed at ENEA
International Nuclear Information System (INIS)
The computer code PRINS, PRogram for INterpreting gamma Spectra, has been developed in collaboration with CENG/SECC (Centre Etude Nucleaire Grenoble / Service Etude Comportement du Combustible). Later it has been updated and improved at ENEA. Properties of the PRINS code are: I) A powerful algorithm to locate the peaks; 2) An accurate evaluation of the errors; 3) Possibility of an automatic channels-energy calibration
POPCYCLE: a computer code for calculating nuclear and fossil plant levelized life-cycle power costs
International Nuclear Information System (INIS)
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)
Usage is described of the computer code DYSAC (Dynamic System Analysis Code) developed for a hybrid computer for the identification and the analysis of system dynamics. A multivariable linear dynamic system is identified based on the autoregressive model using the time series data obtained from a system in operation and the system dynamics thus identified are analyzed. This code includes subroutines for the analysis of step response, frequency response, power spectrum, etc. In order to facilitate handling a large number of various experimental data and to perform the analysis in perspective, considerations for effective utilization of hybrid computer functions and terminal devices are taken in this code, such as; The experimental data record in an analog data recorder are directly input to the analog part of the hybrid computer. The computed results can be plotted on the graphic display and its hard copy is readily available. A series of messages for guidance is given on the display terminal by which the analysis though man-machine interactive computation can be performed. Thus, the required results can be obtained by performing case studies for which necessary parameters are input through the keyboard and the results displayed are checked. (auth.)
International Nuclear Information System (INIS)
Safety analysis is an important tool for justifying the safety of nuclear power plants. Typically, this type of analysis is performed by means of system computer codes with one dimensional approximation for modelling real plant systems. However, in the nuclear area there are issues for which traditional treatment using one dimensional system codes is considered inadequate for modelling local flow and heat transfer phenomena. There is therefore increasing interest in the application of three dimensional computational fluid dynamics (CFD) codes as a supplement to or in combination with system codes. There are a number of both commercial (general purpose) CFD codes as well as special codes for nuclear safety applications available. With further progress in safety analysis techniques, the increasing use of CFD codes for nuclear applications is expected. At present, the main objective with respect to CFD codes is generally to improve confidence in the available analysis tools and to achieve a more reliable approach to safety relevant issues. An exchange of views and experience can facilitate and speed up progress in the implementation of this objective. Both the International Atomic Energy Agency (IAEA) and the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development (OECD/NEA) believed that it would be advantageous to provide a forum for such an exchange. Therefore, within the framework of the Working Group on the Analysis and Management of Accidents of the NEA's Committee on the Safety of Nuclear Installations, the IAEA and the NEA agreed to jointly organize the Technical Meeting on the Use of Computational Fluid Dynamics Codes for Safety Analysis of Reactor Systems, including Containment. The meeting was held in Pisa, Italy, from 11 to 14 November 2002. The entire collection of papers is provided in this report
Calibration and simulation of a HPGe well detector using Monte Carlo computer code
International Nuclear Information System (INIS)
Monte Carlo methods are often used in simulating physical and mathematical systems. This computer code is a class of computational algorithms that rely on repeated random sampling to compute their results. Because of their reliance on repeated computation of random or pseudo-random numbers, these methods are most suited to calculation by a computer and tend to be used when it is unfeasible or impossible to compute an exact result with a deterministic algorithm. The Monte Carlo method is used to determine a detector's response curves which are difficult to obtain experimentally. It deals with random numbers for the simulation of the decay conditions and angle of incidence at a given energy value, studying, thus, the random behavior of the radiation, providing response and efficiency curves. The MCNP5 computer code provides means to simulate gamma ray detectors and has been used for this work for the 50keV - 2000 keV energy range. The HPGe well detector was simulated with the MCNP5 computer code and compared with experimental data. The dimensions of both dead layer and the transition layer were determined, and the response curve for a particular geometry was then obtained and compared with the experimental results, in order to verify the detector's simulation. Both results were in very good agreement. (author)
Development of computer code CRIFLAN for critical flow analysis in reactor safety
International Nuclear Information System (INIS)
The computer code Critical Flow Analysis (CRIFLAN) has been developed for critical flow analysis in reactor safety. The code employs a heuristic model to describe critical discharges of water for both sub cooled and saturated conditions. Unlike many large thermal-hydraulic codes (RELAP5/MOD3, ATHLETE, and ASTEC) currently in use, CRIFLAN is a stand alone code and is not system specific. The proposed CRIFLAN model is validated against correlation based models such as H-F, Moody, and HEM and against the published experimental data. The CRIFLAN predictions are in good agreement with H-F results. It is concluded that the CRIFLAN code offers a reliable, user-friendly analytical tool for critical flow analysis in the field of reactor safety. (author)
Development of a system of computer codes for severe accident analysis and its applications
Energy Technology Data Exchange (ETDEWEB)
Jang, S. H.; Chun, S. W.; Jang, H. S. and others [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)
1993-01-15
As a continuing study for the development of a system of computer codes to analyze severe accidents which had been performed last year, major focuses were on the aspect of application of the developed code systems. As the first step, two most commonly used code packages other than STCP, i.e., MELCOR of NRC and MAAP of IDCOR were reviewed to compare the models that they used. Next, important heat transfer phenomena were surveyed as severe accident progressed. Particularly, debris bed coolability and molten core-concrete interaction were selected as sample models, and they were studied extensively. The recent theoretical works and experiments for these phenomena were surveyed, and also the relevant models adopted by major code packages were compared and assessed. Based on the results obtained in this study, it is expected to be able to take into account these phenomenological uncertainties when one uses the severe accident code packages for probabilistic safety assessments or accident management programs.
A Multiple Sphere T-Matrix Fortran Code for Use on Parallel Computer Clusters
Mackowski, D. W.; Mishchenko, M. I.
2011-01-01
A general-purpose Fortran-90 code for calculation of the electromagnetic scattering and absorption properties of multiple sphere clusters is described. The code can calculate the efficiency factors and scattering matrix elements of the cluster for either fixed or random orientation with respect to the incident beam and for plane wave or localized- approximation Gaussian incident fields. In addition, the code can calculate maps of the electric field both interior and exterior to the spheres.The code is written with message passing interface instructions to enable the use on distributed memory compute clusters, and for such platforms the code can make feasible the calculation of absorption, scattering, and general EM characteristics of systems containing several thousand spheres.
ATHLET-CD and COCOSYS: the mechanistic computer codes of GRS for simulating severe accidents
International Nuclear Information System (INIS)
Simulating accident sequences within the framework of safety analyses of nuclear power plants requires the use of deterministic computer codes furnishing the most realistic results (best estimates) in the light of the state of the art. This requirement exists for design basis accidents as well as accidents and events exceeding the design basis. For simulations of reactor behavior and of the source term from the nuclear steam supply system, the ATHLET (Analysis of Thermohydraulics of Leaks and Transients) code has been developed and validated for transients and accidents without major core damage, and the ATHLET-CD (Core Degradation) code has been developed and validated for accidents resulting in major core damage, while the COCOSYS (Containment Code System) code has been developed and validated for the behavior of the containment and the source term for the environment. (orig.)
LEADS-DC: A computer code for intense dc beam nonlinear transport simulation
Institute of Scientific and Technical Information of China (English)
无
2011-01-01
An intense dc beam nonlinear transport code has been developed. The code is written in Visual FORTRAN 6.6 and has ~13000 lines. The particle distribution in the transverse cross section is uniform or Gaussian. The space charge forces are calculated by the PIC (particle in cell) scheme, and the effects of the applied fields on the particle motion are calculated with the Lie algebraic method through the third order approximation. Obviously,the solutions to the equations of particle motion are self-consistent. The results obtained from the theoretical analysis have been put in the computer code. Many optical beam elements are contained in the code. So, the code can simulate the intense dc particle motions in the beam transport lines, high voltage dc accelerators and ion implanters.
Abstracts of digital computer code packages. Assembled by the Radiation Shielding Information Center
International Nuclear Information System (INIS)
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
Computer codes used in the calculation of high-temperature thermodynamic properties of sodium
International Nuclear Information System (INIS)
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
Development of a system of computer codes for severe accident analysis and its applications
International Nuclear Information System (INIS)
As a continuing study for the development of a system of computer codes to analyze severe accidents which had been performed last year, major focuses were on the aspect of application of the developed code systems. As the first step, two most commonly used code packages other than STCP, i.e., MELCOR of NRC and MAAP of IDCOR were reviewed to compare the models that they used. Next, important heat transfer phenomena were surveyed as severe accident progressed. Particularly, debris bed coolability and molten core-concrete interaction were selected as sample models, and they were studied extensively. The recent theoretical works and experiments for these phenomena were surveyed, and also the relevant models adopted by major code packages were compared and assessed. Based on the results obtained in this study, it is expected to be able to take into account these phenomenological uncertainties when one uses the severe accident code packages for probabilistic safety assessments or accident management programs
A multiple sphere T-matrix Fortran code for use on parallel computer clusters
International Nuclear Information System (INIS)
A general-purpose Fortran-90 code for calculation of the electromagnetic scattering and absorption properties of multiple sphere clusters is described. The code can calculate the efficiency factors and scattering matrix elements of the cluster for either fixed or random orientation with respect to the incident beam and for plane wave or localized-approximation Gaussian incident fields. In addition, the code can calculate maps of the electric field both interior and exterior to the spheres. The code is written with message passing interface instructions to enable the use on distributed memory compute clusters, and for such platforms the code can make feasible the calculation of absorption, scattering, and general EM characteristics of systems containing several thousand spheres.
Window-based computer code package CONPAS for an integrated level 2 PSA
International Nuclear Information System (INIS)
A PC window-based computer code. CONPAS (CONtainment Performance Analysis System), has been developed to integrate the numerical, graphical, and results-operation aspects of Level2 probabilistic safety assessments (PSA) for nuclear power plants automatically. As a main logic for accident progression analysis, it employs a concept of the small containment phenomenological event tree (CPET) helpful to trace out visually individual accident progressions and of the large supporting event tree (LSET) for its detailed quantification. Compared with other existing computer codes for Level 2 PSA, the CONPAS code provides several advanced features: computational aspects including systematic uncertainty analysis, importance analysis, and sensitivity analysis, reporting aspects including tabling and graphic, and user-friend interface
A first accident simulation for Angra-1 power plant using the ALMOD computer code
International Nuclear Information System (INIS)
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)
International Nuclear Information System (INIS)
The purpose of this paper is to discuss the theories, techniques and computer codes that are frequently used in numerical reactor criticality and burnup calculations. It is a part of an integrated nuclear reactor calculation scheme conducted by the Reactors Department, Inshas Nuclear Research Centre. The crude part in numerical reactor criticality and burnup calculations includes the determination of neutron flux distribution which can be obtained in principle as a solution of Boltzmann transport equation. Numerical methods used for solving transport equations are discussed. Emphasis are made on numerical techniques based on multigroup diffusion theory. These numerical techniques include nodal, modal, and finite difference ones. The most commonly known computer codes utilizing these techniques are reviewed. Some of the main computer codes that have been already developed at the Reactors Department and related to numerical reactor criticality and burnup calculations have been presented
FLAME: A finite element computer code for contaminant transport n variably-saturated media
International Nuclear Information System (INIS)
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
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)
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
FURN3D: A computer code for radiative heat transfer in pulverized coal furnaces
Energy Technology Data Exchange (ETDEWEB)
Ahluwalia, R.K.; Im, K.H.
1992-08-01
A computer code FURN3D has been developed for assessing the impact of burning different coals on heat absorption pattern in pulverized coal furnaces. The code is unique in its ability to conduct detailed spectral calculations of radiation transport in furnaces fully accounting for the size distributions of char, soot and ash particles, ash content, and ash composition. The code uses a hybrid technique of solving the three-dimensional radiation transport equation for absorbing, emitting and anisotropically scattering media. The technique achieves an optimal mix of computational speed and accuracy by combining the discrete ordinate method (S{sub 4}), modified differential approximation (MDA) and P, approximation in different range of optical thicknesses. The code uses spectroscopic data for estimating the absorption coefficients of participating gases C0{sub 2}, H{sub 2}0 and CO. It invokes Mie theory for determining the extinction and scattering coefficients of combustion particulates. The optical constants of char, soot and ash are obtained from dispersion relations derived from reflectivity, transmissivity and extinction measurements. A control-volume formulation is adopted for determining the temperature field inside the furnace. A simple char burnout model is employed for estimating heat release and evolution of particle size distribution. The code is written in Fortran 77, has modular form, and is machine-independent. The computer memory required by the code depends upon the number of grid points specified and whether the transport calculations are performed on spectral or gray basis.
FURN3D: A computer code for radiative heat transfer in pulverized coal furnaces
Energy Technology Data Exchange (ETDEWEB)
Ahluwalia, R.K.; Im, K.H.
1992-08-01
A computer code FURN3D has been developed for assessing the impact of burning different coals on heat absorption pattern in pulverized coal furnaces. The code is unique in its ability to conduct detailed spectral calculations of radiation transport in furnaces fully accounting for the size distributions of char, soot and ash particles, ash content, and ash composition. The code uses a hybrid technique of solving the three-dimensional radiation transport equation for absorbing, emitting and anisotropically scattering media. The technique achieves an optimal mix of computational speed and accuracy by combining the discrete ordinate method (S[sub 4]), modified differential approximation (MDA) and P, approximation in different range of optical thicknesses. The code uses spectroscopic data for estimating the absorption coefficients of participating gases C0[sub 2], H[sub 2]0 and CO. It invokes Mie theory for determining the extinction and scattering coefficients of combustion particulates. The optical constants of char, soot and ash are obtained from dispersion relations derived from reflectivity, transmissivity and extinction measurements. A control-volume formulation is adopted for determining the temperature field inside the furnace. A simple char burnout model is employed for estimating heat release and evolution of particle size distribution. The code is written in Fortran 77, has modular form, and is machine-independent. The computer memory required by the code depends upon the number of grid points specified and whether the transport calculations are performed on spectral or gray basis.
