High performance computer code for molecular dynamics simulations
International Nuclear Information System (INIS)
Levay, I.; Toekesi, K.
2007-01-01
Complete text of publication follows. Molecular Dynamics (MD) simulation is a widely used technique for modeling complicated physical phenomena. Since 2005 we are developing a MD simulations code for PC computers. The computer code is written in C++ object oriented programming language. The aim of our work is twofold: a) to develop a fast computer code for the study of random walk of guest atoms in Be crystal, b) 3 dimensional (3D) visualization of the particles motion. In this case we mimic the motion of the guest atoms in the crystal (diffusion-type motion), and the motion of atoms in the crystallattice (crystal deformation). Nowadays, it is common to use Graphics Devices in intensive computational problems. There are several ways to use this extreme processing performance, but never before was so easy to programming these devices as now. The CUDA (Compute Unified Device) Architecture introduced by nVidia Corporation in 2007 is a very useful for every processor hungry application. A Unified-architecture GPU include 96-128, or more stream processors, so the raw calculation performance is 576(!) GFLOPS. It is ten times faster, than the fastest dual Core CPU [Fig.1]. Our improved MD simulation software uses this new technology, which speed up our software and the code run 10 times faster in the critical calculation code segment. Although the GPU is a very powerful tool, it has a strongly paralleled structure. It means, that we have to create an algorithm, which works on several processors without deadlock. Our code currently uses 256 threads, shared and constant on-chip memory, instead of global memory, which is 100 times slower than others. It is possible to implement the total algorithm on GPU, therefore we do not need to download and upload the data in every iteration. On behalf of maximal throughput, every thread run with the same instructions
Computer simulation of variform fuel assemblies using Dragon code
International Nuclear Information System (INIS)
Ju Haitao; Wu Hongchun; Yao Dong
2005-01-01
The DRAGON is a cell code that developed for the CANDU reactor by the Ecole Polytechnique de Montreal of CANADA. Although, the DRAGON is mainly used to simulate the CANDU super-cell fuel assembly, it has an ability to simulate other geometries of the fuel assembly. However, only NEACRP benchmark problem of the BWR lattice cell was analyzed until now except for the CANDU reactor. We also need to develop the code to simulate the variform fuel assemblies, especially, for design of the advanced reactor. We validated that the cell code DRAGON is useful for simulating various kinds of the fuel assembly by analyzing the rod-shape fuel assembly of the PWR and the MTR plate-shape fuel assembly. Some other kinds of geometry of geometry were computed. Computational results show that the DRAGON is able to analyze variform fuel assembly problems and the precision is high. (authors)
Computed radiography simulation using the Monte Carlo code MCNPX
International Nuclear Information System (INIS)
Correa, S.C.A.; Souza, E.M.; Silva, A.X.; Lopes, R.T.
2009-01-01
Simulating x-ray images has been of great interest in recent years as it makes possible an analysis of how x-ray images are affected owing to relevant operating parameters. In this paper, a procedure for simulating computed radiographic images using the Monte Carlo code MCNPX is proposed. The sensitivity curve of the BaFBr image plate detector as well as the characteristic noise of a 16-bit computed radiography system were considered during the methodology's development. The results obtained confirm that the proposed procedure for simulating computed radiographic images is satisfactory, as it allows obtaining results comparable with experimental data. (author)
Computed radiography simulation using the Monte Carlo code MCNPX
Energy Technology Data Exchange (ETDEWEB)
Correa, S.C.A. [Programa de Engenharia Nuclear/COPPE, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil); Centro Universitario Estadual da Zona Oeste (CCMAT)/UEZO, Av. Manuel Caldeira de Alvarenga, 1203, Campo Grande, 23070-200, Rio de Janeiro, RJ (Brazil); Souza, E.M. [Programa de Engenharia Nuclear/COPPE, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil); Silva, A.X., E-mail: ademir@con.ufrj.b [PEN/COPPE-DNC/Poli CT, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil); Cassiano, D.H. [Instituto de Radioprotecao e Dosimetria/CNEN Av. Salvador Allende, s/n, Recreio, 22780-160, Rio de Janeiro, RJ (Brazil); Lopes, R.T. [Programa de Engenharia Nuclear/COPPE, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil)
2010-09-15
Simulating X-ray images has been of great interest in recent years as it makes possible an analysis of how X-ray images are affected owing to relevant operating parameters. In this paper, a procedure for simulating computed radiographic images using the Monte Carlo code MCNPX is proposed. The sensitivity curve of the BaFBr image plate detector as well as the characteristic noise of a 16-bit computed radiography system were considered during the methodology's development. The results obtained confirm that the proposed procedure for simulating computed radiographic images is satisfactory, as it allows obtaining results comparable with experimental data.
Computer codes for simulating atomic-displacement cascades in solids subject to irradiation
International Nuclear Information System (INIS)
Asaoka, Takumi; Taji, Yukichi; Tsutsui, Tsuneo; Nakagawa, Masayuki; Nishida, Takahiko
1979-03-01
In order to study atomic displacement cascades originating from primary knock-on atoms in solids subject to incident radiation, the simulation code CASCADE/CLUSTER is adapted for use on FACOM/230-75 computer system. In addition, the code is modified so as to plot the defect patterns in crystalline solids. As other simulation code of the cascade process, MARLOWE is also available for use on the FACOM system. To deal with the thermal annealing of point defects produced in the cascade process, the code DAIQUIRI developed originally for body-centered cubic crystals is modified to be applicable also for face-centered cubic lattices. By combining CASCADE/CLUSTER and DAIQUIRI, we then prepared a computer code system CASCSRB to deal with heavy irradiation or saturation damage state of solids at normal temperature. Furthermore, a code system for the simulation of heavy irradiations CASCMARL is available, in which MARLOWE code is substituted for CASCADE in the CASCSRB system. (author)
Validation of thermohydraulic codes by comparison of experimental results with computer simulations
International Nuclear Information System (INIS)
Madeira, A.A.; Galetti, M.R.S.; Pontedeiro, A.C.
1989-01-01
The results obtained by simulation of three cases from CANON depressurization experience, using the TRAC-PF1 computer code, version 7.6, implanted in the VAX-11/750 computer of Brazilian CNEN, are presented. The CANON experience was chosen as first standard problem in thermo-hydraulic to be discussed at ENFIR for comparing results from different computer codes with results obtained experimentally. The ability of TRAC-PF1 code to prevent the depressurization phase of a loss of primary collant accident in pressurized water reactors is evaluated. (M.C.K.) [pt
A computer code to simulate X-ray imaging techniques
International Nuclear Information System (INIS)
Duvauchelle, Philippe; Freud, Nicolas; Kaftandjian, Valerie; Babot, Daniel
2000-01-01
A computer code was developed to simulate the operation of radiographic, radioscopic or tomographic devices. The simulation is based on ray-tracing techniques and on the X-ray attenuation law. The use of computer-aided drawing (CAD) models enables simulations to be carried out with complex three-dimensional (3D) objects and the geometry of every component of the imaging chain, from the source to the detector, can be defined. Geometric unsharpness, for example, can be easily taken into account, even in complex configurations. Automatic translations or rotations of the object can be performed to simulate radioscopic or tomographic image acquisition. Simulations can be carried out with monochromatic or polychromatic beam spectra. This feature enables, for example, the beam hardening phenomenon to be dealt with or dual energy imaging techniques to be studied. The simulation principle is completely deterministic and consequently the computed images present no photon noise. Nevertheless, the variance of the signal associated with each pixel of the detector can be determined, which enables contrast-to-noise ratio (CNR) maps to be computed, in order to predict quantitatively the detectability of defects in the inspected object. The CNR is a relevant indicator for optimizing the experimental parameters. This paper provides several examples of simulated images that illustrate some of the rich possibilities offered by our software. Depending on the simulation type, the computation time order of magnitude can vary from 0.1 s (simple radiographic projection) up to several hours (3D tomography) on a PC, with a 400 MHz microprocessor. Our simulation tool proves to be useful in developing new specific applications, in choosing the most suitable components when designing a new testing chain, and in saving time by reducing the number of experimental tests
A computer code to simulate X-ray imaging techniques
Energy Technology Data Exchange (ETDEWEB)
Duvauchelle, Philippe E-mail: philippe.duvauchelle@insa-lyon.fr; Freud, Nicolas; Kaftandjian, Valerie; Babot, Daniel
2000-09-01
A computer code was developed to simulate the operation of radiographic, radioscopic or tomographic devices. The simulation is based on ray-tracing techniques and on the X-ray attenuation law. The use of computer-aided drawing (CAD) models enables simulations to be carried out with complex three-dimensional (3D) objects and the geometry of every component of the imaging chain, from the source to the detector, can be defined. Geometric unsharpness, for example, can be easily taken into account, even in complex configurations. Automatic translations or rotations of the object can be performed to simulate radioscopic or tomographic image acquisition. Simulations can be carried out with monochromatic or polychromatic beam spectra. This feature enables, for example, the beam hardening phenomenon to be dealt with or dual energy imaging techniques to be studied. The simulation principle is completely deterministic and consequently the computed images present no photon noise. Nevertheless, the variance of the signal associated with each pixel of the detector can be determined, which enables contrast-to-noise ratio (CNR) maps to be computed, in order to predict quantitatively the detectability of defects in the inspected object. The CNR is a relevant indicator for optimizing the experimental parameters. This paper provides several examples of simulated images that illustrate some of the rich possibilities offered by our software. Depending on the simulation type, the computation time order of magnitude can vary from 0.1 s (simple radiographic projection) up to several hours (3D tomography) on a PC, with a 400 MHz microprocessor. Our simulation tool proves to be useful in developing new specific applications, in choosing the most suitable components when designing a new testing chain, and in saving time by reducing the number of experimental tests.
A computer code package for electron transport Monte Carlo simulation
International Nuclear Information System (INIS)
Popescu, Lucretiu M.
1999-01-01
A computer code package was developed for solving various electron transport problems by Monte Carlo simulation. It is based on condensed history Monte Carlo algorithm. In order to get reliable results over wide ranges of electron energies and target atomic numbers, specific techniques of electron transport were implemented such as: Moliere multiscatter angular distributions, Blunck-Leisegang multiscatter energy distribution, sampling of electron-electron and Bremsstrahlung individual interactions. Path-length and lateral displacement corrections algorithms and the module for computing collision, radiative and total restricted stopping powers and ranges of electrons are also included. Comparisons of simulation results with experimental measurements are finally presented. (author)
Monocrystal sputtering by the computer simulation code ACOCT
International Nuclear Information System (INIS)
Yamamura, Yasunori; Takeuchi, Wataru.
1987-09-01
A new computer code ACOCT has been developed in order to simulate the atomic collisions in the crystalline target within the binary collision approximation. The present code is more convenient as compared with the MARLOWE code, and takes the higher-order simultaneous collisions into account. To cheke the validity of the ACOCT program, we have calculated sputtering yields for various ion-target combinations and compared with the MARLOWE results. It is found that the calculated yields by the ACOCT program are in good agreements with those by the MARLOWE code. The ejection patterns of sputtered atoms were also calculated for the major surfaces of fcc, bcc, diamond and hcp structures, and we have got reasonable agreements with experimental results. In order to know the effects of the simultaneous collision in the slowing down process the sputtering yields and the projected ranges are calculated, changeing the parameter of the criterion for the simultaneous collision, and the effect of the simultaneous collision is found to depend on the crystal orientation. (author)
Extremely Scalable Spiking Neuronal Network Simulation Code: From Laptops to Exascale Computers
Jordan, Jakob; Ippen, Tammo; Helias, Moritz; Kitayama, Itaru; Sato, Mitsuhisa; Igarashi, Jun; Diesmann, Markus; Kunkel, Susanne
2018-01-01
State-of-the-art software tools for neuronal network simulations scale to the largest computing systems available today and enable investigations of large-scale networks of up to 10 % of the human cortex at a resolution of individual neurons and synapses. Due to an upper limit on the number of incoming connections of a single neuron, network connectivity becomes extremely sparse at this scale. To manage computational costs, simulation software ultimately targeting the brain scale needs to fully exploit this sparsity. Here we present a two-tier connection infrastructure and a framework for directed communication among compute nodes accounting for the sparsity of brain-scale networks. We demonstrate the feasibility of this approach by implementing the technology in the NEST simulation code and we investigate its performance in different scaling scenarios of typical network simulations. Our results show that the new data structures and communication scheme prepare the simulation kernel for post-petascale high-performance computing facilities without sacrificing performance in smaller systems. PMID:29503613
Extremely Scalable Spiking Neuronal Network Simulation Code: From Laptops to Exascale Computers.
Jordan, Jakob; Ippen, Tammo; Helias, Moritz; Kitayama, Itaru; Sato, Mitsuhisa; Igarashi, Jun; Diesmann, Markus; Kunkel, Susanne
2018-01-01
State-of-the-art software tools for neuronal network simulations scale to the largest computing systems available today and enable investigations of large-scale networks of up to 10 % of the human cortex at a resolution of individual neurons and synapses. Due to an upper limit on the number of incoming connections of a single neuron, network connectivity becomes extremely sparse at this scale. To manage computational costs, simulation software ultimately targeting the brain scale needs to fully exploit this sparsity. Here we present a two-tier connection infrastructure and a framework for directed communication among compute nodes accounting for the sparsity of brain-scale networks. We demonstrate the feasibility of this approach by implementing the technology in the NEST simulation code and we investigate its performance in different scaling scenarios of typical network simulations. Our results show that the new data structures and communication scheme prepare the simulation kernel for post-petascale high-performance computing facilities without sacrificing performance in smaller systems.
Extremely Scalable Spiking Neuronal Network Simulation Code: From Laptops to Exascale Computers
Directory of Open Access Journals (Sweden)
Jakob Jordan
2018-02-01
Full Text Available State-of-the-art software tools for neuronal network simulations scale to the largest computing systems available today and enable investigations of large-scale networks of up to 10 % of the human cortex at a resolution of individual neurons and synapses. Due to an upper limit on the number of incoming connections of a single neuron, network connectivity becomes extremely sparse at this scale. To manage computational costs, simulation software ultimately targeting the brain scale needs to fully exploit this sparsity. Here we present a two-tier connection infrastructure and a framework for directed communication among compute nodes accounting for the sparsity of brain-scale networks. We demonstrate the feasibility of this approach by implementing the technology in the NEST simulation code and we investigate its performance in different scaling scenarios of typical network simulations. Our results show that the new data structures and communication scheme prepare the simulation kernel for post-petascale high-performance computing facilities without sacrificing performance in smaller systems.
International Nuclear Information System (INIS)
Zhang, Shuai; Morita, Koji; Shirakawa, Noriyuki; Yamamoto, Yuichi
2010-01-01
The COMPASS code is designed based on the moving particle semi-implicit method to simulate various complex mesoscale phenomena relevant to core disruptive accidents of sodium-cooled fast reactors. In this study, a computational framework for fluid-solid mixture flow simulations was developed for the COMPASS code. The passively moving solid model was used to simulate hydrodynamic interactions between fluid and solids. Mechanical interactions between solids were modeled by the distinct element method. A multi-time-step algorithm was introduced to couple these two calculations. The proposed computational framework for fluid-solid mixture flow simulations was verified by the comparison between experimental and numerical studies on the water-dam break with multiple solid rods. (author)
Dynamic benchmarking of simulation codes
International Nuclear Information System (INIS)
Henry, R.E.; Paik, C.Y.; Hauser, G.M.
1996-01-01
Computer simulation of nuclear power plant response can be a full-scope control room simulator, an engineering simulator to represent the general behavior of the plant under normal and abnormal conditions, or the modeling of the plant response to conditions that would eventually lead to core damage. In any of these, the underlying foundation for their use in analysing situations, training of vendor/utility personnel, etc. is how well they represent what has been known from industrial experience, large integral experiments and separate effects tests. Typically, simulation codes are benchmarked with some of these; the level of agreement necessary being dependent upon the ultimate use of the simulation tool. However, these analytical models are computer codes, and as a result, the capabilities are continually enhanced, errors are corrected, new situations are imposed on the code that are outside of the original design basis, etc. Consequently, there is a continual need to assure that the benchmarks with important transients are preserved as the computer code evolves. Retention of this benchmarking capability is essential to develop trust in the computer code. Given the evolving world of computer codes, how is this retention of benchmarking capabilities accomplished? For the MAAP4 codes this capability is accomplished through a 'dynamic benchmarking' feature embedded in the source code. In particular, a set of dynamic benchmarks are included in the source code and these are exercised every time the archive codes are upgraded and distributed to the MAAP users. Three different types of dynamic benchmarks are used: plant transients; large integral experiments; and separate effects tests. Each of these is performed in a different manner. The first is accomplished by developing a parameter file for the plant modeled and an input deck to describe the sequence; i.e. the entire MAAP4 code is exercised. The pertinent plant data is included in the source code and the computer
Development of a computer code for transients simulation in PWR type reactors
International Nuclear Information System (INIS)
Alvim, A.C.M.; Botelho, D.A.; Oliveira Barroso, A.C. de
1981-01-01
A computer code for the simulation of operacional-transients and accidents in PWR type reactors is being developed at IEN (Instituto de Engenharia Nuclear). Accidents will be considered in which variations in thermohydraulics parameters of fuel and coolant don't cause nucleate boiling in the reactor core, but, otherwise are sufficiently strong to justify a more detailed simulation than that used in linearized models. (E.G.) [pt
SIVAR - Computer code for simulation of fuel rod behavior in PWR during fast transients
International Nuclear Information System (INIS)
Dias, A.F.V.
1980-10-01
Fuel rod behavior during a stationary and a transitory operation, is studied. A computer code aiming at simulating PWR type rods, was developed; however, it can be adapted for simulating other type of rods. A finite difference method was used. (E.G.) [pt
Computer code for simulating pressurized water reactor core
International Nuclear Information System (INIS)
Serrano, A.M.B.
1978-01-01
A computer code was developed for the simulation of the steady-state and transient behaviour of the average channel of a Pressurizer Water Reactor core. Point kinetics equations were used with the reactivity calculated for average temperatures in the channel with the fuel and moderator temperature feedbacks. The radial heat conduction equation in the fuel was solved numerically. For calculating the thermodynamic properties of the coolant, the fundamental equations of conservation (mass, energy and momentum) were solved. The gap and clad were treated as a resistance added to the film coefficient. The fuel system equations were decoupled from the coolant equations. The program permitted the changes in the heat transfer correlations and the flow patterns along the coolant channel. Various test were performed to determine the steady-state and transient response employing the PWR core simulator developed, obtaining results with adequate precision. (author)
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.
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)
Yamanishi, Toshihiko; Okuno, Kenji
1996-09-01
A computer code has been developed to simulate a multistage CECE(Combined Electrolysis Chemical Exchange) column. The solution of basic equations can be found out by the Newton-Raphson method. The independent variables are the atom fractions of D and T in each stage for the case where H is dominant within the column. These variables are replaced by those of H and T under the condition that D is dominant. Some effective techniques have also been developed to get a set of solutions of the basic equations: a setting procedure of initial values of the independent variables; and a procedure for the convergence of the Newton-Raphson method. The computer code allows us to simulate the column behavior under a wide range of the operating conditions. Even for a severe case, where the dominant species changes along the column height, the code can give a set of solutions of the basic equations. (author)
Computer codes for simulation of Angra 1 reactor steam generator
International Nuclear Information System (INIS)
Pinto, A.C.
1978-01-01
A digital computer code is developed for the simulation of the steady-state operation of a u-tube steam generator with natural recirculation used in Pressurized Water Reactors. The steam generator is simulated with two flow channel separated by a metallic wall, with a preheating section with counter flow and a vaporizing section with parallel flow. The program permits the changes in flow patterns and heat transfer correlations, in accordance with the local conditions along the vaporizing section. Various sub-routines are developed for the determination of steam and water properties and a mathematical model is established for the simulation of transients in the same steam generator. The steady state operating conditions in one of the steam generators of ANGRA 1 reactor are determined utilizing this programme. Global results obtained agree with published values [pt
International Nuclear Information System (INIS)
Satake, Shinsuke; Okamoto, Masao; Nakajima, Noriyoshi; Takamaru, Hisanori
2005-11-01
A neoclassical transport simulation code (FORTEC-3D) applicable to three-dimensional configurations has been developed using High Performance Fortran (HPF). Adoption of computing techniques for parallelization and a hybrid simulation model to the δf Monte-Carlo method transport simulation, including non-local transport effects in three-dimensional configurations, makes it possible to simulate the dynamism of global, non-local transport phenomena with a self-consistent radial electric field within a reasonable computation time. In this paper, development of the transport code using HPF is reported. Optimization techniques in order to achieve both high vectorization and parallelization efficiency, adoption of a parallel random number generator, and also benchmark results, are shown. (author)
A molecular dynamics simulation code ISIS
International Nuclear Information System (INIS)
Kambayashi, Shaw
1992-06-01
Computer simulation based on the molecular dynamics (MD) method has become an important tool complementary to experiments and theoretical calculations in a wide range of scientific fields such as physics, chemistry, biology, and so on. In the MD method, the Newtonian equations-of-motion of classical particles are integrated numerically to reproduce a phase-space trajectory of the system. In the 1980's, several new techniques have been developed for simulation at constant-temperature and/or constant-pressure in convenient to compare result of computer simulation with experimental results. We first summarize the MD method for both microcanonical and canonical simulations. Then, we present and overview of a newly developed ISIS (Isokinetic Simulation of Soft-spheres) code and its performance on various computers including vector processors. The ISIS code has a capability to make a MD simulation under constant-temperature condition by using the isokinetic constraint method. The equations-of-motion is integrated by a very accurate fifth-order finite differential algorithm. The bookkeeping method is also utilized to reduce the computational time. Furthermore, the ISIS code is well adopted for vector processing: Speedup ratio ranged from 16 to 24 times is obtained on a VP2600/10 vector processor. (author)
International Nuclear Information System (INIS)
Oelerich, Jan Oliver; Duschek, Lennart; Belz, Jürgen; Beyer, Andreas; Baranovskii, Sergei D.; Volz, Kerstin
2017-01-01
Highlights: • We present STEMsalabim, a modern implementation of the multislice algorithm for simulation of STEM images. • Our package is highly parallelizable on high-performance computing clusters, combining shared and distributed memory architectures. • With STEMsalabim, computationally and memory expensive STEM image simulations can be carried out within reasonable time. - Abstract: We present a new multislice code for the computer simulation of scanning transmission electron microscope (STEM) images based on the frozen lattice approximation. Unlike existing software packages, the code is optimized to perform well on highly parallelized computing clusters, combining distributed and shared memory architectures. This enables efficient calculation of large lateral scanning areas of the specimen within the frozen lattice approximation and fine-grained sweeps of parameter space.
Energy Technology Data Exchange (ETDEWEB)
Oelerich, Jan Oliver, E-mail: jan.oliver.oelerich@physik.uni-marburg.de; Duschek, Lennart; Belz, Jürgen; Beyer, Andreas; Baranovskii, Sergei D.; Volz, Kerstin
2017-06-15
Highlights: • We present STEMsalabim, a modern implementation of the multislice algorithm for simulation of STEM images. • Our package is highly parallelizable on high-performance computing clusters, combining shared and distributed memory architectures. • With STEMsalabim, computationally and memory expensive STEM image simulations can be carried out within reasonable time. - Abstract: We present a new multislice code for the computer simulation of scanning transmission electron microscope (STEM) images based on the frozen lattice approximation. Unlike existing software packages, the code is optimized to perform well on highly parallelized computing clusters, combining distributed and shared memory architectures. This enables efficient calculation of large lateral scanning areas of the specimen within the frozen lattice approximation and fine-grained sweeps of parameter space.
Moon, Hongsik
What is the impact of multicore and associated advanced technologies on computational software for science? Most researchers and students have multicore laptops or desktops for their research and they need computing power to run computational software packages. Computing power was initially derived from Central Processing Unit (CPU) clock speed. That changed when increases in clock speed became constrained by power requirements. Chip manufacturers turned to multicore CPU architectures and associated technological advancements to create the CPUs for the future. Most software applications benefited by the increased computing power the same way that increases in clock speed helped applications run faster. However, for Computational ElectroMagnetics (CEM) software developers, this change was not an obvious benefit - it appeared to be a detriment. Developers were challenged to find a way to correctly utilize the advancements in hardware so that their codes could benefit. The solution was parallelization and this dissertation details the investigation to address these challenges. Prior to multicore CPUs, advanced computer technologies were compared with the performance using benchmark software and the metric was FLoting-point Operations Per Seconds (FLOPS) which indicates system performance for scientific applications that make heavy use of floating-point calculations. Is FLOPS an effective metric for parallelized CEM simulation tools on new multicore system? Parallel CEM software needs to be benchmarked not only by FLOPS but also by the performance of other parameters related to type and utilization of the hardware, such as CPU, Random Access Memory (RAM), hard disk, network, etc. The codes need to be optimized for more than just FLOPs and new parameters must be included in benchmarking. In this dissertation, the parallel CEM software named High Order Basis Based Integral Equation Solver (HOBBIES) is introduced. This code was developed to address the needs of the
LFSC - Linac Feedback Simulation Code
Energy Technology Data Exchange (ETDEWEB)
Ivanov, Valentin; /Fermilab
2008-05-01
The computer program LFSC (
International Nuclear Information System (INIS)
Camargo, C.T.M.
1981-09-01
The Almod 3 computer code was modified, aiming at the simulation of Angra I nuclear power plant behavior during some reactor start-up tests. The results obtained with the modified computer code (Almod 3W) are compared with those obtained with the Retran computer code. (E.G.) [pt
Monte Carlo codes and Monte Carlo simulator program
International Nuclear Information System (INIS)
Higuchi, Kenji; Asai, Kiyoshi; Suganuma, Masayuki.
1990-03-01
Four typical Monte Carlo codes KENO-IV, MORSE, MCNP and VIM have been vectorized on VP-100 at Computing Center, JAERI. The problems in vector processing of Monte Carlo codes on vector processors have become clear through the work. As the result, it is recognized that these are difficulties to obtain good performance in vector processing of Monte Carlo codes. A Monte Carlo computing machine, which processes the Monte Carlo codes with high performances is being developed at our Computing Center since 1987. The concept of Monte Carlo computing machine and its performance have been investigated and estimated by using a software simulator. In this report the problems in vectorization of Monte Carlo codes, Monte Carlo pipelines proposed to mitigate these difficulties and the results of the performance estimation of the Monte Carlo computing machine by the simulator are described. (author)
Porting plasma physics simulation codes to modern computing architectures using the libmrc framework
Germaschewski, Kai; Abbott, Stephen
2015-11-01
Available computing power has continued to grow exponentially even after single-core performance satured in the last decade. The increase has since been driven by more parallelism, both using more cores and having more parallelism in each core, e.g. in GPUs and Intel Xeon Phi. Adapting existing plasma physics codes is challenging, in particular as there is no single programming model that covers current and future architectures. We will introduce the open-source libmrc framework that has been used to modularize and port three plasma physics codes: The extended MHD code MRCv3 with implicit time integration and curvilinear grids; the OpenGGCM global magnetosphere model; and the particle-in-cell code PSC. libmrc consolidates basic functionality needed for simulations based on structured grids (I/O, load balancing, time integrators), and also introduces a parallel object model that makes it possible to maintain multiple implementations of computational kernels, on e.g. conventional processors and GPUs. It handles data layout conversions and enables us to port performance-critical parts of a code to a new architecture step-by-step, while the rest of the code can remain unchanged. We will show examples of the performance gains and some physics applications.
Computer codes in particle transport physics
International Nuclear Information System (INIS)
Pesic, M.