Computing the effects of a contained sodium sheet fire: The 'FEUNA' code
International Nuclear Information System (INIS)
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 400m3 caisson. (author)
International Nuclear Information System (INIS)
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.)
Shift-scale invariance based computer code for wiggler radiation simulation
Smolyakov, N V
2001-01-01
A new package of computer codes for calculating incoherent electromagnetic radiation from a relativistic electron beam moving in arbitrary three-dimensional magnetic field is developed at Hiroshima University. The codes are able to accept either an experimentally measured magnetic field or numerically simulated field map (with field errors, if necessary). The near-field effects as well as the electron beam emittance effects are also included into simulation. The codes are based on the shift-scale invariance property of radiation spectra that enables us to reduce considerably the bulk of individual calculations of single electron radiation.
VARSKIN MOD 2 and SADDE MOD2: Computer codes for assessing skin dose from skin contamination
International Nuclear Information System (INIS)
The computer code VARSKIN has been modified to calculate dose to skin from three-dimensional sources, sources separated from the skin by layers of protective clothing, and gamma dose from certain radionuclides correction for backscatter has also been incorporated for certain geometries. This document describes the new code, VARSKIN Mod 2, including installation and operation instructions, provides detailed descriptions of the models used, and suggests methods for avoiding misuse of the code. The input data file for VARSKIN Mod 2 has been modified to reflect current physical data, to include the contribution to dose from internal conversion and Auger electrons, and to reflect a correction for low-energy electrons. In addition, the computer code SADDE: Scaled Absorbed Dose Distribution Evaluator has been modified to allow the generation of scaled absorbed dose distributions for mixtures of radionuclides and intereat conversion and Auger electrons. This new code, SADDE Mod 2, is also described in this document. Instructions for installation and operation of the code and detailed descriptions of the models used in the code are provided
Development of computer code packages for molten salt reactor core analysis
International Nuclear Information System (INIS)
This paper presents the implementations of the Oak Ridge National Laboratory (ORNL) approach for Molten Salt Reactor (MSR) core analysis with two nuclear reactor core analysis computer code systems. The first code system has been set up with the MCNP6 Monte Carlo code, its depletion module CINDER90 and the PYTHON script language. The second code system has been set up with the NEWT transport calculation module and ORIGEN depletion module connected by TRITON sequence in SCALE code, and the PYTHON script language. The PYTHON script language is used for implementing the online reprocessing of molten-salt fuel, and feeding new fertile material in the computer code simulations. In this paper, simplified nuclear reactor core models of a Molten Salt Breeder Reactor (MSBR), designed by ORNL in the 1960's, and FUJI-U3 designed by Toyohashi University of Technology (TUT) in the 2000's, were analyzed by the two code systems. Using these, various reactor design parameters of the MSRs were compared, such as the multiplication factor, breeding ratio, amount of material, total feeding, neutron flux distribution, and temperature coefficient. (author)
Computer code TOBUNRAD for PWR fuel bundle heat-up calculations
International Nuclear Information System (INIS)
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)
International Nuclear Information System (INIS)
The Nuclear Regulatory Authority is performing an analysis with PC CREAM, developed at the NRPB, for updating computer programs and models used for calculating the transfer of radionuclides through the environment. For CREAM dose assessment verification for local scenarios, this paper presents a comparison of population doses assessed with the computer codes used nowadays and with CREAM, for unitary releases of main radionuclides in nuclear power plant discharges. The results of atmospheric dispersion processes and the transfer of radionuclides through the environment for local scenarios are analysed. The programs used are PLUME for atmospheric dispersion, FARMLAND for the transfer of radionuclides into foodstuffs following atmospheric deposition in the terrestrial environment and ASSESSOR for individual and collective dose assessments.This paper presents the general assumptions made for dose assessments. The results show some differences between doses due to differences in models, in the complexity level of the same models, or in parameters. (author)
NSR-77: a computer code for transient analysis of a light water reactor fuel rod
International Nuclear Information System (INIS)
This report describes computer code NSR-77 written in FORTRAN IV for FACOM-M 200 computer in detail. It has been developed for transient response analysis of a light water reactor fuel rod during an accident such as a reactivityy initiated accident, a loss-of-coolant accident or a power-cooling-mismatch accident. The code consists of subcodes which calculate heat conduction in a fuel rod, gas gap conductance between fuel and cladding, heat transfer from cladding to coolant, fluid hydrodynamics, elastic-plastic fuel and cladding deformation, and material properties, and so on. (author)
Solution of 3-dimensional time-dependent viscous flows. Part 2: Development of the computer code
Weinberg, B. C.; Mcdonald, H.
1980-01-01
There is considerable interest in developing a numerical scheme for solving the time dependent viscous compressible three dimensional flow equations to aid in the design of helicopter rotors. The development of a computer code to solve a three dimensional unsteady approximate form of the Navier-Stokes equations employing a linearized block emplicit technique in conjunction with a QR operator scheme is described. Results of calculations of several Cartesian test cases are presented. The computer code can be applied to more complex flow fields such as these encountered on rotating airfoils.
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.
International Nuclear Information System (INIS)
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
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)
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
International Nuclear Information System (INIS)
This report describes a development of a wind field calculation code and an atmospheric dispersion and dose calculation code which can be used for real-time prediction in an emergency. Models used in the computer codes are a mass-consistent model for wind field and a particle diffusion model for atmospheric dispersion. In order to attain quick response even when the codes are used in a small-scale computer, high-speed iteration method (MILUCR) and kernel density method are applied to the wind field model and the atmospheric and dose calculation model, respectively. In this report, numerical models, computational codes, related files and calculation examples are shown. (author)
De novo computational prediction of non-coding RNA genes in prokaryotic genomes
Tran, Thao T.; Zhou, Fengfeng; Marshburn, Sarah; Stead, Mark; Kushner, Sidney R.; Xu, Ying
2009-01-01
Motivation: The computational identification of non-coding RNA (ncRNA) genes represents one of the most important and challenging problems in computational biology. Existing methods for ncRNA gene prediction rely mostly on homology information, thus limiting their applications to ncRNA genes with known homologues. Results: We present a novel de novo prediction algorithm for ncRNA genes using features derived from the sequences and structures of known ncRNA genes in comparison to decoys. Using...
International Nuclear Information System (INIS)
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
Some questions of using coding theory and analytical calculation methods on computers
International Nuclear Information System (INIS)
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
An improved UO2 thermal conductivity model in the ELESTRES computer code
International Nuclear Information System (INIS)
This paper describes the improved UO2 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 UO2 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 UO2 fuel with different densities. The UO2 thermal conductivity data cover 90% to 99% theoretical density of UO2, 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)
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.
Validation of thermal hydraulic computer codes for advanced light water reactor
International Nuclear Information System (INIS)
The Czech Republic operates 4 WWER-440 units, two WWER-1000 units are being finalised (one of them is undergoing commissioning). Thermal-hydraulics Department of the Nuclear Research Institute Rez performs accident analyses for these plants using a number of computer codes. To model the primary and secondary circuits behaviour the system codes ATHLET, CATHARE, RELAP, TRAC are applied. Containment and pressure-suppressure system are modelled with RALOC and MELCOR codes, the reactor power calculations (point and space-neutron kinetics) are made with DYN3D, NESTLE and CDF codes (FLUENT, TRIO) are used for some specific problems. An integral part of the current Czech project 'New Energy Sources' is selection of a new nuclear source. Within this and the preceding projects financed by the Czech Ministry of Industry and Trade and the EU PHARE, the Department carries and has carried out the systematic validation of thermal-hydraulic and reactor physics computer codes applying data obtained on several experimental facilities as well as the real operational data. The paper provides a concise information on these activities of the NRI and its Thermal-hydraulics Department. A detailed example of the system code validation and the consequent utilisation of the results for a real NPP purposes is included. (author)
RAP-2A Computer code for transients analysis in fast reactors
International Nuclear Information System (INIS)
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
Validation and uncertainty analysis of the Athlet thermal-hydraulic computer code
International Nuclear Information System (INIS)
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)
International Nuclear Information System (INIS)
Two experimental campaigns were carried out to verify: 1) the method of assessing the mean kerma in a household used in the computer code BILL calculating the protection factor afforded by dwellings; 2) in what conditions the kerma calculated in cubic meshes of a given size (code PIECE) agreed with TLD measurements. To that purpose, a house was built near the caesium 137 source of the Ecosystem irradiator located at the Cadarache Nuclear Research Center. During the first campaign, four experiments with different house characteristics were conducted. Some 50 TLSs locations describing the inhabitable volume were defined in order to obtain the mean kerma. 16 locations were considered outside the house. During the second campaign a cobalt 60 source was installed on the side. Only five measurement locations were defined, each with 6 TLDs. The results of dosimetric measurements are presented and compared with the calculations of the two computer codes. The effects of wall heterogeneity were also studied
Sodium combustion computer code ASSCOPS version 2.1. User's manual
International Nuclear Information System (INIS)
ASSCOPS (Analysis of Simultaneous Sodium Combustion in Pool and Spray) has been developed for analyses of thermal consequences of sodium leak and fire accidents in LMFBRs. This report presents a description of the computational models, input and output data as the user's manual of ASSCOPS version 2.1. ASSCOPS is an integrated computational code based on the sodium pool fire code SOFIRE II developed by the Atomics International Division of Rockwell International, and on the sodium spray fire code SPRAY developed by the Hanford Engineering Development Laboratory in the U.S. The users of ASSCOPS need to specify the sodium leak conditions (leak flow rate and temperature, etc.), the cell geometries (cell volume, surface area and thickness of structures, etc.), and the atmospheric initial conditions such as gas temperature, pressure, and composition. ASSCOPS calculates the time histories of atmospheric temperature, pressure and of structural temperature. (author)
Modeling of BWR core meltdown accidents - for application in the MELRPI.MOD2 computer code
International Nuclear Information System (INIS)
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
Solving linear systems in FLICA-4, thermohydraulic code for 3-D transient computations
International Nuclear Information System (INIS)
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
DABIE: a data banking system of integral experiments for reactor core characteristics computer codes
International Nuclear Information System (INIS)
A data banking system of integral experiments for reactor core characteristics computer codes, DABIE, has been developed to lighten the burden on searching so many documents to obtain experiment data required for verification of reactor core characteristics computer code. This data banking system, DABIE, has capabilities of systematic classification, registration and easy retrieval of experiment data. DABIE consists of data bank and supporting programs. Supporting programs are data registration program, data reference program and maintenance program. The system is designed so that user can easily register information of experiment systems including figures as well as geometry data and measured data or obtain those data through TSS terminal interactively. This manual describes the system structure, how-to-use and sample uses of this code system. (author)
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.
Algorithms and computer codes for atomic and molecular quantum scattering theory. Volume I
International Nuclear Information System (INIS)
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
ALICE-87 (Livermore). Precompound Nuclear Model Code. Version for Personal Computer IBM/AT
International Nuclear Information System (INIS)
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
EMPIRE. Pre-equilibrium/Compound Nuclear Model Code for Personal Computer
International Nuclear Information System (INIS)
The pre-equilibrium/compound nuclear model code EMPIRE was implemented for use on personal computer. It is available on a set of diskettes from the NEA Data Bank of Nuclear Energy Agency (Saclay, France) and the IAEA Nuclear Data Section. (author). Refs, figs and tabs
CPS: a continuous-point-source computer code for plume dispersion and deposition calculations
Energy Technology Data Exchange (ETDEWEB)
Peterson, K.R.; Crawford, T.V.; Lawson, L.A.
1976-05-21
The continuous-point-source computer code calculates concentrations and surface deposition of radioactive and chemical pollutants at distances from 0.1 to 100 km, assuming a Gaussian plume. The basic input is atmospheric stability category and wind speed, but a number of refinements are also included.
Energy Technology Data Exchange (ETDEWEB)
Shibata, C.S.; Montes, A. [Instituto de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil); Galvao, R.M.O. [Sao Paulo Univ., SP (Brazil). Inst. de Fisica
1994-04-01
This paper describes the `FLINESH` computer code for magnetic fields calculation developed for the simulation of field configurations in plasma magnetic confinement devices. The expressions for the poloidal field and flux, the program structure and the input parameters description are presented, and also the analysis of the graphic output possibilities. (L.C.J.A.). 12 refs, 14 figs, 2 tabs.
Ivanov, Anisoara; Neacsu, Andrei
2011-01-01
This study describes the possibility and advantages of utilizing simple computer codes to complement the teaching techniques for high school physics. The authors have begun working on a collection of open source programs which allow students to compare the results and graphics from classroom exercises with the correct solutions and further more to…
Peculiarities of spent fuel pool modeling using MELCOR 1.8.5 computer code
International Nuclear Information System (INIS)
Processes that take place in spent nuclear fuel pools are analyzed using the MELCOR 1.8.5. computer code with options for modeling PWR and BWR reactor types. Transients are considered for the initiating events loss of heat removal from the spent fuel pool (SFP) and basic differences in time frames and processes occurring in SFP are determined
SIVAR - Computer code for simulation of fuel rod behavior in PWR during fast transients
International Nuclear Information System (INIS)
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.)
SACRD: a data base for fast reactor safety computer codes, general description
Energy Technology Data Exchange (ETDEWEB)
Greene, N.M.; Forsberg, V.M.; Raiford, G.B.; Arwood, J.W.; Simpson, D.B.; Flanagan, G.F.
1979-01-01
SACRD is a data base of material properties and other handbook data needed in computer codes used for fast reactor safety studies. Data are available in the thermodynamics, heat transfer, fluid mechanics, structural mechanics, aerosol transport, meteorology, neutronics, and dosimetry areas. Tabular, graphical and parameterized data are provided in many cases. A general description of the SACRD system is presented in the report.
Burnup calculations using the ORIGEN code in the CONKEMO computing system
International Nuclear Information System (INIS)
This article describes the CONKEMO computing system for kinetic multigroup calculations of nuclear reactors and their physical characteristics during burnup. The ORIGEN burnup calculation code has been added to the system. The results of an international benchmark calculation are also presented. (author)
Simplified 3D model of a PWR reactor vessel using fluid dynamics code ANSYS CFX computational
International Nuclear Information System (INIS)
This paper presents the results from the calculation of the steady state simulation with model of CFD (computational fluid dynamic) operating under conditions of operation at full power (Hot Full Power). Development and the CFD model results show the usefulness of these codes for calculating 3D of the variable thermohydraulics of these reactors.
Application of Multiple Description Coding for Adaptive QoS Mechanism for Mobile Cloud Computing
Directory of Open Access Journals (Sweden)
Ilan Sadeh
2014-02-01
Full Text Available Multimedia transmission over cloud infrastructure is a hot research topic worldwide. It is very strongly related to video streaming, VoIP, mobile networks, and computer networks. The goal is a reliable integration of telephony, video and audio transmission, computing and broadband transmission based on cloud computing. One right approach to pave the way for mobile multimedia and cloud computing is Multiple Description Coding (MDC, i.e. the solution would be: TCP/IP and similar protocols to be used for transmission of text files, and Multiple Description Coding “Send and Forget” algorithm to be used as transmission method for Multimedia over the cloud. Multiple Description Coding would improve the Quality of Service and would provide new service of rate adaptive streaming. This paper presents a new approach for improving the quality of multimedia and other services in the cloud, by using Multiple Description Coding (MDC. Firsty MDC Send and Forget Algorithm is compared with the existing protocols such as TCP/IP, UDP, RTP, etc. Then the Achievable Rate Region for MDC system is evaluated. Finally, a new subset of Quality of Service that considers the blocking in multi-terminal multimedia network and fidelity losses is considered.
International Nuclear Information System (INIS)
The computer code SSICC (Safety and Stability of Internally Cooled Conductors) has successfully simulated the multiple stability regions observed experimentally by Lue, Miller, and Dresner of Oak Ridge National Laboratory. The simulation requires asymmetrical boundary conditions and a heating pulse duration short compared to the time for reflection of the transient pressure wave back into the heated region of the conductor
The community project COSA: comparison of geo-mechanical computer codes for salt
International Nuclear Information System (INIS)
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
Validation matrices for the computer codes used for ACR safety analysis
International Nuclear Information System (INIS)
This paper discusses the process being used to prepare Validation Matrix Documents (VMs) for validation of the computer codes which will be used for ACR safety analysis. Computer-program validation methodology is organized as a multi-stage process associated with a structured set of documents. The first two stages are implemented with the preparation of a Technical Basis Document and Validation Matrix Documents, which are prepared without reference to specific versions of computer programs. The remaining stages of the process involve the completion of Validation Plans, Validation Exercises and Validation Manuals for specific computer programs. The key features of the ACR design which impact the physical phenomena and the expected ranges of key parameters during postulated accidents are briefly described. An overview is given of the experiments being performed specifically to support the qualification of the safety analysis codes to be used for ACR analysis. Inclusion of these new data sources in the VMs is an important element to ensure that a sound basis is in place for code validation. It is anticipated that the VMs will play a key role in supporting the US NRC's review of the ACR technology base used in the safety analysis code qualification program. The role played by the VMs in linking the fundamental phenomena and the set of design basis events, is also covered in the paper. (author)
A Computer Code For Evaluation of Design Parameters of Concrete Piercing Earth Shock Missile Warhead
Directory of Open Access Journals (Sweden)
P. K. Roy
1985-10-01
Full Text Available A simple and reliable computer code has been devised for evaluating various design parameters, and predicting the penetration performance of concrete piercing earth shock missile-warhead and will be useful to the designers of earth penetrating weapon system.
User's manual for the vertical axis wind turbine performance computer code darter
Energy Technology Data Exchange (ETDEWEB)
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 (SSImple MOmentum, Single Streamtube) and DART (SARrieus Turbine) to DARTER.
JPDYN-IV: a computer code for JPDR-II dynamics analysis
International Nuclear Information System (INIS)
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)
COSMOCR A Numerical Code for Cosmic Ray Studies in Computational Cosmology
Miniati, F
2001-01-01
We present COSMOCR, a numerical code for the investigation of cosmic ray related studies in computational cosmology. The code follows the diffusive shock acceleration, the mechanical and radiative energy losses and the spatial transport of the supra-thermal particles in cosmic environment. Primary cosmic ray electrons and ions are injected at shocks according to the thermal leakage prescription. Secondary electrons are continuously injected as a results of p-p inelastic collisions of primary cosmic ray ions and thermal background nuclei. The code consists of a conservative, finite volume method with a power-law sub-grid model in momentum space. Two slightly different schemes are implemented depending on the stiffness of the cooling terms. Comparisons of numerical results with analytical solution for a number of tests of direct interest show remarkable performance of the present code.
International Nuclear Information System (INIS)
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
The computer code CONDIF-01 (release 2) for transient convective-conductive heat transfer
International Nuclear Information System (INIS)
CONDIF-01 is a finite element computer code developed at J.R.C. Ispra to solve natural and forced convection problems, for use in Post Accident Heat Removal studies following a hypothetical fast-reactor core meltdown. The new version of the code is capable of analysing problems in which there exists initially a liquid (solid) region which may change phase to solid (liquid), as time proceeds. A variant of the enthalpy method is employed to model the phase change process. The presence of structures enclosing the liquid (solid) region is accounted for, but such structures are assumed to remain in the solid phase. Plane and axisymmetric situations may be analysed. The essential characteristics of the code are outlined here. This report gives instructions for preparing input data to CONDIF-01, release 2, and describes two test problems in order to illustrate both the input and the output of the code
''HFLOWR'' - a computer code for predicting flow regimes in a horizontal channel with rod cluster
International Nuclear Information System (INIS)
Indian Pressurised Heavy Water Reactors have horizontal channels containing rod clusters. Occurrence of flow stratification under postulated accident conditions , significantly affects the temperatures of the exposed fuel rods. The various thermodynamic models that are to be used for the analysis depend on the flow regimes, which are expected to occur in the channel. A computer program HFLOWR has been developed to obtain the dimensionless flow regime maps and flow patterns in reactor channels of Indian PHWR under various coolant conditions. This code can be integrated with the safety analysis code, for flow regime prediction under transient/accident conditions. This report describes the various mathematical formulations used in the code and the various analyses done using this code. (author). 6 refs., 1 appendix, 34 figs
Verification of IVA5 computer code for melt-water interaction analysis: Part 2
International Nuclear Information System (INIS)
Experiments with real molten material, especially with prototypic reactor materials, are very important in the course of this validation. Now the author steps forward to model four of the FARO tests performed by JRC Ispra Italy. The purpose of this Chapter is to verify the code for a class of processes called non-explosive melt-water interaction. All elements of the code are addressed in such a complicated simulation, e.g. code architecture, numerical methods, constitutive models, etc. The simulation of the FARO test will reveal the capability of the code to handle such flows with simultaneous evaporation condensation fragmentation of all participating velocity fields, etc. Among these, there are some important constitutive models which are addressed during the computation and their performance within the overall system will be also tested
Methods, algorithms and computer codes for calculation of electron-impact excitation parameters
Bogdanovich, P; Stonys, D
2015-01-01
We describe the computer codes, developed at Vilnius University, for the calculation of electron-impact excitation cross sections, collision strengths, and excitation rates in the plane-wave Born approximation. These codes utilize the multireference atomic wavefunctions which are also adopted to calculate radiative transition parameters of complex many-electron ions. This leads to consistent data sets suitable in plasma modelling codes. Two versions of electron scattering codes are considered in the present work, both of them employing configuration interaction method for inclusion of correlation effects and Breit-Pauli approximation to account for relativistic effects. These versions differ only by one-electron radial orbitals, where the first one employs the non-relativistic numerical radial orbitals, while another version uses the quasirelativistic radial orbitals. The accuracy of produced results is assessed by comparing radiative transition and electron-impact excitation data for neutral hydrogen, helium...
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.
TEMP: a computer code to calculate fuel pin temperatures during a transient
International Nuclear Information System (INIS)
The computer code TEMP calculates fuel pin temperatures during a transient. It was developed to accommodate temperature calculations in any system of axi-symmetric concentric cylinders. When used to calculate fuel pin temperatures, the code will handle a fuel pin as simple as a solid cylinder or as complex as a central void surrounded by fuel that is broken into three regions by two circumferential cracks. Any fuel situation between these two extremes can be analyzed along with additional cladding, heat sink, coolant or capsule regions surrounding the fuel. The one-region version of the code accurately calculates the solution to two problems having closed-form solutions. The code uses an implicit method, an explicit method and a Crank-Nicolson (implicit-explicit) method
International Nuclear Information System (INIS)
As world-wide nuclear facilities continue to grow, the volume of movements between facilities, and the variety of physical forms, sizes, and activity levels, also continue to increase. With this growth there is a need for a better understanding of the regulations relating to the packaging and transport of radioactive materials, the categories and characteristics of the various groups of material, the types of package, and the responsibilities of Consignors and Competent Authorities. Interpretation of the regulations can be a long and involved task. Croft Associates, with the approval of the IAEA, have developed and marketed a search and find software package (Croft Associates Regulations Referencer - CARR) to enable users to locate quickly those parts of the regulations relating to a specific topic defined by the user. (author)
A directory of computer codes suitable for stress analysis of HLW containers - Compas project
International Nuclear Information System (INIS)
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
Recommendations for computer modeling codes to support the UMTRA groundwater restoration project
International Nuclear Information System (INIS)
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
A proposed framework for computational fluid dynamics code calibration/validation
Energy Technology Data Exchange (ETDEWEB)
Oberkampf, W.L.
1993-12-31
The paper reviews the terminology and methodology that have been introduced during the last several years for building confidence n the predictions from Computational Fluid Dynamics (CID) codes. Code validation terminology developed for nuclear reactor analyses and aerospace applications is reviewed and evaluated. Currently used terminology such as ``calibrated code,`` ``validated code,`` and a ``validation experiment`` is discussed along with the shortcomings and criticisms of these terms. A new framework is proposed for building confidence in CFD code predictions that overcomes some of the difficulties of past procedures and delineates the causes of uncertainty in CFD predictions. Building on previous work, new definitions of code verification and calibration are proposed. These definitions provide more specific requirements for the knowledge level of the flow physics involved and the solution accuracy of the given partial differential equations. As part of the proposed framework, categories are also proposed for flow physics research, flow modeling research, and the application of numerical predictions. The contributions of physical experiments, analytical solutions, and other numerical solutions are discussed, showing that each should be designed to achieve a distinctively separate purpose in building confidence in accuracy of CFD predictions. A number of examples are given for each approach to suggest methods for obtaining the highest value for CFD code quality assurance.
A proposed framework for computational fluid dynamics code calibration/validation
International Nuclear Information System (INIS)
The paper reviews the terminology and methodology that have been introduced during the last several years for building confidence n the predictions from Computational Fluid Dynamics (CID) codes. Code validation terminology developed for nuclear reactor analyses and aerospace applications is reviewed and evaluated. Currently used terminology such as ''calibrated code,'' ''validated code,'' and a ''validation experiment'' is discussed along with the shortcomings and criticisms of these terms. A new framework is proposed for building confidence in CFD code predictions that overcomes some of the difficulties of past procedures and delineates the causes of uncertainty in CFD predictions. Building on previous work, new definitions of code verification and calibration are proposed. These definitions provide more specific requirements for the knowledge level of the flow physics involved and the solution accuracy of the given partial differential equations. As part of the proposed framework, categories are also proposed for flow physics research, flow modeling research, and the application of numerical predictions. The contributions of physical experiments, analytical solutions, and other numerical solutions are discussed, showing that each should be designed to achieve a distinctively separate purpose in building confidence in accuracy of CFD predictions. A number of examples are given for each approach to suggest methods for obtaining the highest value for CFD code quality assurance
An Object-Oriented Computer Code for Aircraft Engine Weight Estimation
Tong, Michael T.; Naylor, Bret A.