2004-01-01
Simulation of transport and interaction of various particles in complex media and wide energy range (from 1 MeV up to 1 TeV) is very complicated problem that requires valid model of a real process in nature and appropriate solving tool - computer code and data library. A brief overview of computer codes based on Monte Carlo techniques for simulation of transport and interaction of hadrons and ions in wide energy range in three dimensional (3D) geometry is shown. Firstly, a short attention is paid to underline the approach to the solution of the problem - process in nature - by selection of the appropriate 3D model and corresponding tools - computer codes and cross sections data libraries. Process of data collection and evaluation from experimental measurements and theoretical approach to establishing reliable libraries of evaluated cross sections data is Ion g, difficult and not straightforward activity. For this reason, world reference data centers and specialized ones are acknowledged, together with the currently available, state of art evaluated nuclear data libraries, as the ENDF/B-VI, JEF, JENDL, CENDL, BROND, etc. Codes for experimental and theoretical data evaluations (e.g., SAMMY and GNASH) together with the codes for data processing (e.g., NJOY, PREPRO and GRUCON) are briefly described. Examples of data evaluation and data processing to generate computer usable data libraries are shown. Among numerous and various computer codes developed in transport physics of particles, the most general ones are described only: MCNPX, FLUKA and SHIELD. A short overview of basic application of these codes, physical models implemented with their limitations, energy ranges of particles and types of interactions, is given. General information about the codes covers also programming language, operation system, calculation speed and the code availability. An example of increasing computation speed of running MCNPX code using a MPI cluster compared to the code sequential option
Metropol: A computer code for the simulation of transport of contaminants with groundwater
International Nuclear Information System (INIS)
Sauter, F.J.; Hassanizadeh, S.M.; Leijnse, A.; Glasbergen, P.; Slot, A.F.M.
1990-01-01
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
LFSC - Linac Feedback Simulation Code
International Nuclear Information System (INIS)
Ivanov, Valentin; Fermilab
2008-01-01
The computer program LFSC ( ) is a numerical tool for simulation beam based feedback in high performance linacs. The code LFSC is based on the earlier version developed by a collective of authors at SLAC (L.Hendrickson, R. McEwen, T. Himel, H. Shoaee, S. Shah, P. Emma, P. Schultz) during 1990-2005. That code was successively used in simulation of SLC, TESLA, CLIC and NLC projects. It can simulate as pulse-to-pulse feedback on timescale corresponding to 5-100 Hz, as slower feedbacks, operating in the 0.1-1 Hz range in the Main Linac and Beam Delivery System. The code LFSC is running under Matlab for MS Windows operating system. It contains about 30,000 lines of source code in more than 260 subroutines. The code uses the LIAR ('Linear Accelerator Research code') for particle tracking under ground motion and technical noise perturbations. It uses the Guinea Pig code to simulate the luminosity performance. A set of input files includes the lattice description (XSIF format), and plane text files with numerical parameters, wake fields, ground motion data etc. The Matlab environment provides a flexible system for graphical output
Computation of the bounce-average code
International Nuclear Information System (INIS)
Cutler, T.A.; Pearlstein, L.D.; Rensink, M.E.
1977-01-01
The bounce-average computer code simulates the two-dimensional velocity transport of ions in a mirror machine. The code evaluates and bounce-averages the collision operator and sources along the field line. A self-consistent equilibrium magnetic field is also computed using the long-thin approximation. Optionally included are terms that maintain μ, J invariance as the magnetic field changes in time. The assumptions and analysis that form the foundation of the bounce-average code are described. When references can be cited, the required results are merely stated and explained briefly. A listing of the code is appended
A general purpose code for Monte Carlo simulations
International Nuclear Information System (INIS)
Wilcke, W.W.; Rochester Univ., NY
1984-01-01
A general-purpose computer code MONTHY has been written to perform Monte Carlo simulations of physical systems. To achieve a high degree of flexibility the code is organized like a general purpose computer, operating on a vector describing the time dependent state of the system under simulation. The instruction set of the 'computer' is defined by the user and is therefore adaptable to the particular problem studied. The organization of MONTHY allows iterative and conditional execution of operations. (orig.)
International Nuclear Information System (INIS)
Baes, C.F. III; Sharp, R.D.; Sjoreen, A.L.; Hermann, O.W.
1984-11-01
TERRA is a computer code which calculates concentrations of radionuclides and ingrowing daughters in surface and root-zone soil, produce and feed, beef, and milk from a given deposition rate at any location in the conterminous United States. The code is fully integrated with seven other computer codes which together comprise a Computerized Radiological Risk Investigation System, CRRIS. Output from either the long range (> 100 km) atmospheric dispersion code RETADD-II or the short range (<80 km) atmospheric dispersion code ANEMOS, in the form of radionuclide air concentrations and ground deposition rates by downwind location, serves as input to TERRA. User-defined deposition rates and air concentrations may also be provided as input to TERRA through use of the PRIMUS computer code. The environmental concentrations of radionuclides predicted by TERRA serve as input to the ANDROS computer code which calculates population and individual intakes, exposures, doses, and risks. TERRA incorporates models to calculate uptake from soil and atmospheric deposition on four groups of produce for human consumption and four groups of livestock feeds. During the environmental transport simulation, intermediate calculations of interception fraction for leafy vegetables, produce directly exposed to atmospherically depositing material, pasture, hay, and silage are made based on location-specific estimates of standing crop biomass. Pasture productivity is estimated by a model which considers the number and types of cattle and sheep, pasture area, and annual production of other forages (hay and silage) at a given location. Calculations are made of the fraction of grain imported from outside the assessment area. TERRA output includes the above calculations and estimated radionuclide concentrations in plant produce, milk, and a beef composite by location
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.
Garneli, Varvara; Chorianopoulos, Konstantinos
2018-01-01
Various aspects of computational thinking (CT) could be supported by educational contexts such as simulations and video-games construction. In this field study, potential differences in student motivation and learning were empirically examined through students' code. For this purpose, we performed a teaching intervention that took place over five…
Progress of laser-plasma interaction simulations with the particle-in-cell code
International Nuclear Information System (INIS)
Sakagami, Hitoshi; Kishimoto, Yasuaki; Sentoku, Yasuhiko; Taguchi, Toshihiro
2005-01-01
As the laser-plasma interaction is a non-equilibrium, non-linear and relativistic phenomenon, we must introduce a microscopic method, namely, the relativistic electromagnetic PIC (Particle-In-Cell) simulation code. The PIC code requires a huge number of particles to validate simulation results, and its task is very computation-intensive. Thus simulation researches by the PIC code have been progressing along with advances in computer technology. Recently, parallel computers with tremendous computational power have become available, and thus we can perform three-dimensional PIC simulations for the laser-plasma interaction to investigate laser fusion. Some simulation results are shown with figures. We discuss a recent trend of large-scale PIC simulations that enable direct comparison between experimental facts and computational results. We also discharge/lightning simulations by the extended PIC code, which include various atomic and relaxation processes. (author)
International Nuclear Information System (INIS)
Chalhoub, E.S.
1980-09-01
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) [pt
Advanced thermohydraulic simulation code for transients in LMFBRs (SSC-L code)
Energy Technology Data Exchange (ETDEWEB)
Agrawal, A.K.
1978-02-01
Physical models for various processes that are encountered in preaccident and transient simulation of thermohydraulic transients in the entire liquid metal fast breeder reactor (LMFBR) plant are described in this report. A computer code, SSC-L, was written as a part of the Super System Code (SSC) development project for the ''loop''-type designs of LMFBRs. This code has the self-starting capability, i.e., preaccident or steady-state calculations are performed internally. These results then serve as the starting point for the transient simulation.
Advanced thermohydraulic simulation code for transients in LMFBRs (SSC-L code)
International Nuclear Information System (INIS)
Agrawal, A.K.
1978-02-01
Physical models for various processes that are encountered in preaccident and transient simulation of thermohydraulic transients in the entire liquid metal fast breeder reactor (LMFBR) plant are described in this report. A computer code, SSC-L, was written as a part of the Super System Code (SSC) development project for the ''loop''-type designs of LMFBRs. This code has the self-starting capability, i.e., preaccident or steady-state calculations are performed internally. These results then serve as the starting point for the transient simulation
Computer code for the atomistic simulation of lattice defects and dynamics
International Nuclear Information System (INIS)
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
Wavelet subband coding of computer simulation output using the A++ array class library
Energy Technology Data Exchange (ETDEWEB)
Bradley, J.N.; Brislawn, C.M.; Quinlan, D.J.; Zhang, H.D. [Los Alamos National Lab., NM (United States); Nuri, V. [Washington State Univ., Pullman, WA (United States). School of EECS
1995-07-01
The goal of the project is to produce utility software for off-line compression of existing data and library code that can be called from a simulation program for on-line compression of data dumps as the simulation proceeds. Naturally, we would like the amount of CPU time required by the compression algorithm to be small in comparison to the requirements of typical simulation codes. We also want the algorithm to accomodate a wide variety of smooth, multidimensional data types. For these reasons, the subband vector quantization (VQ) approach employed in has been replaced by a scalar quantization (SQ) strategy using a bank of almost-uniform scalar subband quantizers in a scheme similar to that used in the FBI fingerprint image compression standard. This eliminates the considerable computational burdens of training VQ codebooks for each new type of data and performing nearest-vector searches to encode the data. The comparison of subband VQ and SQ algorithms in indicated that, in practice, there is relatively little additional gain from using vector as opposed to scalar quantization on DWT subbands, even when the source imagery is from a very homogeneous population, and our subjective experience with synthetic computer-generated data supports this stance. It appears that a careful study is needed of the tradeoffs involved in selecting scalar vs. vector subband quantization, but such an analysis is beyond the scope of this paper. Our present work is focused on the problem of generating wavelet transform/scalar quantization (WSQ) implementations that can be ported easily between different hardware environments. This is an extremely important consideration given the great profusion of different high-performance computing architectures available, the high cost associated with learning how to map algorithms effectively onto a new architecture, and the rapid rate of evolution in the world of high-performance computing.
The TESS [Tandem Experiment Simulation Studies] computer code user's manual
International Nuclear Information System (INIS)
Procassini, R.J.
1990-01-01
TESS (Tandem Experiment Simulation Studies) is a one-dimensional, bounded particle-in-cell (PIC) simulation code designed to investigate the confinement and transport of plasma in a magnetic mirror device, including tandem mirror configurations. Mirror plasmas may be modeled in a system which includes an applied magnetic field and/or a self-consistent or applied electrostatic potential. The PIC code TESS is similar to the PIC code DIPSI (Direct Implicit Plasma Surface Interactions) which is designed to study plasma transport to and interaction with a solid surface. The codes TESS and DIPSI are direct descendants of the PIC code ES1 that was created by A. B. Langdon. This document provides the user with a brief description of the methods used in the code and a tutorial on the use of the code. 10 refs., 2 tabs
Parallelization of quantum molecular dynamics simulation code
International Nuclear Information System (INIS)
Kato, Kaori; Kunugi, Tomoaki; Shibahara, Masahiko; Kotake, Susumu
1998-02-01
A quantum molecular dynamics simulation code has been developed for the analysis of the thermalization of photon energies in the molecule or materials in Kansai Research Establishment. The simulation code is parallelized for both Scalar massively parallel computer (Intel Paragon XP/S75) and Vector parallel computer (Fujitsu VPP300/12). Scalable speed-up has been obtained with a distribution to processor units by division of particle group in both parallel computers. As a result of distribution to processor units not only by particle group but also by the particles calculation that is constructed with fine calculations, highly parallelization performance is achieved in Intel Paragon XP/S75. (author)
Computer and compiler effects on code results: status report
International Nuclear Information System (INIS)
1996-01-01
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-assisted Particle-in-Cell code development
International Nuclear Information System (INIS)
Kawata, S.; Boonmee, C.; Teramoto, T.; Drska, L.; Limpouch, J.; Liska, R.; Sinor, M.
1997-12-01
This report presents a new approach for an electromagnetic Particle-in-Cell (PIC) code development by a computer: in general PIC codes have a common structure, and consist of a particle pusher, a field solver, charge and current density collections, and a field interpolation. Because of the common feature, the main part of the PIC code can be mechanically developed on a computer. In this report we use the packages FIDE and GENTRAN of the REDUCE computer algebra system for discretizations of field equations and a particle equation, and for an automatic generation of Fortran codes. The approach proposed is successfully applied to the development of 1.5-dimensional PIC code. By using the generated PIC code the Weibel instability in a plasma is simulated. The obtained growth rate agrees well with the theoretical value. (author)
Evaluation of the SCANAIR Computer Code
International Nuclear Information System (INIS)
Jernkvist, Lars Olof; Massih, Ali
2001-11-01
The SCANAIR computer code, version 3.2, has been evaluated from the standpoint of its capability to analyze, simulate and predict nuclear fuel behavior during severe power transients. SCANAIR calculates the thermal and mechanical behavior of a pressurized water reactor (PWR) fuel rod during a postulated reactivity initiated accident (RIA), and our evaluation indicates that SCANAIR is a state of the art computational tool for this purpose. Our evaluation starts by reviewing the basic theoretical models in SCANAIR, namely the governing equations for heat transfer, the mechanical response of fuel and clad, and the fission gas release behavior. The numerical methods used to solve the governing equations are briefly reviewed, and the range of applicability of the models and their limitations are discussed and illustrated with examples. Next, the main features of the SCANAIR user interface are delineated. The code requires an extensive amount of input data, in order to define burnup-dependent initial conditions to the simulated RIA. These data must be provided in a special format by a thermal-mechanical fuel rod analysis code. The user also has to supply the transient power history under RIA as input, which requires a code for neutronics calculation. The programming structure and documentation of the code are also addressed in our evaluation. SCANAIR is programmed in Fortran-77, and makes use of several general Fortran-77 libraries for handling input/output, data storage and graphical presentation of computed results. The documentation of SCANAIR and its helping libraries is generally of good quality. A drawback with SCANAIR in its present form, is that the code and its pre- and post-processors are tied to computers running the Unix or Linux operating systems. As part of our evaluation, we have performed a large number of computations with SCANAIR, some of which are documented in this report. The computations presented here include a hypothetical RIA in a high
TRANPZ - A computer code for the simulation of reactors with axial dependence
International Nuclear Information System (INIS)
Sampaio, L.C.M.
1980-12-01
A computer code was developed to simulate a PWR reactor in steady state and during transients. The solution of one speed diffusion equation in the axial direction is obtained numerically dividing the core in various axial segments and the axial power distribution is obtained there from. A method was developed to determine the transient solution. The external reactivity effects are caused by the motion of the control rods, starting from the steady condition with the control rods in any position. The heat conduction equation in the fuel is numerically solved in the radial direction. Various tests were performed in steady state and transient conditions and the validity of the present model was verified. Results were compared in steady state condition with the code CITATION and a reasonable agreement was found. (E.G.) [pt
User's manual of Tokamak Simulation Code
International Nuclear Information System (INIS)
Nakamura, Yukiharu; Nishino, Tooru; Tsunematsu, Toshihide; Sugihara, Masayoshi.