2009-01-01
Reliable engine-weight estimation at the conceptual design stage is critical to the development of new aircraft engines. It helps to identify the best engine concept amongst several candidates. At NASA Glenn Research Center (GRC), the Weight Analysis of Turbine Engines (WATE) computer code, originally developed by Boeing Aircraft, has been used to estimate the engine weight of various conceptual engine designs. The code, written in FORTRAN, was originally developed for NASA in 1979. Since then, substantial improvements have been made to the code to improve the weight calculations for most of the engine components. Most recently, to improve the maintainability and extensibility of WATE, the FORTRAN code has been converted into an object-oriented version. The conversion was done within the NASA's NPSS (Numerical Propulsion System Simulation) framework. This enables WATE to interact seamlessly with the thermodynamic cycle model which provides component flow data such as airflows, temperatures, and pressures, etc., that are required for sizing the components and weight calculations. The tighter integration between the NPSS and WATE would greatly enhance system-level analysis and optimization capabilities. It also would facilitate the enhancement of the WATE code for next-generation aircraft and space propulsion systems. In this paper, the architecture of the object-oriented WATE code (or WATE++) is described. Both the FORTRAN and object-oriented versions of the code are employed to compute the dimensions and weight of a 300-passenger aircraft engine (GE90 class). Both versions of the code produce essentially identical results as should be the case.
Validation of computer codes used in safety analyses of CANDU power plants
International Nuclear Information System (INIS)
Since the 1960s, the CANDU industry has been developing and using scientific computer codes for designing and analysing CANDU power plants. In this endeavour, the industry has been following nuclear quality-assurance practices of the day, including verification and validation of design and analysis methodologies. These practices have resulted in a large body of experience and expertise in the development and application of computer codes and their associated documentation. Major computer codes used in safety analyses of operating plants and those under development have been, and continue to be subjected to rigorous processes of development and application. To provide a systematic framework for the validation work done to date and planned for the future, the industry has decided to adopt the methodology of validation matrices for computer-code validation, similar to that developed by the Nuclear Energy Agency of the Organization for Economic Co-operation and Development and focused on thermalhydraulic phenomena in Light Water Reactors (LWR). To manage the development of validation matrices for CANDU power plants and to engage experts who can work in parallel on several topics, the CANDU task has been divided into six scientific disciplines. Teams of specialists in each discipline are developing the matrices. A review of each matrix will show if there are gaps or insufficient data for validation purposes and will thus help to focus future research and development, if needed. Also, the industry is examining its suite of computer codes, and their specific, additional validation needs, if any, will follow from the work on the validation matrices. The team in System Thermalhydraulics is the furthest advanced, since it had the earliest start and the international precedent on LWRs, and has developed its validation matrix. The other teams are at various stages in this multiphase, multi-year program, and their progress to date is presented. (author)
MOSRA-Light; high speed three-dimensional nodal diffusion code for vector computers
International Nuclear Information System (INIS)
MOSRA-Light is a three-dimensional neutron diffusion calculation code for X-Y-Z geometry. It is based on the 4th order polynomial nodal expansion method (NEM). As the 4th order NEM is not sensitive to mesh sizes, accurate calculation is possible by the use of coarse meshes of about 20 cm. The drastic decrease of number of unknowns in a 3-dimensional problem results in very fast computation. Furthermore, it employs newly developed computation algorithm 'boundary separated checkerboard sweep method' appropriate to vector computers. This method is very efficient because the speedup factor by vectorization increases, as a scale of problem becomes larger. Speed-up factor compared to the scalar calculation is from 20 to 40 in the case of PWR core calculation. Considering the both effects by the vectorization and the coarse mesh method, total speedup factor is more than 1000 as compared with conventional scalar code with the finite difference method. MOSRA-Light can be available on most of vector or scalar computers with the UNIX or it's similar operating systems (e.g. freeware like Linux). Users can easily install it by the help of the conversation style installer. This report contains the general theory of NEM, the fast computation algorithm, benchmark calculation results and detailed information for usage of this code including input data instructions and sample input data. (author)
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.
Joint Compute and Forward for the Two Way Relay Channel with Spatially Coupled LDPC Codes
Hern, Brett
2012-01-01
We consider the design and analysis of coding schemes for the binary input two way relay channel with erasure noise. We are particularly interested in reliable physical layer network coding in which the relay performs perfect error correction prior to forwarding messages. The best known achievable rates for this problem can be achieved through either decode and forward or compute and forward relaying. We consider a decoding paradigm called joint compute and forward which we numerically show can achieve the best of these rates with a single encoder and decoder. This is accomplished by deriving the exact performance of a message passing decoder based on joint compute and forward for spatially coupled LDPC ensembles.
ANIGAM: a computer code for the automatic calculation of nuclear group data
International Nuclear Information System (INIS)
The computer code ANIGAM consists mainly of the well-known programmes GAM-I and ANISN as well as of a subroutine which reads the THERMOS cross section library and prepares it for ANISN. ANIGAM has been written for the automatic calculation of microscopic and macroscopic cross sections of light water reactor fuel assemblies. In a single computer run both were calculated, the cross sections representative for fuel assemblies in reactor core calculations and the cross sections of each cell type of a fuel assembly. The calculated data were delivered to EXTERMINATOR and CITATION for following diffusion or burn up calculations by an auxiliary programme. This report contains a detailed description of the computer codes and methods used in ANIGAM, a description of the subroutines, of the OVERLAY structure and an input and output description. (oririg.)
Modeling And Simulation Of Bar Code Scanners Using Computer Aided Design Software
Hellekson, Ron; Campbell, Scott
1988-06-01
Many optical systems have demanding requirements to package the system in a small 3 dimensional space. The use of computer graphic tools can be a tremendous aid to the designer in analyzing the optical problems created by smaller and less costly systems. The Spectra Physics grocery store bar code scanner employs an especially complex 3 dimensional scan pattern to read bar code labels. By using a specially written program which interfaces with a computer aided design system, we have simulated many of the functions of this complex optical system. In this paper we will illustrate how a recent version of the scanner has been designed. We will discuss the use of computer graphics in the design process including interactive tweaking of the scan pattern, analysis of collected light, analysis of the scan pattern density, and analysis of the manufacturing tolerances used to build the scanner.
Speedup of MCACE, a Monte Carlo code for evaluation of shielding safety, by parallel computer, 1
International Nuclear Information System (INIS)
In order to improve the accuracy of shielding analysis, we have modified MCACE, a Monte Carlo code for shielding analysis, to be able to execute on a parallel computer. The suitable algorithms for efficient paralleling has been investigated by static and dynamic analyses of the code. This includes a strategy where new units of batches are assigned to the idling cells dynamically during the execution. The efficiency of paralleling has been measured by using a simulator of a parallel computer. It is found that the load factor of all cells reached nearly 100%, and consequently, it can be said that the most effective paralleling has been achieved. The simulator has estimated the effect of paralleling as the speedup of 7.13 times when a sample problem of 8 batches, 400 particles per one batch, is loaded on parallel computer equipped with 8 cells. (author)
International Nuclear Information System (INIS)
In this work a computer code structure for Fire Protection Measures (FPM) and Fire Fighting Capability (FFC) at Nuclear Power Plants (NPP) is presented. It allows to evaluate the category (satisfactory (s), needs for further evaluation (n), unsatisfactory (u)) to which belongs the given NPP for a self-control in view of an IAEA inspection. This possibility of a self assessment resulted from IAEA documents. Our approach is based on international experience gained in this field and stated in IAEA recommendations. As an illustration we used the FORTRAN programming language statement to make clear the structure of the computer code for the problem taken into account. This computer programme can be conceived so that some literal message in English and Romanian languages be displayed beside the percentage assessments. (author)
ELESTRES 2.1 computer code for high burnup CANDU fuel performance analysis
International Nuclear Information System (INIS)
The ELESTRES (ELEment Simulation and sTRESses) computer code models the thermal, mechanical and micro structural behaviours of 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 in fuel element design analysis and assessments. It is also used to provide initial conditions for evaluating fuel behaviour during high temperature transients. ELESTRES 2.1 was developed for high burnup fuel application, based on an industry standard tool version of the code, through the implementation or modification to code models such as fission gas release, fuel pellet densification, flux depression (radial power distribution in the fuel pellet), fuel pellet thermal conductivity, fuel sheath creep, fuel sheath yield strength, fuel sheath oxidation, two dimensional heat transfer between the fuel pellet and the fuel sheath; and an automatic finite element meshing capability to handle various fuel pellet shapes. The ELESTRES 2.1 code design and development was planned, implemented, verified, validated, and documented in accordance with the AECL software quality assurance program, which meets the requirements of the Canadian Standards Association standard for software quality assurance CSA N286.7-99. This paper presents an overview of the ELESTRES 2.1 code with descriptions of the code's theoretical background, solution methodologies, application range, input data, and interface with other analytical tools. Code verification and validation results, which are also discussed in the paper, have confirmed that ELESTRES 2.1 is capable of modelling important fuel phenomena and the code can be used in the design assessment and the verification of high burnup fuels. (author)
Qualification Programs for Computer Codes Used for Safety and Transient Analysis in Canada
International Nuclear Information System (INIS)
Computer codes are used in the Canadian nuclear industry for design support and safety analysis of CANDU reactors. Although the codes in these suites had been validated against experiment as they were developed and used, the methods were not formal and were therefore difficult to audit for completeness. The Canadian nuclear industry therefore initiated programs of formal verification and validation in the mid 1990's, to demonstrate a very low likelihood of significant errors or un-quantifiable uncertainties in the codes, to provide a documentation base which would establish software quality and to demonstrate to a regulator that applicable requirements have been met. Atomic Energy of Canada Limited (AECL) established a safety analysis code Validation Project and Ontario Power Generation Incorporated (OPGI) - formerly Ontario Hydro - established a Verification and Validation of Safety Analysis Software Project as part of its Integrated Improvement Program (IIP). Underlying forces driving these projects included concerns raised by the regulator regarding the quality of safety analysis software and the need to conform to newly-issued Canadian quality assurance standards, specifically CSA N286.7. The objective of these programs is to formally qualify the safety analysis codes for their intended applications. It was apparent that the formal qualification was costly and that substantial savings in both manpower and expenses could be achieved if the efforts could be combined. AECL and the utilities therefore initiated a systematic evaluation of all safety analysis codes, with a view to selecting a reference code suite with code versions to be validated, verified, and used by the entire industry. This code suite would then form an 'Industry Standard Toolset', or IST. The development of this initiative is described in a companion paper. (authors)
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.
International Nuclear Information System (INIS)
The Nuclear Reload Management Program of the Nuclear Power Division (NPD) of the Electric Power Research Institute (EPRI) has the responsibility for initiating and managing applied research in selected nuclear engineering analysis functions for nuclear utilities. The computer systems that result from the research projects consist of large FORTRAN programs containing elaborate computational algorithms used to access such areas as core physics, fuel performance, thermal hydraulics, and transient analysis. This paper summarizes a study of computing technology trends sponsored by the NPD. The approach taken was to interview hardware and software vendors, industry observers, and utility personnel focusing on expected changes that will occur in the computing industry over the next 3 to 5 yr. Particular emphasis was placed on how these changes will impact engineering/scientific computer code development, maintenance, and use. In addition to the interviews, a workshop was held with attendees from EPRI, Power Computing Company, industry, and utilities. The workshop provided a forum for discussing issues and providing input into EPRI's long-term computer code planning process
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.
Two-phase wall friction model for the trace computer code
International Nuclear Information System (INIS)
The wall drag model in the TRAC/RELAP5 Advanced Computational Engine computer code (TRACE) has certain known deficiencies. For example, in an annular flow regime, the code predicts an unphysical high liquid velocity compared to the experimental data. To address those deficiencies, a new wall frictional drag package has been developed and implemented in the TRACE code to model the wall drag for two-phase flow system code. The modeled flow regimes are (1) annular/mist, (2) bubbly/slug, and (3) bubbly/slug with wall nucleation. The new models use void fraction (instead of flow quality) as the correlating variable to minimize the calculation oscillation. In addition, the models allow for transitions between the three regimes. The annular/mist regime is subdivided into three separate regimes for pure annular flow, annular flow with entrainment, and film breakdown. For adiabatic two-phase bubbly/slug flows, the vapor phase primarily exists outside of the boundary layer, and the wall shear uses single-phase liquid velocity for friction calculation. The vapor phase wall friction drag is set to zero for bubbly/slug flows. For bubbly/slug flows with wall nucleation, the bubbles are presented within the hydrodynamic boundary layer, and the two-phase wall friction drag is significantly higher with a pronounced mass flux effect. An empirical correlation has been studied and applied to account for nucleate boiling. Verification and validation tests have been performed, and the test results showed a significant code improvement. (authors)
GAM-HEAT: A computer code to compute heat transfer in complex enclosures. Revision 2
Energy Technology Data Exchange (ETDEWEB)
Cooper, R.E.; Taylor, J.R.
1992-12-01
This report discusses the GAM{underscore}HEAT code which was developed for heat transfer analyses associated with postulated Double Ended Guilliotine Break Loss Of Coolant Accidents (DEGB LOCA) resulting in a drained reactor vessel. In these analyses the gamma radiation resulting from fission product decay constitutes the primary source of energy as a function of time. This energy is deposited into the various reactor components and is re-radiated as thermal energy. The code accounts for all radiant heat exchanges within and leaving the reactor enclosure. The SRS reactors constitute complex radiant exchange enclosures since there are many assemblies of various types within the primary enclosure and most of the assemblies themselves constitute enclosures. GAM-HEAT accounts for this complexity by processing externally generated view factors and connectivity matrices as discussed below, and also accounts for convective, conductive, and advective heat exchanges. The code is structured such that it is applicable for many situations involving heat exchange between surfaces within a radiatively passive medium.
GAM-HEAT: A computer code to compute heat transfer in complex enclosures
Energy Technology Data Exchange (ETDEWEB)
Cooper, R.E.; Taylor, J.R.
1992-12-01
This report discusses the GAM[underscore]HEAT code which was developed for heat transfer analyses associated with postulated Double Ended Guilliotine Break Loss Of Coolant Accidents (DEGB LOCA) resulting in a drained reactor vessel. In these analyses the gamma radiation resulting from fission product decay constitutes the primary source of energy as a function of time. This energy is deposited into the various reactor components and is re-radiated as thermal energy. The code accounts for all radiant heat exchanges within and leaving the reactor enclosure. The SRS reactors constitute complex radiant exchange enclosures since there are many assemblies of various types within the primary enclosure and most of the assemblies themselves constitute enclosures. GAM-HEAT accounts for this complexity by processing externally generated view factors and connectivity matrices as discussed below, and also accounts for convective, conductive, and advective heat exchanges. The code is structured such that it is applicable for many situations involving heat exchange between surfaces within a radiatively passive medium.
International Nuclear Information System (INIS)
Activities within the frame of the IAEA's Technical Working Group on Advanced Technologies for HWRs (TWG-HWR) are conducted in a project within the IAEA's subprogramme on nuclear power reactor technology development. The objective of the activities on HWRs is to foster, within the frame of the TWG-HWR, information exchange and co-operative research on technology development for current and future HWRs, with an emphasis on safety, economics and fuel resource sustainability. One of the activities recommended by the TWG-HWR was an international standard problem exercise entitled: Intercomparison and validation of computer codes for thermalhydraulics safety analyses. Intercomparison and validation of computer codes used in different countries for thermalhydraulics safety analyses will enhance the confidence in the predictions made by these codes. However, the intercomparison and validation exercise needs a set of reliable experimental data. The RD-14M Large-Loss Of Coolant Accident (LOCA) test B9401 simulating HWR LOCA behaviour that was conducted by Atomic Energy of Canada Ltd (AECL) was selected for this validation project. This report provides a comparison of the results obtained from six participating countries, utilizing four different computer codes. General conclusions are reached and recommendations made
Radiometry of UVB: Comparisons with results of Lowtran 7 and Premar computer codes
International Nuclear Information System (INIS)
A prototype UVMFR instrument (Ultra Violet MultiFilter Radiometer) with 4 different wavelength sensors in the UVB band has been thoroughly tested in Brasimone (44,11 deg N; 11.11 deg E) prior to tis installation for long term measurement campaigns of the UVB flux at the ground (890 m. local height) with separation of the direct and diffuse components of radiation. It has also been used for the first validation of a new computer code (PREMAR) developed by ENEA which solves the radiation transfer equation in the atmosphere by a Montecarlo approach. The new code considers a multilayer geometry, allows for the computation of albedo effects and exploits the rich library and potentialities of the LOWTRAN-7 U.S. computer code. With reference to the best available data, in days with an optimal meteorology to avoid significant cloud effects, an intercomparison of the instrument and code results has been performed at different times (varying solar zenital angles). A good agreement has been obtained between experiment and calculations as to the diffuse / total radiation ratio, and the deduced local albedo has been found to correspond rather well to theoretical estimates
Homogeneous zones definition in deterministic codes and effect on computed neutronic parameters
Energy Technology Data Exchange (ETDEWEB)
Varvayanni, M. [NCSR ' DEMOKRITOS' , P.O. Box 60228, 15310 Aghia Paraskevi, Attiki (Greece)], E-mail: melina@ipta.demokritos.gr; Savva, P.; Catsaros, N. [NCSR ' DEMOKRITOS' , P.O. Box 60228, 15310 Aghia Paraskevi, Attiki (Greece); Antonopoulos-Domis, M. [School of Electrical and Computer Engineering, Aristotle University of Thessaloniki, Thessaloniki (Greece)
2009-05-15
The design or modification and in general the analysis and control of nuclear reactors require complex calculations, which are carried out by numerical codes including neutronic and thermal-hydraulic components. Among the neutronic codes, the deterministic ones which solve the neutron transport/diffusion equation simulate the reactor core by dividing it into homogenized zones, i.e. volumes within which the macroscopic nuclear properties are considered uniform. These codes have been extensively used and tested for several decades and are shown to perform well when they analyze reactor cores containing regions with relatively homogeneous distributions of fuel, moderator and absorbing materials. In this work, the sensitivity of computed key neutronic parameters to the partitioning of the reactor core in homogenized zones is examined. Application is made for a configuration of the Greek Research Reactor (GRR-1) core, which is pool type, fueled by slab-type fuel elements. For the calculations, the neutronic code system consisting of XSDRNPM (cell-calculations) and CITATION (core analysis) is used with two different definitions of homogeneous zones for the special/control fuel assemblies. The effect on computations of neutron flux distribution, void-induced reactivity and total control rod worth is examined based on corresponding measurements. It is shown that with a more appropriate partition in homogeneous zones, the agreement of computed results with measurements can be remarkably improved concerning mainly the neutron flux, while the control rods worth is the less affected quantity.
Homogeneous zones definition in deterministic codes and effect on computed neutronic parameters
International Nuclear Information System (INIS)
The design or modification and in general the analysis and control of nuclear reactors require complex calculations, which are carried out by numerical codes including neutronic and thermal-hydraulic components. Among the neutronic codes, the deterministic ones which solve the neutron transport/diffusion equation simulate the reactor core by dividing it into homogenized zones, i.e. volumes within which the macroscopic nuclear properties are considered uniform. These codes have been extensively used and tested for several decades and are shown to perform well when they analyze reactor cores containing regions with relatively homogeneous distributions of fuel, moderator and absorbing materials. In this work, the sensitivity of computed key neutronic parameters to the partitioning of the reactor core in homogenized zones is examined. Application is made for a configuration of the Greek Research Reactor (GRR-1) core, which is pool type, fueled by slab-type fuel elements. For the calculations, the neutronic code system consisting of XSDRNPM (cell-calculations) and CITATION (core analysis) is used with two different definitions of homogeneous zones for the special/control fuel assemblies. The effect on computations of neutron flux distribution, void-induced reactivity and total control rod worth is examined based on corresponding measurements. It is shown that with a more appropriate partition in homogeneous zones, the agreement of computed results with measurements can be remarkably improved concerning mainly the neutron flux, while the control rods worth is the less affected quantity.
International Nuclear Information System (INIS)
A digital computer code TRANP was developed to simulate the steady-state and transient behavior of a pressurizer water reactor primary circuit. The development of this code was based on the combining of three codes already developed for the simulation of a PWR core, a pressurizer, a steam generator and a main coolant pump, representing the primary circuit components. (Author)
International Nuclear Information System (INIS)
The REBO computer code has been written for the automatic generation, with relatively simple input data, of cylindrical coordinates for the three dimensional finite element grid of thick walled nozzle cylindrical vessel junctions with curved transitions. The REBO is a FORTRAN IV code written for the IBM system 370. The main feature of the code are presented and a user's manual is given
Once-through CANDU reactor models for the ORIGEN2 computer code
International Nuclear Information System (INIS)
Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % 235U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given
RAP-4A Computer code for thermohydraulic calculation of liquid metal cooled fuel clusters
International Nuclear Information System (INIS)
RAP-4A is a programme for calculating the fuel clusters thermal-hydraulic parameters in a fast liquid metal-cooled reactor. The code gives the possibility to calculate steady state axial distribution temperature, enthalpy, pressure drop and mass velocity . A monodimensional mathematical model along the cluster allowing the study of the single and two phase flow is used by taking into account the mixing between adjacent subchannels. Physical and mathematical models, general features and an example are presented. RAP-4A code is written in FORTRAN-IV language on IBM 370/135 computer
International Nuclear Information System (INIS)
The ASSERT subchannel code has been developed specifically to model flow and phase distributions within CANDU fuel bundles. ASSERT uses a drift-flux model which permits the phases to have unequal velocities, and can thus model phase separation tendencies which may occur in horizontal flow. The basic principles of ASSERT are outlined, and computed results are compared against data from various experiments for validation purposes. The paper concludes with an example of the use of the code to predict critical heat flux in CANDU geometries
International Nuclear Information System (INIS)
The computer code HADOC (Hanford Acute Dose Calculations) is described and instructions for its use are presented. The code calculates external dose from air submersion and inhalation doses following acute radionuclide releases. Atmospheric dispersion is calculated using the Hanford model with options to determine maximum conditions. Building wake effects and terrain variation may also be considered. Doses are calculated using dose conversion factor supplied in a data library. Doses are reported for one and fifty year dose commitment periods for the maximum individual and the regional population (within 50 miles). The fractional contribution to dose by radionuclide and exposure mode are also printed if requested
The MELTSPREAD-1 computer code for the analysis of transient spreading in containments
International Nuclear Information System (INIS)
A one-dimensional, multicell, Eulerian finite difference computer code (MELTSPREAD-1) has been developed to provide an improved prediction of the gravity driven spreading and thermal interactions of molten corium flowing over a concrete or steel surface. In this paper, the modeling incorporated into the code is described and the spreading models are benchmarked against a simple ''dam break'' problem as well as water simulant spreading data obtained in a scaled apparatus of the Mk I containment. Results are also presented for a scoping calculation of the spreading behavior and shell thermal response in the full scale Mk I system following vessel meltthrough. 24 refs., 15 figs
Metropol: A computer code for the simulation of transport of contaminants with groundwater
International Nuclear Information System (INIS)
In this report a description is given of the computer code Metropol. This code simulates the three-dimensional flow of groundwater with varying density and the simultaneous transport of contaminants in low concentration and is based on the finite element method. The basic equations for groundwater flow and transport are described as well as the mathematical techniques used to solve these equations. Pre-processing facilities for mesh generation and post-processing facilities such as particle tracking are also discussed. This work was part of the Community Mirage project Second phase, research area Calculation tools
Aquelarre. A computer code for fast neutron cross sections from the statistical model
International Nuclear Information System (INIS)
A Fortran V computer code for Univac 1108/6 using the partial statistical (or compound nucleus) model is described. The code calculates fast neutron cross sections for the (n, n'), (n, p), (n, d) and (n, α reactions and the angular distributions and Legendre moments.for the (n, n) and (n, n') processes in heavy and intermediate spherical nuclei. A local Optical Model with spin-orbit interaction for each level is employed, allowing for the width fluctuation and Moldauer corrections, as well as the inclusion of discrete and continuous levels. (Author) 67 refs
Experimental assessment of computer codes used for safety analysis of integral reactors
Energy Technology Data Exchange (ETDEWEB)
Falkov, A.A.; Kuul, V.S.; Samoilov, O.B. [OKB Mechanical Engineering, Nizhny Novgorod (Russian Federation)
1995-09-01
Peculiarities of integral reactor thermohydraulics in accidents are associated with presence of noncondensable gas in built-in pressurizer, absence of pumped ECCS, use of guard vessel for LOCAs localisation and passive RHRS through in-reactor HX`s. These features defined the main trends in experimental investigations and verification efforts for computer codes applied. The paper reviews briefly the performed experimental investigation of thermohydraulics of AST-500, VPBER600-type integral reactors. The characteristic of UROVEN/MB-3 code for LOCAs analysis in integral reactors and results of its verification are given. The assessment of RELAP5/mod3 applicability for accident analysis in integral reactor is presented.
SAGAPO. A computer code for the thermo-fluiddynamic analysis of gas cooled fuel element bundles
International Nuclear Information System (INIS)
This paper is a guide for the users of the Fortran computer code SAGAPO, which has been developed by the author for the thermo-fluiddynamic analysis of gas cooled fuel element bundles. The physical models and the mathematical procedures used in SAGAPO have been already described by the author of this work in a previous paper. Thus this work contains only a description of the structure of the code, together with the other informations necessary to the users. A listing of SAGAPO is included in the appendix, together with an example of input preparation and parts of printed results. (orig.)
TRIGEN (TRIga oriGEN) procedure: an origen code version for compatible IBM computer
International Nuclear Information System (INIS)
In the frame of management and development activities of the TRIGA RC-1 reactor plant, the need to create a code library devoted to the reactor operation needs has been felt. In this context, as to the inventory of fission products, the Oak Ridge National Laboratories ORIGEN code has been adapted to a IBM personal MS-DOS computer. An interactive procedure, called TRIGEN combining TRIGA and ORIGEN, was so developed. This procedure, against the input demands on RC-1 working conditions, produces an output tabulate relative to the activities, measured in curies, of fission products for every nuclide. (orig.)
Using the ORIGEN-2 computer code for near core activation calculations
International Nuclear Information System (INIS)
The ORIGEN2 computer code is a useful tool for calculating radionuclide inventories resulting from irradiation of materials in a reactor. It is widely used to calculate activation products in irradiated metals that form the structural portion of fuel assemblies. The code is straightforward for materials within the active fuel region of a reactor core, which are subject to core average conditions. For materials outside the active core, ORIGEN2 cannot be used directly. However, ORIGEN2 can be used with the appropriate methodology to calculate the activation of materials in near core locations. This paper presents the background and a methodology for estimating radionuclide inventories in activated metals in near core locations
V.S.O.P. (99/05) computer code system
International Nuclear Information System (INIS)
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 (∼65000 Fortran statements). (orig.)
International Nuclear Information System (INIS)
The information on the status of the work on development of the system of the nuclear safety codes for fast liquid metal reactors is presented in paper. The purpose of the work is to create an instrument for NPP neutronic, thermohydraulic and strength justification including human and environment radiation safety. The main task that is to be solved by the system of codes developed is the analysis of the broad spectrum of phenomena taking place on the NPP (including reactor itself, NPP components, containment rooms, industrial site and surrounding area) and analysis of the impact of the regular and accidental releases on the environment. The code system is oriented on the ability of fully integrated modeling of the NPP behavior in the coupled definition accounting for the wide range of significant phenomena taking place on the NPP under normal and accident conditions. It is based on the models that meet the state-of-the-art knowledge level. The codes incorporate advanced numerical methods and modern programming technologies oriented on the high-performance computing systems. The information on the status of the work on verification of the separate codes of the system of codes is also presented. (author)
V.S.O.P. (99/05) computer code system
Energy Technology Data Exchange (ETDEWEB)
Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Scherer, W.