1992-12-01
User's manual for use of Tokamak Simulation Code (TSC), which simulates the time-evolutional process of deformable motion of axisymmetric toroidal plasma, is summarized. For the use at JAERI computer system, the TSC is linked with the data management system GAEA. This manual is forcused on the procedure for the input and output by using the GAEA system. Model equations to give axisymmetric motion, outline of code system, optimal method to get the well converged solution are also described. (author)
Tokamak simulation code manual
International Nuclear Information System (INIS)
Chung, Moon Kyoo; Oh, Byung Hoon; Hong, Bong Keun; Lee, Kwang Won
1995-01-01
The method to use TSC (Tokamak Simulation Code) developed by Princeton plasma physics laboratory is illustrated. In KT-2 tokamak, time dependent simulation of axisymmetric toroidal plasma and vertical stability have to be taken into account in design phase using TSC. In this report physical modelling of TSC are described and examples of application in JAERI and SERI are illustrated, which will be useful when TSC is installed KAERI computer system. (Author) 15 refs., 6 figs., 3 tabs
van Heerwaarden, Chiel C.; van Stratum, Bart J. H.; Heus, Thijs; Gibbs, Jeremy A.; Fedorovich, Evgeni; Mellado, Juan Pedro
2017-08-01
This paper describes MicroHH 1.0, a new and open-source (www.microhh.org) computational fluid dynamics code for the simulation of turbulent flows in the atmosphere. It is primarily made for direct numerical simulation but also supports large-eddy simulation (LES). The paper covers the description of the governing equations, their numerical implementation, and the parameterizations included in the code. Furthermore, the paper presents the validation of the dynamical core in the form of convergence and conservation tests, and comparison of simulations of channel flows and slope flows against well-established test cases. The full numerical model, including the associated parameterizations for LES, has been tested for a set of cases under stable and unstable conditions, under the Boussinesq and anelastic approximations, and with dry and moist convection under stationary and time-varying boundary conditions. The paper presents performance tests showing good scaling from 256 to 32 768 processes. The graphical processing unit (GPU)-enabled version of the code can reach a speedup of more than an order of magnitude for simulations that fit in the memory of a single GPU.
The GC computer code for flow sheet simulation of pyrochemical processing of spent nuclear fuels
International Nuclear Information System (INIS)
Ahluwalia, R.K.; Geyer, H.K.
1996-01-01
The GC computer code has been developed for flow sheet simulation of pyrochemical processing of spent nuclear fuel. It utilizes a robust algorithm SLG for analyzing simultaneous chemical reactions between species distributed across many phases. Models have been developed for analysis of the oxide fuel reduction process, salt recovery by electrochemical decomposition of lithium oxide, uranium separation from the reduced fuel by electrorefining, and extraction of fission products into liquid cadmium. The versatility of GC is demonstrated by applying the code to a flow sheet of current interest
Computer Modeling and Simulation
Energy Technology Data Exchange (ETDEWEB)
Pronskikh, V. S. [Fermilab
2014-05-09
Verification and validation of computer codes and models used in simulation are two aspects of the scientific practice of high importance and have recently been discussed by philosophers of science. While verification is predominantly associated with the correctness of the way a model is represented by a computer code or algorithm, validation more often refers to model’s relation to the real world and its intended use. It has been argued that because complex simulations are generally not transparent to a practitioner, the Duhem problem can arise for verification and validation due to their entanglement; such an entanglement makes it impossible to distinguish whether a coding error or model’s general inadequacy to its target should be blamed in the case of the model failure. I argue that in order to disentangle verification and validation, a clear distinction between computer modeling (construction of mathematical computer models of elementary processes) and simulation (construction of models of composite objects and processes by means of numerical experimenting with them) needs to be made. Holding on to that distinction, I propose to relate verification (based on theoretical strategies such as inferences) to modeling and validation, which shares the common epistemology with experimentation, to simulation. To explain reasons of their intermittent entanglement I propose a weberian ideal-typical model of modeling and simulation as roles in practice. I suggest an approach to alleviate the Duhem problem for verification and validation generally applicable in practice and based on differences in epistemic strategies and scopes
General purpose code for Monte Carlo simulations
International Nuclear Information System (INIS)
Wilcke, W.W.
1983-01-01
A general-purpose computer called MONTHY has been written to perform Monte Carlo simulations of physical systems. To achieve a high degree of flexibility the code is organized like a general purpose computer, operating on a vector describing the time dependent state of the system under simulation. The instruction set of the computer is defined by the user and is therefore adaptable to the particular problem studied. The organization of MONTHY allows iterative and conditional execution of operations
International Nuclear Information System (INIS)
Hall, D.G.; Watkins, J.C.
1987-01-01
This report documents an evaluation of the TRAC-PF1/MOD1 reactor safety analysis computer code during computer simulations of feedwater line break transients. The experimental data base for the evaluation included the results of three bottom feedwater line break tests performed in the Semiscale Mod-2C test facility. The tests modeled 14.3% (S-FS-7), 50% (S-FS-11), and 100% (S-FS-6B) breaks. The test facility and the TRAC-PF1/MOD1 model used in the calculations are described. Evaluations of the accuracy of the calculations are presented in the form of comparisons of measured and calculated histories of selected parameters associated with the primary and secondary systems. In addition to evaluating the accuracy of the code calculations, the computational performance of the code during the simulations was assessed. A conclusion was reached that the code is capable of making feedwater line break transient calculations efficiently, but there is room for significant improvements in the simulations that were performed. Recommendations are made for follow-on investigations to determine how to improve future feedwater line break calculations and for code improvements to make the code easier to use
High-performance computational fluid dynamics: a custom-code approach
International Nuclear Information System (INIS)
Fannon, James; Náraigh, Lennon Ó; Loiseau, Jean-Christophe; Valluri, Prashant; Bethune, Iain
2016-01-01
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. (paper)
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.
Use of the buble rise model in the simulation of the pressurizer in ALMOD 3 computer code
International Nuclear Information System (INIS)
Madeira, A.A.; Camargo, C.T.M.
1984-01-01
The implementation of the buble rise model in the ALMOD 3 computer code to simulate the pressurizer of a nuclear power plant is presented. Some transients for Angra I were calculated and results obtained from the original and modified models were compared. (E.G.) [pt
FELIX - a computer code for simulation of criticality excursions in liquid fissile solutions
International Nuclear Information System (INIS)
Gmal, B.; Weber, J.
1989-01-01
Knowledge of characteristic parameters like evolved power fission yield during an accidental excursion is of essential importance to estimate possible radiological consequences and resulting safety hazards. The computer code 'FELIX' simulates excursion characteristics of aqueous critical assemblies: Starting out from given initial conditions the space-dependent neutron kinetic equations are solved in one-dimensional geometry. Power, fission yield, reactivity and temperature are calculated as a function of time. Reactivity-feedback includes density effects and radiolytic gas voids. Results from calculations are compared with CRAC-experiments. (orig.)
Development of a space radiation Monte Carlo computer simulation based on the FLUKA and ROOT codes
Pinsky, L; Ferrari, A; Sala, P; Carminati, F; Brun, R
2001-01-01
This NASA funded project is proceeding to develop a Monte Carlo-based computer simulation of the radiation environment in space. With actual funding only initially in place at the end of May 2000, the study is still in the early stage of development. The general tasks have been identified and personnel have been selected. The code to be assembled will be based upon two major existing software packages. The radiation transport simulation will be accomplished by updating the FLUKA Monte Carlo program, and the user interface will employ the ROOT software being developed at CERN. The end-product will be a Monte Carlo-based code which will complement the existing analytic codes such as BRYNTRN/HZETRN presently used by NASA to evaluate the effects of radiation shielding in space. The planned code will possess the ability to evaluate the radiation environment for spacecraft and habitats in Earth orbit, in interplanetary space, on the lunar surface, or on a planetary surface such as Mars. Furthermore, it will be usef...
V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009
Energy Technology Data Exchange (ETDEWEB)
Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.
2010-07-15
V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)
V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009
International Nuclear Information System (INIS)
Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.
2010-07-01
V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)
FLASH: A finite element computer code for variably saturated flow
International Nuclear Information System (INIS)
Baca, R.G.; Magnuson, S.O.
1992-05-01
A numerical model was developed for use in performance assessment studies at the INEL. The numerical model, referred to as the FLASH computer code, is designed to simulate two-dimensional fluid flow in fractured-porous media. The code is specifically designed to model variably saturated flow in an arid site vadose zone and saturated flow in an unconfined aquifer. In addition, the code also has the capability to simulate heat conduction in the vadose zone. This report presents the following: description of the conceptual frame-work and mathematical theory; derivations of the finite element techniques and algorithms; computational examples that illustrate the capability of the code; and input instructions for the general use of the code. The FLASH computer code is aimed at providing environmental scientists at the INEL with a predictive tool for the subsurface water pathway. This numerical model is expected to be widely used in performance assessments for: (1) the Remedial Investigation/Feasibility Study process and (2) compliance studies required by the US Department of Energy Order 5820.2A
Computer Simulation of Reading.
Leton, Donald A.
In recent years, coding and decoding have been claimed to be the processes for converting one language form to another. But there has been little effort to locate these processes in the human learner or to identify the nature of the internal codes. Computer simulation of reading is useful because the similarities in the human reception and…
International Nuclear Information System (INIS)
Lysenko, W.P.
1984-04-01
We have developed the RFQLIB simulation system to provide a means to systematically generate the new versions of radio-frequency quadrupole (RFQ) linac simulation codes that are required by the constantly changing needs of a research environment. This integrated system simplifies keeping track of the various versions of the simulation code and makes it practical to maintain complete and up-to-date documentation. In this scheme, there is a certain standard version of the simulation code that forms a library upon which new versions are built. To generate a new version of the simulation code, the routines to be modified or added are appended to a standard command file, which contains the commands to compile the new routines and link them to the routines in the library. The library itself is rarely changed. Whenever the library is modified, however, this modification is seen by all versions of the simulation code, which actually exist as different versions of the command file. All code is written according to the rules of structured programming. Modularity is enforced by not using COMMON statements, simplifying the relation of the data flow to a hierarchy diagram. Simulation results are similar to those of the PARMTEQ code, as expected, because of the similar physical model. Different capabilities, such as those for generating beams matched in detail to the structure, are available in the new code for help in testing new ideas in designing RFQ linacs
Towards advanced code simulators
International Nuclear Information System (INIS)
Scriven, A.H.
1990-01-01
The Central Electricity Generating Board (CEGB) uses advanced thermohydraulic codes extensively to support PWR safety analyses. A system has been developed to allow fully interactive execution of any code with graphical simulation of the operator desk and mimic display. The system operates in a virtual machine environment, with the thermohydraulic code executing in one virtual machine, communicating via interrupts with any number of other virtual machines each running other programs and graphics drivers. The driver code itself does not have to be modified from its normal batch form. Shortly following the release of RELAP5 MOD1 in IBM compatible form in 1983, this code was used as the driver for this system. When RELAP5 MOD2 became available, it was adopted with no changes needed in the basic system. Overall the system has been used for some 5 years for the analysis of LOBI tests, full scale plant studies and for simple what-if studies. For gaining rapid understanding of system dependencies it has proved invaluable. The graphical mimic system, being independent of the driver code, has also been used with other codes to study core rewetting, to replay results obtained from batch jobs on a CRAY2 computer system and to display suitably processed experimental results from the LOBI facility to aid interpretation. For the above work real-time execution was not necessary. Current work now centers on implementing the RELAP 5 code on a true parallel architecture machine. Marconi Simulation have been contracted to investigate the feasibility of using upwards of 100 processors, each capable of a peak of 30 MIPS to run a highly detailed RELAP5 model in real time, complete with specially written 3D core neutronics and balance of plant models. This paper describes the experience of using RELAP5 as an analyzer/simulator, and outlines the proposed methods and problems associated with parallel execution of RELAP5
Building a dynamic code to simulate new reactor concepts
International Nuclear Information System (INIS)
Catsaros, N.; Gaveau, B.; Jaekel, M.-T.; Maillard, J.; Maurel, G.; Savva, P.; Silva, J.; Varvayanni, M.
2012-01-01
Highlights: ► We develop a stochastic neutronic code based on an existing High Energy Physics code. ► The code simulates innovative reactor designs including Accelerator Driven Systems. ► Core materials evolution will be dynamically simulated, including fuel burnup. ► Continuous feedback between the main inter-related parameters will be established. ► A description of the current research development and achievements is also given. - Abstract: Innovative nuclear reactor designs have been proposed, such as the Accelerator Driven Systems (ADSs), the “candle” reactors, etc. These reactor designs introduce computational nuclear technology problems the solution of which necessitates a new, global and dynamic computational approach of the system. A continuous feedback procedure must be established between the main inter-related parameters of the system such as the chemical, physical and isotopic composition of the core, the neutron flux distribution and the temperature field. Furthermore, as far as ADSs are concerned, the ability of the computational tool to simulate the nuclear cascade created from the interaction of accelerated protons with the spallation target as well as the produced neutrons, is also required. The new Monte Carlo code ANET (Advanced Neutronics with Evolution and Thermal hydraulic feedback) is being developed based on the GEANT3 High Energy Physics code, aiming to progressively satisfy all the above requirements. A description of the capabilities and methodologies implemented in the present version of ANET is given here, together with some illustrative applications of the code.
Development of a graphical interface computer code for reactor fuel reloading optimization
International Nuclear Information System (INIS)
Do Quang Binh; Nguyen Phuoc Lan; Bui Xuan Huy
2007-01-01
This report represents the results of the project performed in 2007. The aim of this project is to develop a graphical interface computer code that allows refueling engineers to design fuel reloading patterns for research reactor using simulated graphical model of reactor core. Besides, this code can perform refueling optimization calculations based on genetic algorithms as well as simulated annealing. The computer code was verified based on a sample problem, which relies on operational and experimental data of Dalat research reactor. This code can play a significant role in in-core fuel management practice at nuclear research reactor centers and in training. (author)
The use of micro-computers in the simulation of ion beam optics
International Nuclear Information System (INIS)
Spaedtke, P.; Ivens, D.