2005-11-01
V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to HTRs and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. V.S.O.P.(99 / 05) represents the further development of V.S.O.P. (99). Compared to its precursor, the code system has been improved in many details. Major improvements and extensions have been included concerning the neutron spectrum calculation, the 3-d neutron diffusion options, and the thermal hydraulic section with respect to 'multi-pass'-fuelled pebblebed cores. This latest code version was developed and tested under the WINDOWS-XP - operating system. The storage requirement for the executables and the basic libraries associated with the code amounts to about 15 MB. Another 5 MB are required - if desired - for storage of the source code ({approx}65000 Fortran statements). (orig.)
International Nuclear Information System (INIS)
EPICSHOW (Electron Photon Interactive Code - Show Data) is an interactive graphics code that allows users to view and interact with neutron, photon, electron and light charged particle data. Besides on screen graphics the code provides hard copy in the form of tabulated listings and Postscript output files. The code has been implemented on UNIX, IBM-PC, Power MAC, and even Laptop computers. It should be relatively easy to use it on almost any computer. All of the data included in this system is based on the Lawrence Livermore National Laboratory Databases and the neutron and photon data is used in the TART97 Monte Carlo transport code system. (author)
International Nuclear Information System (INIS)
In the fuel rods of the first DUELL experiment highly asymmetric fuel structures were found which had been caused by a steep transversal neutron flux gradient and eccentric pellet location. The TEXDIF-P computer code was developed to explain this phenomenon in quantitative terms. This computer code solves for an encapsulated fuel rod the equation of two-dimensional heat conduction using the finite differences method. Any distribution may be specified of the heat source density and of the gap between the fuel pellet and the cladding tube. By use of the modular structure the material relations are easily exchangeable. The TEXDIF-P code can be applied both to oxide and to carbide fuel rods. Coupling of the POUMEC subprogram of SATURN-1 allows the dynamic calculation of pore migration. Independent of this, the program includes an option for determination of the limit of the pore migration zone via a relation covering the minimum pore migration path according to Olander. TEXDIF-P has been used so far to verify the first start-up ramp experiment of DUELL. The agreement between the computation and the findings of post-examinations is quite satisfactory regarding the size and the location of the central void. Also the limit of the compacted zone is fairly well reproduced by the computation. The assumption on the size of the transversal neutron flux gradient has been essentially confirmed retroactively by transversal γ-scanning. (orig.)
High-performance computational fluid dynamics: a custom-code approach
Fannon, James; Loiseau, Jean-Christophe; Valluri, Prashant; Bethune, Iain; Náraigh, Lennon Ó.
2016-07-01
We introduce a modified and simplified version of the pre-existing fully parallelized three-dimensional Navier–Stokes flow solver known as TPLS. We demonstrate how the simplified version can be used as a pedagogical tool for the study of computational fluid dynamics (CFDs) and parallel computing. TPLS is at its heart a two-phase flow solver, and uses calls to a range of external libraries to accelerate its performance. However, in the present context we narrow the focus of the study to basic hydrodynamics and parallel computing techniques, and the code is therefore simplified and modified to simulate pressure-driven single-phase flow in a channel, using only relatively simple Fortran 90 code with MPI parallelization, but no calls to any other external libraries. The modified code is analysed in order to both validate its accuracy and investigate its scalability up to 1000 CPU cores. Simulations are performed for several benchmark cases in pressure-driven channel flow, including a turbulent simulation, wherein the turbulence is incorporated via the large-eddy simulation technique. The work may be of use to advanced undergraduate and graduate students as an introductory study in CFDs, while also providing insight for those interested in more general aspects of high-performance computing.
International Nuclear Information System (INIS)
The RALLY computer code pack (RALLY pack) is a set of computer codes destinate to the reliability of complex systems, aiming to a risk analysis. Three of the six codes, are commented, presenting their purpose, input description, calculation methods and results obtained with each one of those computer codes. The computer codes are: TREBIL, to obtain the fault tree logical equivalent; CRESSEX, to obtain the minimal cut and the punctual values of the non-reliability and non-availability of the system; and STREUSL, for the dispersion calculation of those values around the media. In spite of the CRESSEX, in its version available at CNEN, uses a little long method to obtain the minimal cut in an HB-CNEN system, the three computer programs show good results, mainly the STREUSL, which permits the simulation of various components. (E.G.)
Computer code for the thermal-hydraulic analysis of ITU TRIGA Mark-II reactor
International Nuclear Information System (INIS)
Istanbul Technical University (ITU) TRIGA Mark-II reactor core consists of ninety vertical cylindrical elements located in five rings. Sixty-nine of them are fuel elements. The reactor is operated and cooled with natural convection by pool water, which is also cooled and purified in external coolant circuits by forced convection. This characteristic leads to consider both the natural and forced convection heat transfer in a 'porous-medium analysis'. The safety analysis of the reactor requires a thermal-hydraulic model of the reactor to determine the thermal-hydraulic parameters in each mode of operation. In this study, a computer code cooled TRIGA-PM (TRIGA - Porous Medium) for the thermal-hydraulic analysis of ITU is considered. TRIGA Mark-II reactor code has been developed to obtain velocity, pressure and temperature distributions in the reactor pool as a function of core design parameters and pool configuration. The code is a transient, thermal-hydraulic code and requires geometric and physical modelling parameters. In the model, although the reactor is considered as only porous medium, the other part of the reactor pool is considered partly as continuum and partly as porous medium. COMMIX-1C code is used for the benchmark purpose of TRIGA-PM code. For the normal operating conditions of the reactor, estimations of TRIGA-PM are in good agreement with those of COMMIX-1C. After some more improvements, this code will be employed for the estimation of LOCA scenario, which can not be analyses by COMMIX-1C and the other multi-purpose codes, considering a break at one of the beam tubes of the reactor
The CAIN computer code for the generation of MABEL input data sets: a user's manual
International Nuclear Information System (INIS)
CAIN is an interactive FORTRAN computer code designed to overcome the substantial effort involved in manually creating the thermal-hydraulics input data required by MABEL-2. CAIN achieves this by processing output from either of the whole-core codes, RELAP or TRAC, interpolating where necessary, and by scanning RELAP/TRAC output in order to generate additional information. This user's manual describes the actions required in order to create RELAP/TRAC data sets from magnetic tape, to create the other input data sets required by CAIN, and to operate the interactive command procedure for the execution of CAIN. In addition, the CAIN code is described in detail. This programme of work is part of the Nuclear Installations Inspectorate (NII)'s contribution to the United Kingdom Atomic Energy Authority's independent safety assessment of pressurized water reactors. (author)
A computer code for calculating a γ-external dose from a randomly distributed radioactive cloud
International Nuclear Information System (INIS)
A computer code ( CIDE ) has been developed to calculate a γ-external dose from a randomly distributed radioactive cloud. Atmospheric dispersion of radioactive materials accidentally released from a nuclear reactor needs to be estimated considering time-dependent meteorological data and terrain heights. Particle-in-Cell model is useful for that purpose, but it is not easy to calculate the dose from the randomly distributed concentration by numerical integration. In this study the mean concentration in a cell evaluated by PIC model was assumed to be uniformly distributed over that cell, which was integrated as a constant concentration by a point kernel method. The dose was obtained by summing the attributable cell doses. When the concentration of plume had a Gaussian distribution, the results of CIDE code well agreed with those of GAMPLE, which was the code for calculating the dose from the Gaussian distribution. The choice of cell sizes affecting the accuracy of the calculated results was discussed. (author)
Enhancement of the Probabilistic CEramic Matrix Composite ANalyzer (PCEMCAN) Computer Code
Shah, Ashwin
2000-01-01
This report represents a final technical report for Order No. C-78019-J entitled "Enhancement of the Probabilistic Ceramic Matrix Composite Analyzer (PCEMCAN) Computer Code." The scope of the enhancement relates to including the probabilistic evaluation of the D-Matrix terms in MAT2 and MAT9 material properties card (available in CEMCAN code) for the MSC/NASTRAN. Technical activities performed during the time period of June 1, 1999 through September 3, 1999 have been summarized, and the final version of the enhanced PCEMCAN code and revisions to the User's Manual is delivered along with. Discussions related to the performed activities were made to the NASA Project Manager during the performance period. The enhanced capabilities have been demonstrated using sample problems.
Laser bar code applied in computer aided design of power fittings
Yang, Xiaohong; Yang, Fan
2010-10-01
A computer aided process planning system is developed based on laser bar code technology to automatize and standardize processing-paper making. The system sorts fittings by analyzing their types, structures, dimensions, materials, and technics characteristics, groups and encodes the fittings with similar technology characteristics base on the theory of Group Technology (GT). The system produces standard technology procedures using integrative-parts method and stores them into technics databases. To work out the technology procedure of fittings, the only thing for users need to do is to scan the bar code of fittings with a laser code reader. The system can produce process-paper using decision trees method and then print the process-cards automatically. The software has already been applied in some power stations and is praised by the users.
A computer simulation code of heat input due to incidence of fast ion beam
International Nuclear Information System (INIS)
A computer code has been developed to evaluate heat flux due to bombardment of ion beam to the beam limiters and beam dump of the neutral beam injector. In this code, energetic ions extracted from the ion source are represented by finite number of test particles, and their trajectories are calculated in the presense of magnetic field. They are bent and reflected as they pass through the bending magnet region. Finally they bombard the wall of the beam limiters and beam dump, where their energies are deposited. The heat flux can be derived from the number of bombarding test particles in a unit area. The code has been applied to the design of beam line hardwares of the JT-60 neutral beam injectors. (author)
International Nuclear Information System (INIS)
A computer code (MIGSTEM-FIT) has been developed to determine the prediction parameters, retardation factor, water flow velocity, dispersion coefficient, etc., of radionuclide migration in soil layer from the concentration distribution of radionuclide in soil layer or in effluent. In this code, the solution of the predicting equation for radionuclide migration is compared with the concentration distribution measured, and the most adequate values of parameter can be determined by the flexible tolerance method. The validity of finite differential method, which was one of the method to solve the predicting equation, was confirmed by comparison with the analytical solution, and also the validity of fitting method was confirmed by the fitting of the concentration distribution calculated from known parameters. From the examination about the error, it was found that the error of the parameter obtained by using this code was smaller than that of the concentration distribution measured. (author)
Large break LOCA analysis for retrofitted ECCS at MAPS using modified computer code ATMIKA
International Nuclear Information System (INIS)
Full text: Computer code ATMIKA which has been used for thermal hydraulic analysis is based on unequal velocity equal temperature (UVET) model. Thermal hydraulic transient was predicted using three conservation equations and drift flux model. The modified drift flux model is now able to predict counter current flow and the relative velocity in vertical channel more accurately. Apart from this, stratification model is also introduced to predict the fuel behaviour under stratified condition. Many more improvements were carried out with respect to solution of conservation equation, heat transfer package and frictional pressure drop model. All these modifications have been well validated with published data on RD-12/RD-14 experiments. This paper describes the code modifications and also deals with the application of the code for the large break LOCA analysis for retrofitted emergency core cooling system (ECCS) being implemented at Madras Atomic Power Station (MAPS). This paper also brings out the effect of accumulator on stratification and fuel behaviour
WOLF: a computer code package for the calculation of ion beam trajectories
International Nuclear Information System (INIS)
The WOLF code solves POISSON'S equation within a user-defined problem boundary of arbitrary shape. The code is compatible with ANSI FORTRAN and uses a two-dimensional Cartesian coordinate geometry represented on a triangular lattice. The vacuum electric fields and equipotential lines are calculated for the input problem. The use may then introduce a series of emitters from which particles of different charge-to-mass ratios and initial energies can originate. These non-relativistic particles will then be traced by WOLF through the user-defined region. Effects of ion and electron space charge are included in the calculation. A subprogram PISA forms part of this code and enables optimization of various aspects of the problem. The WOLF package also allows detailed graphics analysis of the computed results to be performed
Hydra-II, a computer code for hydrothermal analysis of spent fuel storage systems
Energy Technology Data Exchange (ETDEWEB)
McCann, R.A.
1988-03-01
HYDRA-II is a hydrothermal computer code designed to accurately predict steady-state fluid flow and temperature distributions in spent nuclear fuel storage and transportation systems. The code is capable of three-dimensional analysis of coupled conduction, convection, and thermal radiation problems. It provides a finite difference solution in cartesian coordinates to the equations governing the conservation of mass, momentum, and energy. A cylindrical coordinate system may also be used to enclose the cartesian coordinate system. This exterior coordinate system is useful for modeling cylindrical cask bodies. The basic equations and an overview of the numerics employed in their solution are presented. Selected results from an extensive code verification/validation effort are also presented. Comparisons are made between the results of simulations of a multiassembly storage system and actual experimental data. The effects of backfill gas composition and pressure and cask orientation are illustrated.