1989-01-01
With computer simulation codes specific problems of the ion beam optics can be studied, which is useful in the design as in optimization of existing systems. Several such codes have been developed, unfortunately requiring substantial computer resources. Recent advances of mini- and micro-computers have now made it possible to develop simulation codes which can be run on these small computers also. In this paper, some of these codes will be presented and their computing time discussed. (author)
Fel simulations using distributed computing
Einstein, J.; Biedron, S.G.; Freund, H.P.; Milton, S.V.; Van Der Slot, P. J M; Bernabeu, G.
2016-01-01
While simulation tools are available and have been used regularly for simulating light sources, including Free-Electron Lasers, the increasing availability and lower cost of accelerated computing opens up new opportunities. This paper highlights a method of how accelerating and parallelizing code
Development of parallel benchmark code by sheet metal forming simulator 'ITAS'
International Nuclear Information System (INIS)
Watanabe, Hiroshi; Suzuki, Shintaro; Minami, Kazuo
1999-03-01
This report describes the development of parallel benchmark code by sheet metal forming simulator 'ITAS'. ITAS is a nonlinear elasto-plastic analysis program by the finite element method for the purpose of the simulation of sheet metal forming. ITAS adopts the dynamic analysis method that computes displacement of sheet metal at every time unit and utilizes the implicit method with the direct linear equation solver. Therefore the simulator is very robust. However, it requires a lot of computational time and memory capacity. In the development of the parallel benchmark code, we designed the code by MPI programming to reduce the computational time. In numerical experiments on the five kinds of parallel super computers at CCSE JAERI, i.e., SP2, SR2201, SX-4, T94 and VPP300, good performances are observed. The result will be shown to the public through WWW so that the benchmark results may become a guideline of research and development of the parallel program. (author)
International Nuclear Information System (INIS)
Watanabe, Nishio; Shimomura, Yoshiharu
1985-01-01
The derivation of basic equations of the computer simulation code 'MARLOWE' was examined in detail, which was treated by the binary collision approximation developed by Robinson and Torrens. The 'MARLOWE' program was used for the simulation of the three dimensional structure of displacement cascade damages of Au, Cu and Al, which were generated by primary knock-on atoms (PKA) of 1 keV to 40 keV. Results were seriously affected by the selection of parameter of Frenkel defect formation E disp and ion movement E quit with the close Frenkel defect recombination criteria and E disp = E quit , it was found that E disp of 11 eV, 5 eV, 5 eV are reasonable for the simulation calculation for Au, Cu, and Al, respectively. Cascade seems to have subcascade structures even for 40 keV PKA. (author)
International Nuclear Information System (INIS)
Popescu, Lucretiu M.
2000-01-01
A computer code package (PTSIM) for particle transport Monte Carlo simulation was developed using object oriented techniques of design and programming. A flexible system for simulation of coupled photon, electron transport, facilitating development of efficient simulation applications, was obtained. For photons: Compton and photo-electric effects, pair production and Rayleigh interactions are simulated, while for electrons, a class II condensed history scheme was considered, in which catastrophic interactions (Moeller electron-electron interaction, bremsstrahlung, etc.) are treated in detail and all other interactions with reduced individual effect on electron history are grouped together using continuous slowing down approximation and energy straggling theories. Electron angular straggling is simulated using Moliere theory or a mixed model in which scatters at large angles are treated as distinct events. Comparisons with experimentally benchmarks for electron transmission and bremsstrahlung emissions energy and angular spectra, and for dose calculations are presented
Computer code for the atomistic simulation of lattice defects and dynamics
International Nuclear Information System (INIS)
Schiffgens, J.O.; Graves, N.J.; Oster, C.A.
1980-04-01
The computer code COMENT used to simulate the effect of temperature, impurities, and pre-existing defects on radiation-induced defect production mechanisms, defect properties, defect migration, and defect stability. This report documents Version IV of COMENT (models, methods, and implementation) and defines current code options. Version IV of COMENT generates only face-centered-cubic (fcc) crystal lattices. However, an effort was made to structure COMENT to allow addition of new options with a minimum of change in the existing version of the code. This document describes the calling program and thirty-two user-defined subroutines. Fourteen subroutines (ALOYORD, DASPKA, DFCT, DSLOAN, DSLOIN, EXPAND, POT1, POT2, POT3, POT4, POT5, POT6, POT7, and THRMAL) are associated with the selection of program options; only a few of these are used in any given analysis. Seven of the other subroutines (CRYSTL, IEAF, INCBOX, LABLE, MINILAT, SPEFORS, and SQUEZ are used to establish a variety of arrays and conditions required for each analysis; most of them are used once in a given calculation. The remaining eleven subroutines (DAMP, DIRECT, IDDEF, NEAF, INBIN, FILBIN, FTBIN, PAC3, UNPAC3, PACF, and UNPACF) are used many times in each calculation; the last eight of these are used many times in each time step during the integration and, therefore, are written in COMPASS (CDC assembly language). The COMPASS subroutines are described in sufficient detail to permit easy conversion to some other assembly language or to FORTRAN
International Nuclear Information System (INIS)
Shekhar Kumar; Koganti, S.B.
2003-07-01
Benchmarking and application of a computer code SIMPSEX for high plutonium FBR flowsheets was reported recently in an earlier report (IGC-234). Improvements and recompilation of the code (Version 4.01, March 2003) required re-validation with the existing benchmarks as well as additional benchmark flowsheets. Improvements in the high Pu region (Pu Aq >30 g/L) resulted in better results in the 75% Pu flowsheet benchmark. Below 30 g/L Pu Aq concentration, results were identical to those from the earlier version (SIMPSEX Version 3, code compiled in 1999). In addition, 13 published flowsheets were taken as additional benchmarks. Eleven of these flowsheets have a wide range of feed concentrations and few of them are β-γ active runs with FBR fuels having a wide distribution of burnup and Pu ratios. A published total partitioning flowsheet using externally generated U(IV) was also simulated using SIMPSEX. SIMPSEX predictions were compared with listed predictions from conventional SEPHIS, PUMA, PUNE and PUBG. SIMPSEX results were found to be comparable and better than the result from above listed codes. In addition, recently reported UREX demo results along with AMUSE simulations are also compared with SIMPSEX predictions. Results of the benchmarking SIMPSEX with these 14 benchmark flowsheets are discussed in this report. (author)
SURE: a system of computer codes for performing sensitivity/uncertainty analyses with the RELAP code
International Nuclear Information System (INIS)
Bjerke, M.A.
1983-02-01
A package of computer codes has been developed to perform a nonlinear uncertainty analysis on transient thermal-hydraulic systems which are modeled with the RELAP computer code. Using an uncertainty around the analyses of experiments in the PWR-BDHT Separate Effects Program at Oak Ridge National Laboratory. The use of FORTRAN programs running interactively on the PDP-10 computer has made the system very easy to use and provided great flexibility in the choice of processing paths. Several experiments simulating a loss-of-coolant accident in a nuclear reactor have been successfully analyzed. It has been shown that the system can be automated easily to further simplify its use and that the conversion of the entire system to a base code other than RELAP is possible
Vectorization of nuclear codes on FACOM 230-75 APU computer
International Nuclear Information System (INIS)
Harada, Hiroo; Higuchi, Kenji; Ishiguro, Misako; Tsutsui, Tsuneo; Fujii, Minoru
1983-02-01
To provide for the future usage of supercomputer, we have investigated the vector processing efficiency of nuclear codes which are being used at JAERI. The investigation is performed by using FACOM 230-75 APU computer. The codes are CITATION (3D neutron diffusion), SAP5 (structural analysis), CASCMARL (irradiation damage simulation). FEM-BABEL (3D neutron diffusion by FEM), GMSCOPE (microscope simulation). DWBA (cross section calculation at molecular collisions). A new type of cell density calculation for particle-in-cell method is also investigated. For each code we have obtained a significant speedup which ranges from 1.8 (CASCMARL) to 7.5 (GMSCOPE), respectively. We have described in this report the running time dynamic profile analysis of the codes, numerical algorithms used, program restructuring for the vectorization, numerical experiments of the iterative process, vectorized ratios, speedup ratios on the FACOM 230-75 APU computer, and some vectorization views. (author)
FAST: a three-dimensional time-dependent FEL simulation code
International Nuclear Information System (INIS)
Saldin, E.L.; Schneidmiller, E.A.; Yurkov, M.V.
1999-01-01
In this report we briefly describe the three-dimensional, time-dependent FEL simulation code FAST. The equations of motion of the particles and Maxwell's equations are solved simultaneously taking into account the slippage effect. Radiation fields are calculated using an integral solution of Maxwell's equations. A special technique has been developed for fast calculations of the radiation field, drastically reducing the required CPU time. As a result, the developed code allows one to use a personal computer for time-dependent simulations. The code allows one to simulate the radiation from the electron bunch of any transverse and longitudinal bunch shape; to simulate simultaneously an external seed with superimposed noise in the electron beam; to take into account energy spread in the electron beam and the space charge fields; and to simulate a high-gain, high-efficiency FEL amplifier with a tapered undulator. It is important to note that there are no significant memory limitations in the developed code and an electron bunch of any length can be simulated
Monte Carlo simulation code modernization
CERN. Geneva
2015-01-01
The continual development of sophisticated transport simulation algorithms allows increasingly accurate description of the effect of the passage of particles through matter. This modelling capability finds applications in a large spectrum of fields from medicine to astrophysics, and of course HEP. These new capabilities however come at the cost of a greater computational intensity of the new models, which has the effect of increasing the demands of computing resources. This is particularly true for HEP, where the demand for more simulation are driven by the need of both more accuracy and more precision, i.e. better models and more events. Usually HEP has relied on the "Moore's law" evolution, but since almost ten years the increase in clock speed has withered and computing capacity comes in the form of hardware architectures of many-core or accelerated processors. To harness these opportunities we need to adapt our code to concurrent programming models taking advantages of both SIMD and SIMT architectures. Th...
PLASMOR: A laser-plasma simulation code. Pt. 2
International Nuclear Information System (INIS)
Salzman, D.; Krumbein, A.D.; Szichman, H.
1987-06-01
This report supplements a previous one which describes the PLASMOR hydrodynamics code. The present report documents the recent changes and additions made in the code. In particular described are two new subroutines for radiative preheat, a system of preprocessors which prepare the code before run, a list of postprocessors which simulate experimental setups, and the basic data sets required to run PLASMOR. In the Appendix a new computer-based manual which lists the main features of PLASMOR is reproduced
Computer code to simulate transients in a steam generator of PWR nuclear power plants
International Nuclear Information System (INIS)
Silva, J.M. da.
1979-01-01
A digital computer code KIBE was developed to simulate the transient behavior of a Steam Generator used in Pressurized Water Reactor Power PLants. The equations of Conservation of mass, energy and momentum were numerically integrated by an implicit method progressively in the several axial sections into which the Steam Generator was divided. Forced convection heat transfer was assumed on the primary side, while on the secondary side all the different modes of heat transfer were permitted and deternined from the various correlations. The stability of the stationary state was verified by its reproducibility during the integration of the conservation equation without any pertubation. Transient behavior resulting from pertubations in the flow and the internal energy (temperature) at the inlet of the primary side were simulated. The results obtained exhibited satisfactory behaviour. (author) [pt
Accelerator simulation using computers
International Nuclear Information System (INIS)
Lee, M.; Zambre, Y.; Corbett, W.
1992-01-01
Every accelerator or storage ring system consists of a charged particle beam propagating through a beam line. Although a number of computer programs exits that simulate the propagation of a beam in a given beam line, only a few provide the capabilities for designing, commissioning and operating the beam line. This paper shows how a ''multi-track'' simulation and analysis code can be used for these applications
Computer codes for designing proton linear accelerators
International Nuclear Information System (INIS)
Kato, Takao
1992-01-01
Computer codes for designing proton linear accelerators are discussed from the viewpoint of not only designing but also construction and operation of the linac. The codes are divided into three categories according to their purposes: 1) design code, 2) generation and simulation code, and 3) electric and magnetic fields calculation code. The role of each category is discussed on the basis of experience at KEK (the design of the 40-MeV proton linac and its construction and operation, and the design of the 1-GeV proton linac). We introduce our recent work relevant to three-dimensional calculation and supercomputer calculation: 1) tuning of MAFIA (three-dimensional electric and magnetic fields calculation code) for supercomputer, 2) examples of three-dimensional calculation of accelerating structures by MAFIA, 3) development of a beam transport code including space charge effects. (author)
Icarus: A 2-D Direct Simulation Monte Carlo (DSMC) Code for Multi-Processor Computers
International Nuclear Information System (INIS)
BARTEL, TIMOTHY J.; PLIMPTON, STEVEN J.; GALLIS, MICHAIL A.
2001-01-01
Icarus is a 2D Direct Simulation Monte Carlo (DSMC) code which has been optimized for the parallel computing environment. The code is based on the DSMC method of Bird[11.1] and models from free-molecular to continuum flowfields in either cartesian (x, y) or axisymmetric (z, r) coordinates. Computational particles, representing a given number of molecules or atoms, are tracked as they have collisions with other particles or surfaces. Multiple species, internal energy modes (rotation and vibration), chemistry, and ion transport are modeled. A new trace species methodology for collisions and chemistry is used to obtain statistics for small species concentrations. Gas phase chemistry is modeled using steric factors derived from Arrhenius reaction rates or in a manner similar to continuum modeling. Surface chemistry is modeled with surface reaction probabilities; an optional site density, energy dependent, coverage model is included. Electrons are modeled by either a local charge neutrality assumption or as discrete simulational particles. Ion chemistry is modeled with electron impact chemistry rates and charge exchange reactions. Coulomb collision cross-sections are used instead of Variable Hard Sphere values for ion-ion interactions. The electro-static fields can either be: externally input, a Langmuir-Tonks model or from a Green's Function (Boundary Element) based Poison Solver. Icarus has been used for subsonic to hypersonic, chemically reacting, and plasma flows. The Icarus software package includes the grid generation, parallel processor decomposition, post-processing, and restart software. The commercial graphics package, Tecplot, is used for graphics display. All of the software packages are written in standard Fortran
Framework for utilizing computational devices within simulation
Directory of Open Access Journals (Sweden)
Miroslav Mintál
2013-12-01
Full Text Available Nowadays there exist several frameworks to utilize a computation power of graphics cards and other computational devices such as FPGA, ARM and multi-core processors. The best known are either low-level and need a lot of controlling code or are bounded only to special graphic cards. Furthermore there exist more specialized frameworks, mainly aimed to the mathematic field. Described framework is adjusted to use in a multi-agent simulations. Here it provides an option to accelerate computations when preparing simulation and mainly to accelerate a computation of simulation itself.