Hydra-II, a computer code for hydrothermal analysis of spent fuel storage systems
International Nuclear Information System (INIS)
HYDRA-II is a hydrothermal computer code designed to accurately predict steady-state fluid flow and temperature distributions in spent nuclear fuel storage and transportation systems. The code is capable of three-dimensional analysis of coupled conduction, convection, and thermal radiation problems. It provides a finite difference solution in cartesian coordinates to the equations governing the conservation of mass, momentum, and energy. A cylindrical coordinate system may also be used to enclose the cartesian coordinate system. This exterior coordinate system is useful for modeling cylindrical cask bodies. The basic equations and an overview of the numerics employed in their solution are presented. Selected results from an extensive code verification/validation effort are also presented. Comparisons are made between the results of simulations of a multiassembly storage system and actual experimental data. The effects of backfill gas composition and pressure and cask orientation are illustrated
Computer codes for the calculation of vibrations in machines and structures
International Nuclear Information System (INIS)
After an introductory paper on the typical requirements to be met by vibration calculations, the first two sections of the conference papers present universal as well as specific finite-element codes tailored to solve individual problems. The calculation of dynamic processes increasingly now in addition to the finite elements applies the method of multi-component systems which takes into account rigid bodies or partial structures and linking and joining elements. This method, too, is explained referring to universal computer codes and to special versions. In mechanical engineering, rotary vibrations are a major problem, and under this topic, conference papers exclusively deal with codes that also take into account special effects such as electromechanical coupling, non-linearities in clutches, etc. (orig./HP)
Validation of the transportation computer codes HIGHWAY, INTERLINE, RADTRAN 4, and RISKIND
International Nuclear Information System (INIS)
The computer codes HIGHWAY, INTERLINE, RADTRAN 4, and RISKIND were used to estimate radiation doses from the transportation of radioactive material in the Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Environmental Impact Statement. HIGHWAY and INTERLINE were used to estimate transportation routes for truck and rail shipments, respectively. RADTRAN 4 was used to estimate collective doses from incident-free transportation and the risk (probability x consequence) from transportation accidents. RISKIND was used to estimate incident-free radiation doses for maximally exposed individuals and the consequences from reasonably foreseeable transportation accidents. The purpose of this analysis is to validate the estimates made by these computer codes; critiques of the conceptual models used in RADTRAN 4 are also discussed. Validation is defined as ''the test and evaluation of the completed software to ensure compliance with software requirements.'' In this analysis, validation means that the differences between the estimates generated by these codes and independent observations are small (i.e., within the acceptance criterion established for the validation analysis). In some cases, the independent observations used in the validation were measurements; in other cases, the independent observations used in the validation analysis were generated using hand calculations. The results of the validation analyses performed for HIGHWAY, INTERLINE, RADTRAN 4, and RISKIND show that the differences between the estimates generated using the computer codes and independent observations were small. Based on the acceptance criterion established for the validation analyses, the codes yielded acceptable results; in all cases the estimates met the requirements for successful validation
Interactive computer codes for education and training on nuclear safety and radioprotection
International Nuclear Information System (INIS)
Two interactive computer codes for education and training on nuclear safety and radioprotection developed at RA6 Reactor Division-Bariloche Atomic Center-CNEA are presented on this paper. The first code named SIMREACT has been developed in order to simulate the control of a research nuclear reactor in real time with a simple but accurate approach. The code solves the equations of neutron punctual kinetics with time variable reactivity. Utilizing the timer of the computer and the controls of a PC keyboard, with an adequate graphic interface, a simulation in real time of the temporal behavior of a research reactor is obtained. The reactivity can be changed by means of the extraction or insertion of control rods. It was implemented also the simulation of automatic pilot and scram. The use of this code is focalized on practices of nuclear reactor control like start-up from the subcritical state with external source up to power to a desired level, change of power level, calibration of a control rod with different methods, and approach to critical condition by interpolation of the answer in function of reactivity. The second code named LICEN has been developed in order to help the studies of all the topics included in examination programs for obtaining licenses for research reactor operators and radioprotection officials. Using the PC mouse, with an adequate graphic interface, the student can gradually learn the topics related with general and special licenses. The general option includes nuclear reactor engineering, radioprotection, nuclear safety, documentation and normative. The specific option includes mandatory documentation, description of the installation and task on normal and emergency situations. For each of these topics there are sub-items with all the relevant information. The objective of this code is to joint in one electronic place a large amount of information which usually it is disseminated on difficult to find separated papers. (author)
HYDRA-II: A hydrothermal analysis computer code: Volume 3, Verification/validation assessments
International Nuclear Information System (INIS)
HYDRA-II is a hydrothermal computer code capable of three-dimensional analysis of coupled conduction, convection, and thermal radiation problems. This code is especially appropriate for simulating the steady-state performance of spent fuel storage systems. The code has been evaluated for this application for the US Department of Energy's Commercial Spent Fuel Management Program. HYDRA-II provides a finite difference solution in cartesian coordinates to the equations governing the conservation of mass, momentum, and energy. A cylindrical coordinate system may also be used to enclose the cartesian coordinate system. This exterior coordinate system is useful for modeling cylindrical cask bodies. The difference equations for conservation of momentum are enhanced by the incorporation of directional porosities and permeabilities that aid in modeling solid structures whose dimensions may be smaller than the computational mesh. The equation for conservation of energy permits modeling of orthotropic physical properties and film resistances. Several automated procedures are available to model radiation transfer within enclosures and from fuel rod to fuel rod. The documentation of HYDRA-II is presented in three separate volumes. Volume I - Equations and Numerics describes the basic differential equations, illustrates how the difference equations are formulated, and gives the solution procedures employed. Volume II - User's Manual contains code flow charts, discusses the code structure, provides detailed instructions for preparing an input file, and illustrates the operation of the code by means of a model problem. This volume, Volume III - Verification/Validation Assessments, provides a comparison between the analytical solution and the numerical simulation for problems with a known solution. This volume also documents comparisons between the results of simulations of single- and multiassembly storage systems and actual experimental data. 11 refs., 55 figs., 13 tabs
HYDRA-II: A hydrothermal analysis computer code: Volume 2, User's manual
International Nuclear Information System (INIS)
HYDRA-II is a hydrothermal computer code capable of three-dimensional analysis of coupled conduction, convection, and thermal radiation problems. This code is especially appropriate for simulating the steady-state performance of spent fuel storage systems. The code has been evaluated for this application for the US Department of Energy's Commercial Spent Fuel Management Program. HYDRA-II provides a finite-difference solution in cartesian coordinates to the equations governing the conservation of mass, momentum, and energy. A cylindrical coordinate system may also be used to enclose the cartesian coordinate system. This exterior coordinate system is useful for modeling cylindrical cask bodies. The difference equations for conservation of momentum incorporate directional porosities and permeabilities that are available to model solid structures whose dimensions may be smaller than the computational mesh. The equation for conservation of energy permits modeling of orthotropic physical properties and film resistances. Several automated methods are available to model radiation transfer within enclosures and from fuel rod to fuel rod. The documentation of HYDRA-II is presented in three separate volumes. Volume 1 - Equations and Numerics describes the basic differential equations, illustrates how the difference equations are formulated, and gives the solution procedures employed. This volume, Volume 2 - User's Manual, contains code flow charts, discusses the code structure, provides detailed instructions for preparing an input file, and illustrates the operation of the code by means of a sample problem. The final volume, Volume 3 - Verification/Validation Assessments, provides a comparison between the analytical solution and the numerical simulation for problems with a known solution. 6 refs
HYDRA-II: A hydrothermal analysis computer code: Volume 1, Equations and numerics
Energy Technology Data Exchange (ETDEWEB)
McCann, R.A.
1987-04-01
HYDRA-II is a hydrothermal computer code capable of three-dimensional analysis of coupled conduction, convection, and thermal radiation problems. This code is especially appropriate for simulating the steady-state performance of spent fuel storage systems. The code has been evaluated for this application for the US Department of Energy's Commercial Spent Fuel Management Program. HYDRA-II provides a finite difference solution in Cartesian coordinates to the equations governing the conservation of mass, momentum, and energy. A cylindrical coordinate system may also be used to enclose the Cartesian coordinate system. This exterior coordinate system is useful for modeling cylindrical cask bodies. The difference equations for conservation of momentum are enhanced by the incorporation of directional porosities and permeabilities that aid in modeling solid structures whose dimensions may be smaller than the computational mesh. The equation for conservation of energy permits of modeling of orthotropic physical properties and film resistances. Several automated procedures are available to model radiation transfer within enclosures and from fuel rod to fuel rod. The documentation of HYDRA-II is presented in three separate volumes. This volume, Volume I - Equations and Numerics, describes the basic differential equations, illustrates how the difference equations are formulated, and gives the solution procedures employed. Volume II - User's Manual contains code flow charts, discusses the code structure, provides detailed instructions for preparing an input file, and illustrates the operation of the code by means of a model problem. The final volume, Volume III - Verification/Validation Assessments, presents results of numerical simulations of single- and multiassembly storage systems and comparisons with experimental data. 4 refs.
HYDRA-II: A hydrothermal analysis computer code: Volume 3, Verification/validation assessments
Energy Technology Data Exchange (ETDEWEB)
McCann, R.A.; Lowery, P.S.
1987-10-01
HYDRA-II is a hydrothermal computer code capable of three-dimensional analysis of coupled conduction, convection, and thermal radiation problems. This code is especially appropriate for simulating the steady-state performance of spent fuel storage systems. The code has been evaluated for this application for the US Department of Energy's Commercial Spent Fuel Management Program. HYDRA-II provides a finite difference solution in cartesian coordinates to the equations governing the conservation of mass, momentum, and energy. A cylindrical coordinate system may also be used to enclose the cartesian coordinate system. This exterior coordinate system is useful for modeling cylindrical cask bodies. The difference equations for conservation of momentum are enhanced by the incorporation of directional porosities and permeabilities that aid in modeling solid structures whose dimensions may be smaller than the computational mesh. The equation for conservation of energy permits modeling of orthotropic physical properties and film resistances. Several automated procedures are available to model radiation transfer within enclosures and from fuel rod to fuel rod. The documentation of HYDRA-II is presented in three separate volumes. Volume I - Equations and Numerics describes the basic differential equations, illustrates how the difference equations are formulated, and gives the solution procedures employed. Volume II - User's Manual contains code flow charts, discusses the code structure, provides detailed instructions for preparing an input file, and illustrates the operation of the code by means of a model problem. This volume, Volume III - Verification/Validation Assessments, provides a comparison between the analytical solution and the numerical simulation for problems with a known solution. This volume also documents comparisons between the results of simulations of single- and multiassembly storage systems and actual experimental data. 11 refs., 55 figs., 13 tabs.
HYDRA-II: A hydrothermal analysis computer code: Volume 2, User's manual
Energy Technology Data Exchange (ETDEWEB)
McCann, R.A.; Lowery, P.S.; Lessor, D.L.
1987-09-01
HYDRA-II is a hydrothermal computer code capable of three-dimensional analysis of coupled conduction, convection, and thermal radiation problems. This code is especially appropriate for simulating the steady-state performance of spent fuel storage systems. The code has been evaluated for this application for the US Department of Energy's Commercial Spent Fuel Management Program. HYDRA-II provides a finite-difference solution in cartesian coordinates to the equations governing the conservation of mass, momentum, and energy. A cylindrical coordinate system may also be used to enclose the cartesian coordinate system. This exterior coordinate system is useful for modeling cylindrical cask bodies. The difference equations for conservation of momentum incorporate directional porosities and permeabilities that are available to model solid structures whose dimensions may be smaller than the computational mesh. The equation for conservation of energy permits modeling of orthotropic physical properties and film resistances. Several automated methods are available to model radiation transfer within enclosures and from fuel rod to fuel rod. The documentation of HYDRA-II is presented in three separate volumes. Volume 1 - Equations and Numerics describes the basic differential equations, illustrates how the difference equations are formulated, and gives the solution procedures employed. This volume, Volume 2 - User's Manual, contains code flow charts, discusses the code structure, provides detailed instructions for preparing an input file, and illustrates the operation of the code by means of a sample problem. The final volume, Volume 3 - Verification/Validation Assessments, provides a comparison between the analytical solution and the numerical simulation for problems with a known solution. 6 refs.
International Nuclear Information System (INIS)
Increasing use is being made of computer codes to predict the dynamic response of the containment of fast reactors following a hypothetical energy excursion. This paper reports on UK work to validate those aspects of the containment codes concerned with the prediction of the deformation of the primary vessel and other thin shell structures. The two containment codes ASTARTE and SEURBNUK are being developed in the UK and their general features and ability to reproduce experimental results have been reviewed elsewhere. Both codes have the basic capability for solving the time dependent flow of compressible fluids in two-dimensional axisymmetric geometry. ASTARTE uses a Langrangian finite difference formulation whereas SEURBNUK, under joint development by the UKAEA and JRC Ispra, uses an Eulerian approach. A structural analysis capability is also required and for thin tanks, eg the primary vessel, the deformation has a significant effect on the pressure field and must be computed simultaneously with the fluid motion. A thin shell model has therefore been incorporated into the codes to calculate the deformation of vessels either within the coolant or bounding it. The model assumes that the thickness of the shell is small compared with its characteristic length and that the normal stress through the shell can be neglected. Bending theory is incorporated by allowing the stresses to vary through the thickness. Material behaviour can be elasto-plastic; the von Mises yield criterion and Prandtl-Reuss flow rule are used in conjunction with the mechanical sub-layer model. Coupling of shell and fluid is achieved by different techniques in the two codes but in both it is implicit that there is free slip of fluid along the shell. (orig.)