Energy Technology Data Exchange (ETDEWEB)
HOLM,ELIZABETH A.; BATTAILE,CORBETT C.; BUCHHEIT,THOMAS E.; FANG,HUEI ELIOT; RINTOUL,MARK DANIEL; VEDULA,VENKATA R.; GLASS,S. JILL; KNOROVSKY,GERALD A.; NEILSEN,MICHAEL K.; WELLMAN,GERALD W.; SULSKY,DEBORAH; SHEN,YU-LIN; SCHREYER,H. BUCK
2000-04-01
Computational materials simulations have traditionally focused on individual phenomena: grain growth, crack propagation, plastic flow, etc. However, real materials behavior results from a complex interplay between phenomena. In this project, the authors explored methods for coupling mesoscale simulations of microstructural evolution and micromechanical response. In one case, massively parallel (MP) simulations for grain evolution and microcracking in alumina stronglink materials were dynamically coupled. In the other, codes for domain coarsening and plastic deformation in CuSi braze alloys were iteratively linked. this program provided the first comparison of two promising ways to integrate mesoscale computer codes. Coupled microstructural/micromechanical codes were applied to experimentally observed microstructures for the first time. In addition to the coupled codes, this project developed a suite of new computational capabilities (PARGRAIN, GLAD, OOF, MPM, polycrystal plasticity, front tracking). The problem of plasticity length scale in continuum calculations was recognized and a solution strategy was developed. The simulations were experimentally validated on stockpile materials.
Computer codes and methods for simulating accelerator driven systems
International Nuclear Information System (INIS)
Sartori, E.; Byung Chan Na
2003-01-01
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)
GPU-accelerated micromagnetic simulations using cloud computing
Energy Technology Data Exchange (ETDEWEB)
Jermain, C.L., E-mail: clj72@cornell.edu [Cornell University, Ithaca, NY 14853 (United States); Rowlands, G.E.; Buhrman, R.A. [Cornell University, Ithaca, NY 14853 (United States); Ralph, D.C. [Cornell University, Ithaca, NY 14853 (United States); Kavli Institute at Cornell, Ithaca, NY 14853 (United States)
2016-03-01
Highly parallel graphics processing units (GPUs) can improve the speed of micromagnetic simulations significantly as compared to conventional computing using central processing units (CPUs). We present a strategy for performing GPU-accelerated micromagnetic simulations by utilizing cost-effective GPU access offered by cloud computing services with an open-source Python-based program for running the MuMax3 micromagnetics code remotely. We analyze the scaling and cost benefits of using cloud computing for micromagnetics. - Highlights: • The benefits of cloud computing for GPU-accelerated micromagnetics are examined. • We present the MuCloud software for running simulations on cloud computing. • Simulation run times are measured to benchmark cloud computing performance. • Comparison benchmarks are analyzed between CPU and GPU based solvers.
GPU-accelerated micromagnetic simulations using cloud computing
International Nuclear Information System (INIS)
Jermain, C.L.; Rowlands, G.E.; Buhrman, R.A.; Ralph, D.C.
2016-01-01
Highly parallel graphics processing units (GPUs) can improve the speed of micromagnetic simulations significantly as compared to conventional computing using central processing units (CPUs). We present a strategy for performing GPU-accelerated micromagnetic simulations by utilizing cost-effective GPU access offered by cloud computing services with an open-source Python-based program for running the MuMax3 micromagnetics code remotely. We analyze the scaling and cost benefits of using cloud computing for micromagnetics. - Highlights: • The benefits of cloud computing for GPU-accelerated micromagnetics are examined. • We present the MuCloud software for running simulations on cloud computing. • Simulation run times are measured to benchmark cloud computing performance. • Comparison benchmarks are analyzed between CPU and GPU based solvers.
ANNarchy: a code generation approach to neural simulations on parallel hardware
Vitay, Julien; Dinkelbach, Helge Ü.; Hamker, Fred H.
2015-01-01
Many modern neural simulators focus on the simulation of networks of spiking neurons on parallel hardware. Another important framework in computational neuroscience, rate-coded neural networks, is mostly difficult or impossible to implement using these simulators. We present here the ANNarchy (Artificial Neural Networks architect) neural simulator, which allows to easily define and simulate rate-coded and spiking networks, as well as combinations of both. The interface in Python has been designed to be close to the PyNN interface, while the definition of neuron and synapse models can be specified using an equation-oriented mathematical description similar to the Brian neural simulator. This information is used to generate C++ code that will efficiently perform the simulation on the chosen parallel hardware (multi-core system or graphical processing unit). Several numerical methods are available to transform ordinary differential equations into an efficient C++code. We compare the parallel performance of the simulator to existing solutions. PMID:26283957
Computational fluid dynamics simulations of light water reactor flows
International Nuclear Information System (INIS)
Tzanos, C.P.; Weber, D.P.
1999-01-01
Advances in computational fluid dynamics (CFD), turbulence simulation, and parallel computing have made feasible the development of three-dimensional (3-D) single-phase and two-phase flow CFD codes that can simulate fluid flow and heat transfer in realistic reactor geometries with significantly reduced reliance, especially in single phase, on empirical correlations. The objective of this work was to assess the predictive power and computational efficiency of a CFD code in the analysis of a challenging single-phase light water reactor problem, as well as to identify areas where further improvements are needed
International Nuclear Information System (INIS)
1988-03-01
HYDROCOIN is an international study for examining ground-water flow modeling strategies and their influence on safety assessments of geologic repositories for nuclear waste. This report summarizes only the combined NRC project temas' simulation efforts on the computer code bench-marking problems. The codes used to simulate thesee seven problems were SWIFT II, FEMWATER, UNSAT2M USGS-3D, AND TOUGH. In general, linear problems involving scalars such as hydraulic head were accurately simulated by both finite-difference and finite-element solution algorithms. Both types of codes produced accurate results even for complex geometrics such as intersecting fractures. Difficulties were encountered in solving problems that invovled nonlinear effects such as density-driven flow and unsaturated flow. In order to fully evaluate the accuracy of these codes, post-processing of results using paricle tracking algorithms and calculating fluxes were examined. This proved very valuable by uncovering disagreements among code results even through the hydraulic-head solutions had been in agreement. 9 refs., 111 figs., 6 tabs
A simulation of a pebble bed reactor core by the MCNP-4C computer code
Directory of Open Access Journals (Sweden)
Bakhshayesh Moshkbar Khalil
2009-01-01
Full Text Available Lack of energy is a major crisis of our century; the irregular increase of fossil fuel costs has forced us to search for novel, cheaper, and safer sources of energy. Pebble bed reactors - an advanced new generation of reactors with specific advantages in safety and cost - might turn out to be the desired candidate for the role. The calculation of the critical height of a pebble bed reactor at room temperature, while using the MCNP-4C computer code, is the main goal of this paper. In order to reduce the MCNP computing time compared to the previously proposed schemes, we have devised a new simulation scheme. Different arrangements of kernels in fuel pebble simulations were investigated and the best arrangement to decrease the MCNP execution time (while keeping the accuracy of the results, chosen. The neutron flux distribution and control rods worth, as well as their shadowing effects, have also been considered in this paper. All calculations done for the HTR-10 reactor core are in good agreement with experimental results.
International Nuclear Information System (INIS)
Fletcher, C.D.
1986-01-01
The capability to perform thermal-hydraulic analyses of a space reactor using the ATHENA computer code is demonstrated. The fast reactor, liquid-lithium coolant loops, and lithium-filled heat pipes of the preliminary General electric SP-100 design were modeled with ATHENA. Two demonstration transient calculations were performed simulating accident conditions. Calculated results are available for display using the Nuclear Plant Analyzer color graphics analysis tool in addition to traditional plots. ATHENA-calculated results appear reasonable, both for steady state full power conditions, and for the two transients. This analysis represents the first known transient thermal-hydraulic simulation using an integral space reactor system model incorporating heat pipes. 6 refs., 17 figs., 1 tab
International Nuclear Information System (INIS)
Bertolotto, D.
2011-11-01
The current doctoral research is focused on the development and validation of a coupled computational tool, to combine the advantages of computational fluid dynamics (CFD) in analyzing complex flow fields and of state-of-the-art system codes employed for nuclear power plant (NPP) simulations. Such a tool can considerably enhance the analysis of NPP transient behavior, e.g. in the case of pressurized water reactor (PWR) accident scenarios such as Main Steam Line Break (MSLB) and boron dilution, in which strong coolant flow asymmetries and multi-dimensional mixing effects strongly influence the reactivity of the reactor core, as described in Chap. 1. To start with, a literature review on code coupling is presented in Chap. 2, together with the corresponding ongoing projects in the international community. Special reference is made to the framework in which this research has been carried out, i.e. the Paul Scherrer Institute's (PSI) project STARS (Steady-state and Transient Analysis Research for the Swiss reactors). In particular, the codes chosen for the coupling, i.e. the CFD code ANSYS CFX V11.0 and the system code US-NRC TRACE V5.0, are part of the STARS codes system. Their main features are also described in Chap. 2. The development of the coupled tool, named CFX/TRACE from the names of the two constitutive codes, has proven to be a complex and broad-based task, and therefore constraints had to be put on the target requirements, while keeping in mind a certain modularity to allow future extensions to be made with minimal efforts. After careful consideration, the coupling was defined to be on-line, parallel and with non-overlapping domains connected by an interface, which was developed through the Parallel Virtual Machines (PVM) software, as described in Chap. 3. Moreover, two numerical coupling schemes were implemented and tested: a sequential explicit scheme and a sequential semi-implicit scheme. Finally, it was decided that the coupling would be single
Parallelization of simulation code for liquid-gas model of lattice-gas fluid
International Nuclear Information System (INIS)
Kawai, Wataru; Ebihara, Kenichi; Kume, Etsuo; Watanabe, Tadashi
2000-03-01
A simulation code for hydrodynamical phenomena which is based on the liquid-gas model of lattice-gas fluid is parallelized by using MPI (Message Passing Interface) library. The parallelized code can be applied to the larger size of the simulations than the non-parallelized code. The calculation times of the parallelized code on VPP500 (Vector-Parallel super computer with dispersed memory units), AP3000 (Scalar-parallel server with dispersed memory units), and a workstation cluster decreased in inverse proportion to the number of processors. (author)
A first accident simulation for Angra-1 power plant using the ALMOD computer code
International Nuclear Information System (INIS)
Camargo, C.T.M.
1981-02-01
The acquisition of the Almod computer code from GRS-Munich to CNEN has permited doing calculations of transients in PWR nuclear power plants, in which doesn't occur loss of coolant. The implementation of the german computer code Almod and its application in the calculation of Angra-1, a nuclear power plant different from the KWU power plants, demanded study and models adaptation; and due to economic reasons simplifications and optimizations were necessary. The first results define the analytical potential of the computer code, confirm the adequacy of the adaptations done and provide relevant conclusions about the Angra-1 safety analysis, showing at the same time areas in which the model can be applied or simply improved. (Author) [pt
Development of an integral computer code for simulation of heat exchangers
International Nuclear Information System (INIS)
Horvat, A.; Catton, I.
2001-01-01
Heat exchangers are one of the basic installations in power and process industries. The present guidelines provide an ad-hoc solution to certain design problems. A unified approach based on simultaneous modeling of thermal-hydraulics and structural behavior does not exist. The present paper describes the development of integral numerical code for simulation of heat exchangers. The code is based on Volume Averaging Technique (VAT) for porous media flow modeling. The calculated values of the whole-section drag and heat transfer coefficients show an excellent agreement with already published values. The matching results prove the correctness of the selected approach and verify the developed numerical code used for this calculation.(author)
Thermal-hydraulic and aerosol containment phenomena modelling in ASTEC severe accident computer code
International Nuclear Information System (INIS)
Kljenak, Ivo; Dapper, Maik; Dienstbier, Jiri; Herranz, Luis E.; Koch, Marco K.; Fontanet, Joan
2010-01-01
Transients in containment systems of different scales (Phebus.FP containment, KAEVER vessel, Battelle Model Containment, LACE vessel and VVER-1000 nuclear power plant containment) involving thermal-hydraulic phenomena and aerosol behaviour, were simulated with the computer integral code ASTEC. The results of the simulations in the first four facilities were compared with experimental results, whereas the results of the simulated accident in the VVER-1000 containment were compared to results obtained with the MELCOR code. The main purpose of the simulations was the validation of the CPA module of the ASTEC code. The calculated results support the applicability of the code for predicting in-containment thermal-hydraulic and aerosol phenomena during a severe accident in a nuclear power plant.
International Nuclear Information System (INIS)
Schelonka, E.P.
1979-01-01
Development and application of a series of simulation codes used for computer security analysis and design are described. Boolean relationships for arrays of barriers within functional modules are used to generate composite effectiveness indices. The general case of multiple layers of protection with any specified barrier survival criteria is given. Generalized reduction algorithms provide numerical security indices in selected subcategories and for the system as a whole. 9 figures, 11 tables
Tutorial: Parallel Computing of Simulation Models for Risk Analysis.
Reilly, Allison C; Staid, Andrea; Gao, Michael; Guikema, Seth D
2016-10-01
Simulation models are widely used in risk analysis to study the effects of uncertainties on outcomes of interest in complex problems. Often, these models are computationally complex and time consuming to run. This latter point may be at odds with time-sensitive evaluations or may limit the number of parameters that are considered. In this article, we give an introductory tutorial focused on parallelizing simulation code to better leverage modern computing hardware, enabling risk analysts to better utilize simulation-based methods for quantifying uncertainty in practice. This article is aimed primarily at risk analysts who use simulation methods but do not yet utilize parallelization to decrease the computational burden of these models. The discussion is focused on conceptual aspects of embarrassingly parallel computer code and software considerations. Two complementary examples are shown using the languages MATLAB and R. A brief discussion of hardware considerations is located in the Appendix. © 2016 Society for Risk Analysis.
Fire simulation in nuclear facilities: the FIRAC code and supporting experiments
International Nuclear Information System (INIS)
Burkett, M.W.; Martin, R.A.; Fenton, D.L.; Gunaji, M.V.
1984-01-01
The fire accident analysis computer code FIRAC was designed to estimate radioactive and nonradioactive source terms and predict fire-induced flows and thermal and material transport within the ventilation systems of nuclear fuel cycle facilities. FIRAC maintains its basic structure and features and has been expanded and modified to include the capabilities of the zone-type compartment fire model computer code FIRIN developed by Battelle Pacific Northwest Laboratory. The two codes have been coupled to provide an improved simulation of a fire-induced transient within a facility. The basic material transport capability of FIRAC has been retained and includes estimates of entrainment, convection, deposition, and filtration of material. The interrelated effects of filter plugging, heat transfer, gas dynamics, material transport, and fire and radioactive source terms also can be simulated. Also, a sample calculation has been performed to illustrate some of the capabilities of the code and how a typical facility is modeled with FIRAC. In addition to the analytical work being performed at Los Alamos, experiments are being conducted at the New Mexico State University to support the FIRAC computer code development and verification. This paper summarizes two areas of the experimental work that support the material transport capabiities of the code: the plugging of high-efficiency particulate air (HEPA) filters by combustion aerosols and the transport and deposition of smoke in ventilation system ductwork
Fire simulation in nuclear facilities--the FIRAC code and supporting experiments
International Nuclear Information System (INIS)
Burkett, M.W.; Martin, R.A.; Fenton, D.L.; Gunaji, M.V.
1985-01-01
The fire accident analysis computer code FIRAC was designed to estimate radioactive and nonradioactive source terms and predict fire-induced flows and thermal and material transport within the ventilation systems of nuclear fuel cycle facilities. FIRAC maintains its basic structure and features and has been expanded and modified to include the capabilities of the zone-type compartment fire model computer code FIRIN developed by Battelle Pacific Northwest Laboratory. The two codes have been coupled to provide an improved simulation of a fire-induced transient within a facility. The basic material transport capability of FIRAC has been retained and includes estimates of entrainment, convection, deposition, and filtration of material. The interrelated effects of filter plugging, heat transfer, gas dynamics, material transport, and fire and radioactive source terms also can be simulated. Also, a sample calculation has been performed to illustrate some of the capabilities of the code and how a typical facility is modeled with FIRAC. In addition to the analytical work being performed at Los Alamos, experiments are being conducted at the New Mexico State University to support the FIRAC computer code development and verification. This paper summarizes two areas of the experimental work that support the material transport capabilities of the code: the plugging of high-efficiency particulate air (HEPA) filters by combustion aerosols and the transport and deposition of smoke in ventilation system ductwork
Independent validation testing of the FLAME computer code, Version 1.0
International Nuclear Information System (INIS)
Martian, P.; Chung, J.N.
1992-07-01
Independent testing of the FLAME computer code, Version 1.0, was conducted to determine if the code is ready for use in hydrological and environmental studies at Department of Energy sites. This report describes the technical basis, approach, and results of this testing. Validation tests, (i.e., tests which compare field data to the computer generated solutions) were used to determine the operational status of the FLAME computer code and were done on a qualitative basis through graphical comparisons of the experimental and numerical data. These tests were specifically designed to check: (1) correctness of the FORTRAN coding, (2) computational accuracy, and (3) suitability to simulating actual hydrologic conditions. This testing was performed using a structured evaluation protocol which consisted of: (1) independent applications, and (2) graduated difficulty of test cases. Three tests ranging in complexity from simple one-dimensional steady-state flow field problems under near-saturated conditions to two-dimensional transient flow problems with very dry initial conditions
Distribution of absorbed dose in human eye simulated by SRNA-2KG computer code
International Nuclear Information System (INIS)
Ilic, R.; Pesic, M.; Pavlovic, R.; Mostacci, D.
2003-01-01
Rapidly increasing performances of personal computers and development of codes for proton transport based on Monte Carlo methods will allow, very soon, the introduction of the computer planning proton therapy as a normal activity in regular hospital procedures. A description of SRNA code used for such applications and results of calculated distributions of proton-absorbed dose in human eye are given in this paper. (author)
Improvement of the computing speed of the FBR fuel pin bundle deformation analysis code 'BAMBOO'
International Nuclear Information System (INIS)
Ito, Masahiro; Uwaba, Tomoyuki
2005-04-01
JNC has developed a coupled analysis system of a fuel pin bundle deformation analysis code 'BAMBOO' and a thermal hydraulics analysis code ASFRE-IV' for the purpose of evaluating the integrity of a subassembly under the BDI condition. This coupled analysis took much computation time because it needs convergent calculations to obtain numerically stationary solutions for thermal and mechanical behaviors. We improved the computation time of the BAMBOO code analysis to make the coupled analysis practicable. 'BAMBOO' is a FEM code and as such its matrix calculations consume large memory area to temporarily stores intermediate results in the solution of simultaneous linear equations. The code used the Hard Disk Drive (HDD) for the virtual memory area to save Random Access Memory (RAM) of the computer. However, the use of the HDD increased the computation time because Input/Output (I/O) processing with the HDD took much time in data accesses. We improved the code in order that it could conduct I/O processing only with the RAM in matrix calculations and run with in high-performance computers. This improvement considerably increased the CPU occupation rate during the simulation and reduced the total simulation time of the BAMBOO code to about one-seventh of that before the improvement. (author)
International Nuclear Information System (INIS)
Rakhno, I.L.; Roginets, L.P.
1999-01-01
When performing radiation treatment of products using an electron beam much time and money should be spent for numerous measurements to make optimal choice of treatment mode. Direct radiation treatment simulation by means of the EGS4 computer code fails to describe such measurement results correctly. In the paper a multi-step radiation treatment planning procedure is suggested which consists in fitting the EGS4 simulation results to reference measurement results, and using the fitted electron beam parameters and other ones in subsequent computer simulations. It is shown that the fitting procedure should be performed separately for each material or product type. The procedure suggested allows to replace measurements by computer simulations and therefore reduces significantly time and money required for such measurements. (author)
TESLA: Large Signal Simulation Code for Klystrons
International Nuclear Information System (INIS)
Vlasov, Alexander N.; Cooke, Simon J.; Chernin, David P.; Antonsen, Thomas M. Jr.; Nguyen, Khanh T.; Levush, Baruch
2003-01-01
TESLA (Telegraphist's Equations Solution for Linear Beam Amplifiers) is a new code designed to simulate linear beam vacuum electronic devices with cavities, such as klystrons, extended interaction klystrons, twistrons, and coupled cavity amplifiers. The model includes a self-consistent, nonlinear solution of the three-dimensional electron equations of motion and the solution of time-dependent field equations. The model differs from the conventional Particle in Cell approach in that the field spectrum is assumed to consist of a carrier frequency and its harmonics with slowly varying envelopes. Also, fields in the external cavities are modeled with circuit like equations and couple to fields in the beam region through boundary conditions on the beam tunnel wall. The model in TESLA is an extension of the model used in gyrotron code MAGY. The TESLA formulation has been extended to be capable to treat the multiple beam case, in which each beam is transported inside its own tunnel. The beams interact with each other as they pass through the gaps in their common cavities. The interaction is treated by modification of the boundary conditions on the wall of each tunnel to include the effect of adjacent beams as well as the fields excited in each cavity. The extended version of TESLA for the multiple beam case, TESLA-MB, has been developed for single processor machines, and can run on UNIX machines and on PC computers with a large memory (above 2GB). The TESLA-MB algorithm is currently being modified to simulate multiple beam klystrons on multiprocessor machines using the MPI (Message Passing Interface) environment. The code TESLA has been verified by comparison with MAGIC for single and multiple beam cases. The TESLA code and the MAGIC code predict the same power within 1% for a simple two cavity klystron design while the computational time for TESLA is orders of magnitude less than for MAGIC 2D. In addition, recently TESLA was used to model the L-6048 klystron, code
Sensitivity analysis of the titan hybrid deterministic transport code for SPECT simulation
International Nuclear Information System (INIS)
Royston, Katherine K.; Haghighat, Alireza
2011-01-01
Single photon emission computed tomography (SPECT) has been traditionally simulated using Monte Carlo methods. The TITAN code is a hybrid deterministic transport code that has recently been applied to the simulation of a SPECT myocardial perfusion study. For modeling SPECT, the TITAN code uses a discrete ordinates method in the phantom region and a combined simplified ray-tracing algorithm with a fictitious angular quadrature technique to simulate the collimator and generate projection images. In this paper, we compare the results of an experiment with a physical phantom with predictions from the MCNP5 and TITAN codes. While the results of the two codes are in good agreement, they differ from the experimental data by ∼ 21%. In order to understand these large differences, we conduct a sensitivity study by examining the effect of different parameters including heart size, collimator position, collimator simulation parameter, and number of energy groups. (author)
Nishiura, Daisuke; Furuichi, Mikito; Sakaguchi, Hide
2015-09-01
The computational performance of a smoothed particle hydrodynamics (SPH) simulation is investigated for three types of current shared-memory parallel computer devices: many integrated core (MIC) processors, graphics processing units (GPUs), and multi-core CPUs. We are especially interested in efficient shared-memory allocation methods for each chipset, because the efficient data access patterns differ between compute unified device architecture (CUDA) programming for GPUs and OpenMP programming for MIC processors and multi-core CPUs. We first introduce several parallel implementation techniques for the SPH code, and then examine these on our target computer architectures to determine the most effective algorithms for each processor unit. In addition, we evaluate the effective computing performance and power efficiency of the SPH simulation on each architecture, as these are critical metrics for overall performance in a multi-device environment. In our benchmark test, the GPU is found to produce the best arithmetic performance as a standalone device unit, and gives the most efficient power consumption. The multi-core CPU obtains the most effective computing performance. The computational speed of the MIC processor on Xeon Phi approached that of two Xeon CPUs. This indicates that using MICs is an attractive choice for existing SPH codes on multi-core CPUs parallelized by OpenMP, as it gains computational acceleration without the need for significant changes to the source code.
The NEST Dry-Run Mode: Efficient Dynamic Analysis of Neuronal Network Simulation Code
Directory of Open Access Journals (Sweden)
Susanne Kunkel
2017-06-01
Full Text Available NEST is a simulator for spiking neuronal networks that commits to a general purpose approach: It allows for high flexibility in the design of network models, and its applications range from small-scale simulations on laptops to brain-scale simulations on supercomputers. Hence, developers need to test their code for various use cases and ensure that changes to code do not impair scalability. However, running a full set of benchmarks on a supercomputer takes up precious compute-time resources and can entail long queuing times. Here, we present the NEST dry-run mode, which enables comprehensive dynamic code analysis without requiring access to high-performance computing facilities. A dry-run simulation is carried out by a single process, which performs all simulation steps except communication as if it was part of a parallel environment with many processes. We show that measurements of memory usage and runtime of neuronal network simulations closely match the corresponding dry-run data. Furthermore, we demonstrate the successful application of the dry-run mode in the areas of profiling and performance modeling.
Numerical simulation of fragmentation of hot metal and oxide melts with the computer code IVA3
International Nuclear Information System (INIS)
Mussa, S.; Tromm, W.
1994-01-01
The phenomena of fragmentation of melts caused by water-inlet from the bottom with the computer code IVA3/11,12,13/ are investigated. With the computer code IVA3 three-component-multiphase flows can be numerically simulated. Two geometrical models are used. Both consist of a cylindrical vessel for water lying beneath a cylindrical vessel for melt. The vessels are connected to each other through a hole. Steel and UO 2 melts are. The following parameters were varied: the type of the melt (steel,UO 2 ), the water supply pressure and the geometry of the hole in the bottom plate through which the water and melt vessels are connected. As results of the numerical simulations temperature and pressure versus time curves are plotted. Additionally the volume flow rates and the volume fractions of the various phases in the vessels and the increase in surface and enthalpy of the melt during the time of simulation are depicted. With steel melts the rate of fragmentation increases with increasing water pressure and melt temperature, whereby stable channels are formed in the melt layer showing a very low flow resistance for steam. With UO 2 the formations of channels are also observed. However, these channels are not so stable that they eventually break apart and lead to the fragmentation of the UO 2 melt in drops. The fragmentation of the steel melt in water vessel is less than that of UO 2 . No essential solidification of the melt is observed in the respective duration of the simulations. However, a small drop in the melt temperature is observed. With a slight or no water pressure the melt flows from the upper vessel into the water vessel via the connecting hole. The processes take place in a very slow manner and with such a low steam production so that despite the occuring pressure peaks no sign of steam explosions could be observed. (orig./HP) [de
A three-dimensional magnetostatics computer code for insertion devices
International Nuclear Information System (INIS)
Chubar, O.; Elleaume, P.; Chavanne, J.