Computation of Thermodynamic Equilibria Pertinent to Nuclear Materials in Multi-Physics Codes
Piro, Markus Hans Alexander
Nuclear energy plays a vital role in supporting electrical needs and fulfilling commitments to reduce greenhouse gas emissions. Research is a continuing necessity to improve the predictive capabilities of fuel behaviour in order to reduce costs and to meet increasingly stringent safety requirements by the regulator. Moreover, a renewed interest in nuclear energy has given rise to a "nuclear renaissance" and the necessity to design the next generation of reactors. In support of this goal, significant research efforts have been dedicated to the advancement of numerical modelling and computational tools in simulating various physical and chemical phenomena associated with nuclear fuel behaviour. This undertaking in effect is collecting the experience and observations of a past generation of nuclear engineers and scientists in a meaningful way for future design purposes. There is an increasing desire to integrate thermodynamic computations directly into multi-physics nuclear fuel performance and safety codes. A new equilibrium thermodynamic solver is being developed with this matter as a primary objective. This solver is intended to provide thermodynamic material properties and boundary conditions for continuum transport calculations. There are several concerns with the use of existing commercial thermodynamic codes: computational performance; limited capabilities in handling large multi-component systems of interest to the nuclear industry; convenient incorporation into other codes with quality assurance considerations; and, licensing entanglements associated with code distribution. The development of this software in this research is aimed at addressing all of these concerns. The approach taken in this work exploits fundamental principles of equilibrium thermodynamics to simplify the numerical optimization equations. In brief, the chemical potentials of all species and phases in the system are constrained by estimates of the chemical potentials of the system
Manual of a suite of computer codes, EXPRESS (EXact PREparedness Supporting System)
International Nuclear Information System (INIS)
The emergency response supporting system EXPRESS (EXact PREparedness Supporting System) is constructed in JAERI for low cost engineering work stations under the UNIX operation. The purpose of this system is real-time predictions of affected areas due to radioactivities discharged into atmosphere from nuclear facilities. The computational models in EXPRESS are the mass-consistent wind field model EXPRESS-I and the particle dispersion model EXPRESS-II for atmospheric dispersions. In order to attain the quick response even when the codes are used in a small-scale computer, a high-speed iteration method MILUCR (Modified Incomplete Linear Unitary Conjugate Residual) is applied to EXPRESS-I and kernel density method is to EXPRESS-II. This manual describes the model configurations, code structures, related files, namelists and sample outputs of EXPRESS-I and -II. (author)
Investigation of the efficiency and qualification of computer codes for PSA
International Nuclear Information System (INIS)
An international selection of computer codes for the quantification of level 1 PSA models was evaluated due to the consistence of results of different Benchmark exercises. The exercises in this project are based on those developed during the first benchmark project (Phase I). Due to several large differences in the results during Phase I of the Benchmark, the exercises in Phase II were more precisely defined. Due to the improved definition of the benchmark exercises, the results delivered from the different computer codes for Phase II are much more consistent. In general, the results of Benchmark II show, that the exercises were defined well enough to allow consistant results to be generated. Thus, the exercises can also be used to support the evaluation of additional PSA programs. (orig.)
Computing element evolution towards Exascale and its impact on legacy simulation codes
International Nuclear Information System (INIS)
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.)
COSA II Further benchmark exercises to compare geomechanical computer codes for salt
International Nuclear Information System (INIS)
Project COSA (COmputer COdes COmparison for SAlt) was a benchmarking exercise involving the numerical modelling of the geomechanical behaviour of heated rock salt. Its main objective was to assess the current European capability to predict the geomechanical behaviour of salt, in the context of the disposal of heat-producing radioactive waste in salt formations. Twelve organisations participated in the exercise in which their solutions to a number of benchmark problems were compared. The project was organised in two distinct phases: The first, from 1984-1986, concentrated on the verification of the computer codes. The second, from 1986-1988 progressed to validation, using three in-situ experiments at the Asse research facility in West Germany as a basis for comparison. This document reports the activities of the second phase of the project and presents the results, assessments and conclusions
Computing element evolution towards Exascale and its impact on legacy simulation codes
Colin de Verdière, Guillaume J. L.
2015-12-01
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.
Abstracts of digital computer code packages assembled by the Radiation Shielding Information Center
International Nuclear Information System (INIS)
This publication, ORNL/RSIC-13, Volumes I to III Revised, has resulted from an internal audit of the first 168 packages of computing technology in the Computer Codes Collection (CCC) of the Radiation Shielding Information Center (RSIC). It replaces the earlier three documents published as single volumes between 1966 to 1972. A significant number of the early code packages were considered to be obsolete and were removed from the collection in the audit process and the CCC numbers were not reassigned. Others not currently being used by the nuclear R and D community were retained in the collection to preserve technology not replaced by newer methods, or were considered of potential value for reference purposes. Much of the early technology, however, has improved through developer/RSIC/user interaction and continues at the forefront of the advancing state-of-the-art
Computing element evolution towards Exascale and its impact on legacy simulation codes
Energy Technology Data Exchange (ETDEWEB)
Colin de Verdiere, Guillaume J.L. [CEA, DAM, DIF, Arpajon (France)
2015-12-15
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.)
HIBRA: A computer code for heavy ion binary reaction analysis employing ion track detectors
Jamil, Khalid; Ahmad, Siraj-ul-Islam; Manzoor, Shahid
2016-01-01
Collisions of heavy ions many times result in production of only two reaction products. Study of heavy ions using ion track detectors allows experimentalists to observe the track length in the plane of the detector, depth of the tracks in the volume of the detector and angles between the tracks on the detector surface, all known as track parameters. How to convert these into useful physics parameters such as masses, energies, momenta of the reaction products and the Q-values of the reaction? This paper describes the (a) model used to analyze binary reactions in terms of measured etched track parameters of the reaction products recorded in ion track detectors, and (b) the code developed for computing useful physics parameters for fast and accurate analysis of a large number of binary events. A computer code, HIBRA (Heavy Ion Binary Reaction Analysis) has been developed both in C++ and FORTRAN programming languages. It has been tested on the binary reactions from 12.5 MeV/u 84Kr ions incident upon U (natural) target deposited on mica ion track detector. The HIBRA code can be employed with any ion track detector for which range-velocity relation is available including the widely used CR-39 ion track detectors. This paper provides the source code of HIBRA in C++ language along with input and output data to test the program.
Selection of Computer Codes for Shallow Land Waste Disposal in PPTA Serpong
International Nuclear Information System (INIS)
Selection of Computer Codes for Shallow Land Waste Disposal in PPTA Serpong. Models and computer codes have been selected for safety assessment of near surface waste disposal facility. This paper provides a summary and overview of the methodology and codes selected. The methodology allows analyses of dose to individuals from offsite releases under normal conditions as well as on-site doses to inadvertent intruders. A demonstration in the case of shallow land waste disposal in Nuclear Research Establishment are in Serpong has been given for normal release scenario. The assessment includes infiltration of rainfall, source-term, ground water (well) and surface water transport, food-chain and dosimetry. The results show dose history of maximally exposed individuals. The codes used are VS2DT, PAGAN and GENII. The application of 1 m silt loam as a moisture barrier cover decreases flow in the disposal unit by a factor of 27. The selected radionuclides show variety of dose histories according to their chemical and physical characteristics and behavior in the environment
Improvement of Level-1 PSA computer code package -A study for nuclear safety improvement-
International Nuclear Information System (INIS)
This year is the second year of the Government-sponsored Mid- and Long-Term Nuclear Power Technology Development Project. The scope of this subproject titled on 'The Improvement of Level-1 PSA Computer Codes' is divided into three main activities : (1) Methodology development on the under-developed fields such as risk assessment technology for plant shutdown and external events, (2) Computer code package development for Level-1 PSA, (3) Applications of new technologies to reactor safety assessment. At first, in the area of PSA methodology development, foreign PSA reports on shutdown and external events have been reviewed and various PSA methodologies have been compared. Level-1 PSA code KIRAP and CCF analysis code COCOA are converted from KOS to Windows. Human reliability database has been also established in this year. In the area of new technology applications, fuzzy set theory and entropy theory are used to estimate component life and to develop a new measure of uncertainty importance. Finally, in the field of application study of PSA technique to reactor regulation, a strategic study to develop a dynamic risk management tool PEPSI and the determination of inspection and test priority of motor operated valves based on risk importance worths have been studied. (Author)
Smoothing spline analysis of variance approach for global sensitivity analysis of computer codes
International Nuclear Information System (INIS)
The paper investigates a nonparametric regression method based on smoothing spline analysis of variance (ANOVA) approach to address the problem of global sensitivity analysis (GSA) of complex and computationally demanding computer codes. The two steps algorithm of this method involves an estimation procedure and a variable selection. The latter can become computationally demanding when dealing with high dimensional problems. Thus, we proposed a new algorithm based on Landweber iterations. Using the fact that the considered regression method is based on ANOVA decomposition, we introduced a new direct method for computing sensitivity indices. Numerical tests performed on several analytical examples and on an application from petroleum reservoir engineering showed that the method gives competitive results compared to a more standard Gaussian process approach
PyCUDA: GPU Run-Time Code Generation for High-Performance Computing
Klöckner, Andreas; Lee, Yunsup; Catanzaro, Bryan; Ivanov, Paul; Fasih, Ahmed
2009-01-01
High-performance scientific computing has recently seen a surge of interest in heterogeneous systems, with an emphasis on modern Graphics Processing Units (GPUs). These devices offer tremendous potential for performance and efficiency in important large-scale applications of computational science. However, exploiting this potential can be challenging, as one must adapt to the specialized and rapidly evolving computing environment currently exhibited by GPUs. One way of addressing this challenge is to embrace better techniques and develop tools tailored to their needs. This article presents one simple technique, GPU run-time code generation (RTCG), and PyCUDA, an open-source toolkit that supports this technique. In introducing PyCUDA, this article proposes the combination of a dynamic, high-level scripting language with the massive performance of a GPU as a compelling two-tiered computing platform, potentially offering significant performance and productivity advantages over conventional single-tier, static sy...
International Nuclear Information System (INIS)
A computer code WHAMS for calculating pressure and velocity transients in liquid filled piping networks is described. 11 different boundary conditions are applied of which two specific ones are described in some detail. Several example calculations are described and results are compared with those of other programs. WHAMS is capable of analyzing a network of 75 pipes which can be coupled in an arbitrary way. The program is written in Fortran IV language on UNIVAC 1108 computer and the program size is approximately 60 kwords. (author)
Implementation of an anisotropic turbulence model in the COMMIX- 1C/ATM computer code
International Nuclear Information System (INIS)
The computer code COMMIX-1C/ATM, which describes single-phase, three-dimensional transient thermofluiddynamic problems, has provided the framework for the extension of the standard k-var-epsilon turbulence model to a six-equation model with additional transport equations for the turbulence heat fluxes and the variance of temperature fluctuations. The new, model, which allows simulation of anisotropic turbulence in stratified shear flows, is referred to as the Anisotropic Turbulence Model (ATM) has been verified with numerical computations of stable and unstable stratified shear flow between parallel plates
ANCON: A code for the evaluation of complex fault trees in personal computers
International Nuclear Information System (INIS)
Performing probabilistic safety analysis has been recognized worldwide as one of the more effective ways for further enhancing safety of Nuclear Power Plants. The evaluation of fault trees plays a fundamental role in these analysis. Some existing limitations in RAM and execution speed of personal computers (PC) has restricted so far their use in the analysis of complex fault trees. Starting from new approaches in the data structure and other possibilities the ANCON code can evaluate complex fault trees in a PC, allowing the user to do a more comprehensive analysis of the considered system in reduced computing time
Radiation damage calculation by NPRIM computer code with JENDL3.3
International Nuclear Information System (INIS)
The Neutron Damage Evaluation Group of the Atomic Energy Society of Japan starts an identification of neutron-induced radiation damage in materials for typical neutron fields. For this study, a computer code, NPRIM, has been developed to be free from a tedious computational effort, which has been devoted to the calculation of derived quantities such as dpa and helium production rate. Neutron cross sections concerning to damage reactions based on JENDL3.3 are given with 640-group-structure. The impact of cross sections based on JENDL3.3 to damage calculation results has been described in this paper. (author)
International Nuclear Information System (INIS)
The computer codes were developed to evaluate internal radiation dose when radioactive isotopes released from nuclear facilities are taken through ingestion and inhalation pathways. Food chain models and relevant data base representing the agricultural and social environment of Korea are set up. An equilibrium model-KFOOD, which can deal with routine releases from a nuclear facility and a dynamic model-ECOREA, which is suitable for the description of acute radioactivity release following nuclear accident. (Author)
Tight bounds on computing error-correcting codes by bounded-depth circuits with arbitrary gates
DEFF Research Database (Denmark)
Gál, Anna; Hansen, Kristoffer Arnsfelt; Koucký, Michal; Pudlák, Pavel; Viola, Emanuele
We bound the minimum number w of wires needed to compute any (asymptotically good) error-correcting code C:{0,1}Ω(n) -> {0,1}n with minimum distance Ω(n), using unbounded fan-in circuits of depth d with arbitrary gates. Our main results are: (1) If d=2 then w = Θ(n ({log n/ log log n})2). (2) If d...
Tight bounds on computing error-correcting codes by bounded-depth circuits with arbitrary gates
DEFF Research Database (Denmark)
Gal, A.; Hansen, Kristoffer Arnsfelt; Koucky, Michal; Pudlak, P.; Viola, E.
2013-01-01
We bound the minimum number w of wires needed to compute any (asymptotically good) error-correcting code C:{0,1}Ω(n)→{0,1}n with minimum distance Ω(n), using unbounded fan-in circuits of depth d with arbitrary gates. Our main results are: 1) if d=2, then w=Θ(n (lgn/lglgn)2); 2) if d=3, then w...