1998-01-01
RADIA is a three-dimensional magnetostatics computer code optimized for the design of undulators and wigglers. It solves boundary magnetostatics problems with magnetized and current-carrying volumes using the boundary integral approach. The magnetized volumes can be arbitrary polyhedrons with non-linear (iron) or linear anisotropic (permanent magnet) characteristics. The current-carrying elements can be straight or curved blocks with rectangular cross sections. Boundary conditions are simulated by the technique of mirroring. Analytical formulae used for the computation of the field produced by a magnetized volume of a polyhedron shape are detailed. The RADIA code is written in object-oriented C++ and interfaced to Mathematica (Mathematica is a registered trademark of Wolfram Research, Inc.). The code outperforms currently available finite-element packages with respect to the CPU time of the solver and accuracy of the field integral estimations. An application of the code to the case of a wedge-pole undulator is presented
Linking CATHENA with other computer codes through a remote process
Energy Technology Data Exchange (ETDEWEB)
Vasic, A.; Hanna, B.N.; Waddington, G.M. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Sabourin, G. [Atomic Energy of Canada Limited, Montreal, Quebec (Canada); Girard, R. [Hydro-Quebec, Montreal, Quebec (Canada)
2005-07-01
'Full text:' CATHENA (Canadian Algorithm for THErmalhydraulic Network Analysis) is a computer code developed by Atomic Energy of Canada Limited (AECL). The code uses a transient, one-dimensional, two-fluid representation of two-phase flow in piping networks. CATHENA is used primarily for the analysis of postulated upset conditions in CANDU reactors; however, the code has found a wider range of applications. In the past, the CATHENA thermalhydraulics code included other specialized codes, i.e. ELOCA and the Point LEPreau CONtrol system (LEPCON) as callable subroutine libraries. The combined program was compiled and linked as a separately named code. This code organizational process is not suitable for independent development, maintenance, validation and version tracking of separate computer codes. The alternative solution to provide code development independence is to link CATHENA to other computer codes through a Parallel Virtual Machine (PVM) interface process. PVM is a public domain software package, developed by Oak Ridge National Laboratory and enables a heterogeneous collection of computers connected by a network to be used as a single large parallel machine. The PVM approach has been well accepted by the global computing community and has been used successfully for solving large-scale problems in science, industry, and business. Once development of the appropriate interface for linking independent codes through PVM is completed, future versions of component codes can be developed, distributed separately and coupled as needed by the user. This paper describes the coupling of CATHENA to the ELOCA-IST and the TROLG2 codes through a PVM remote process as an illustration of possible code connections. ELOCA (Element Loss Of Cooling Analysis) is the Industry Standard Toolset (IST) code developed by AECL to simulate the thermo-mechanical response of CANDU fuel elements to transient thermalhydraulics boundary conditions. A separate ELOCA driver program
Linking CATHENA with other computer codes through a remote process
International Nuclear Information System (INIS)
Vasic, A.; Hanna, B.N.; Waddington, G.M.; Sabourin, G.; Girard, R.
2005-01-01
'Full text:' CATHENA (Canadian Algorithm for THErmalhydraulic Network Analysis) is a computer code developed by Atomic Energy of Canada Limited (AECL). The code uses a transient, one-dimensional, two-fluid representation of two-phase flow in piping networks. CATHENA is used primarily for the analysis of postulated upset conditions in CANDU reactors; however, the code has found a wider range of applications. In the past, the CATHENA thermalhydraulics code included other specialized codes, i.e. ELOCA and the Point LEPreau CONtrol system (LEPCON) as callable subroutine libraries. The combined program was compiled and linked as a separately named code. This code organizational process is not suitable for independent development, maintenance, validation and version tracking of separate computer codes. The alternative solution to provide code development independence is to link CATHENA to other computer codes through a Parallel Virtual Machine (PVM) interface process. PVM is a public domain software package, developed by Oak Ridge National Laboratory and enables a heterogeneous collection of computers connected by a network to be used as a single large parallel machine. The PVM approach has been well accepted by the global computing community and has been used successfully for solving large-scale problems in science, industry, and business. Once development of the appropriate interface for linking independent codes through PVM is completed, future versions of component codes can be developed, distributed separately and coupled as needed by the user. This paper describes the coupling of CATHENA to the ELOCA-IST and the TROLG2 codes through a PVM remote process as an illustration of possible code connections. ELOCA (Element Loss Of Cooling Analysis) is the Industry Standard Toolset (IST) code developed by AECL to simulate the thermo-mechanical response of CANDU fuel elements to transient thermalhydraulics boundary conditions. A separate ELOCA driver program starts, ends
Numerical simulation code for combustion of sodium liquid droplet and its verification
International Nuclear Information System (INIS)
Okano, Yasushi
1997-11-01
The computer programs for sodium leak and burning phenomena had been developed based on mechanistic approach. Direct numerical simulation code for sodium liquid droplet burning had been developed for numerical analysis of droplet combustion in forced convection air flow. Distributions of heat generation and temperature and reaction rate of chemical productions, such as sodium oxide and hydroxide, are calculated and evaluated with using this numerical code. Extended MAC method coupled with a higher-order upwind scheme had been used for combustion simulation of methane-air mixture. In the numerical simulation code for combustion of sodium liquid droplet, chemical reaction model of sodium was connected with the extended MAC method. Combustion of single sodium liquid droplet was simulated in this report for the verification of developed numerical simulation code. The changes of burning rate and reaction product with droplet diameter and inlet wind velocity were investigated. These calculation results were qualitatively and quantitatively conformed to the experimental and calculation observations in combustion engineering. It was confirmed that the numerical simulation code was available for the calculation of sodium liquid droplet burning. (author)
Optimization of the particle pusher in a diode simulation code
International Nuclear Information System (INIS)
Theimer, M.M.; Quintenz, J.P.
1979-09-01
The particle pusher in Sandia's particle-in-cell diode simulation code has been rewritten to reduce the required run time of a typical simulation. The resulting new version of the code has been found to run up to three times as fast as the original with comparable accuracy. The cost of this optimization was an increase in storage requirements of about 15%. The new version has also been written to run efficiently on a CRAY-1 computing system. Steps taken to affect this reduced run time are described. Various test cases are detailed
Monte Carlo simulation of Ising models by multispin coding on a vector computer
Wansleben, Stephan; Zabolitzky, John G.; Kalle, Claus
1984-11-01
Rebbi's efficient multispin coding algorithm for Ising models is combined with the use of the vector computer CDC Cyber 205. A speed of 21.2 million updates per second is reached. This is comparable to that obtained by special- purpose computers.
The TeraShake Computational Platform for Large-Scale Earthquake Simulations
Cui, Yifeng; Olsen, Kim; Chourasia, Amit; Moore, Reagan; Maechling, Philip; Jordan, Thomas
Geoscientific and computer science researchers with the Southern California Earthquake Center (SCEC) are conducting a large-scale, physics-based, computationally demanding earthquake system science research program with the goal of developing predictive models of earthquake processes. The computational demands of this program continue to increase rapidly as these researchers seek to perform physics-based numerical simulations of earthquake processes for larger meet the needs of this research program, a multiple-institution team coordinated by SCEC has integrated several scientific codes into a numerical modeling-based research tool we call the TeraShake computational platform (TSCP). A central component in the TSCP is a highly scalable earthquake wave propagation simulation program called the TeraShake anelastic wave propagation (TS-AWP) code. In this chapter, we describe how we extended an existing, stand-alone, wellvalidated, finite-difference, anelastic wave propagation modeling code into the highly scalable and widely used TS-AWP and then integrated this code into the TeraShake computational platform that provides end-to-end (initialization to analysis) research capabilities. We also describe the techniques used to enhance the TS-AWP parallel performance on TeraGrid supercomputers, as well as the TeraShake simulations phases including input preparation, run time, data archive management, and visualization. As a result of our efforts to improve its parallel efficiency, the TS-AWP has now shown highly efficient strong scaling on over 40K processors on IBM’s BlueGene/L Watson computer. In addition, the TSCP has developed into a computational system that is useful to many members of the SCEC community for performing large-scale earthquake simulations.
Connecting Neural Coding to Number Cognition: A Computational Account
Prather, Richard W.
2012-01-01
The current study presents a series of computational simulations that demonstrate how the neural coding of numerical magnitude may influence number cognition and development. This includes behavioral phenomena cataloged in cognitive literature such as the development of numerical estimation and operational momentum. Though neural research has…
Quantum computation with Turaev-Viro codes
International Nuclear Information System (INIS)
Koenig, Robert; Kuperberg, Greg; Reichardt, Ben W.
2010-01-01
For a 3-manifold with triangulated boundary, the Turaev-Viro topological invariant can be interpreted as a quantum error-correcting code. The code has local stabilizers, identified by Levin and Wen, on a qudit lattice. Kitaev's toric code arises as a special case. The toric code corresponds to an abelian anyon model, and therefore requires out-of-code operations to obtain universal quantum computation. In contrast, for many categories, such as the Fibonacci category, the Turaev-Viro code realizes a non-abelian anyon model. A universal set of fault-tolerant operations can be implemented by deforming the code with local gates, in order to implement anyon braiding. We identify the anyons in the code space, and present schemes for initialization, computation and measurement. This provides a family of constructions for fault-tolerant quantum computation that are closely related to topological quantum computation, but for which the fault tolerance is implemented in software rather than coming from a physical medium.
Uses of Computer Simulation Models in Ag-Research and Everyday Life
When the news media talks about models they could be talking about role models, fashion models, conceptual models like the auto industry uses, or computer simulation models. A computer simulation model is a computer code that attempts to imitate the processes and functions of certain systems. There ...
COMPBRN III: a computer code for modeling compartment fires
International Nuclear Information System (INIS)
Ho, V.; Siu, N.; Apostolakis, G.; Flanagan, G.F.
1986-07-01
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
Nexus: A modular workflow management system for quantum simulation codes
Krogel, Jaron T.
2016-01-01
The management of simulation workflows represents a significant task for the individual computational researcher. Automation of the required tasks involved in simulation work can decrease the overall time to solution and reduce sources of human error. A new simulation workflow management system, Nexus, is presented to address these issues. Nexus is capable of automated job management on workstations and resources at several major supercomputing centers. Its modular design allows many quantum simulation codes to be supported within the same framework. Current support includes quantum Monte Carlo calculations with QMCPACK, density functional theory calculations with Quantum Espresso or VASP, and quantum chemical calculations with GAMESS. Users can compose workflows through a transparent, text-based interface, resembling the input file of a typical simulation code. A usage example is provided to illustrate the process.
Geochemical computer codes. A review
International Nuclear Information System (INIS)
Andersson, K.
1987-01-01
In this report a review of available codes is performed and some code intercomparisons are also discussed. The number of codes treating natural waters (groundwater, lake water, sea water) is large. Most geochemical computer codes treat equilibrium conditions, although some codes with kinetic capability are available. A geochemical equilibrium model consists of a computer code, solving a set of equations by some numerical method and a data base, consisting of thermodynamic data required for the calculations. There are some codes which treat coupled geochemical and transport modeling. Some of these codes solve the equilibrium and transport equations simultaneously while other solve the equations separately from each other. The coupled codes require a large computer capacity and have thus as yet limited use. Three code intercomparisons have been found in literature. It may be concluded that there are many codes available for geochemical calculations but most of them require a user that us quite familiar with the code. The user also has to know the geochemical system in order to judge the reliability of the results. A high quality data base is necessary to obtain a reliable result. The best results may be expected for the major species of natural waters. For more complicated problems, including trace elements, precipitation/dissolution, adsorption, etc., the results seem to be less reliable. (With 44 refs.) (author)
ZENO: N-body and SPH Simulation Codes
Barnes, Joshua E.
2011-02-01
The ZENO software package integrates N-body and SPH simulation codes with a large array of programs to generate initial conditions and analyze numerical simulations. Written in C, the ZENO system is portable between Mac, Linux, and Unix platforms. It is in active use at the Institute for Astronomy (IfA), at NRAO, and possibly elsewhere. Zeno programs can perform a wide range of simulation and analysis tasks. While many of these programs were first created for specific projects, they embody algorithms of general applicability and embrace a modular design strategy, so existing code is easily applied to new tasks. Major elements of the system include: Structured data file utilities facilitate basic operations on binary data, including import/export of ZENO data to other systems.Snapshot generation routines create particle distributions with various properties. Systems with user-specified density profiles can be realized in collisionless or gaseous form; multiple spherical and disk components may be set up in mutual equilibrium.Snapshot manipulation routines permit the user to sift, sort, and combine particle arrays, translate and rotate particle configurations, and assign new values to data fields associated with each particle.Simulation codes include both pure N-body and combined N-body/SPH programs: Pure N-body codes are available in both uniprocessor and parallel versions.SPH codes offer a wide range of options for gas physics, including isothermal, adiabatic, and radiating models. Snapshot analysis programs calculate temporal averages, evaluate particle statistics, measure shapes and density profiles, compute kinematic properties, and identify and track objects in particle distributions.Visualization programs generate interactive displays and produce still images and videos of particle distributions; the user may specify arbitrary color schemes and viewing transformations.
Software Engineering for Scientific Computer Simulations
Post, Douglass E.; Henderson, Dale B.; Kendall, Richard P.; Whitney, Earl M.
2004-11-01
Computer simulation is becoming a very powerful tool for analyzing and predicting the performance of fusion experiments. Simulation efforts are evolving from including only a few effects to many effects, from small teams with a few people to large teams, and from workstations and small processor count parallel computers to massively parallel platforms. Successfully making this transition requires attention to software engineering issues. We report on the conclusions drawn from a number of case studies of large scale scientific computing projects within DOE, academia and the DoD. The major lessons learned include attention to sound project management including setting reasonable and achievable requirements, building a good code team, enforcing customer focus, carrying out verification and validation and selecting the optimum computational mathematics approaches.
International Nuclear Information System (INIS)
Rockhold, M.L.; Wurstner, S.K.
1991-03-01
The objective of this work was to test the ability of the PORFLO-3 computer code to simulate water infiltration and solute transport in dry soils. Data from a field-scale unsaturated zone flow and transport experiment, conducted near Las Cruces, New Mexico, were used for model validation. A spatial moment analysis was used to provide a quantitative basis for comparing the mean simulated and observed flow behavior. The scope of this work was limited to two-dimensional simulations of the second experiment at the Las Cruces trench site. Three simulation cases are presented. The first case represents a uniform soil profile, with homogeneous, isotropic hydraulic and transport properties. The second and third cases represent single stochastic realizations of randomly heterogeneous hydraulic conductivity fields, generated from the cumulative probability distribution of the measured data. Two-dimensional simulations produced water content changes that matched the observed data reasonably well. Models that explicitly incorporated heterogeneous hydraulic conductivity fields reproduced the characteristics of the observed data somewhat better than a uniform, homogeneous model. Improved predictions of water content changes at specific spatial locations were obtained by adjusting the soil hydraulic properties. The results of this study should only be considered a qualitative validation of the PORFLO-3 code. However, the results of this study demonstrate the importance of site-specific data for model calibration. Applications of the code for waste management and remediation activities will require site-specific data for model calibration before defensible predictions of unsaturated flow and containment transport can be made. 23 refs., 16 figs., 3 tabs
A general concurrent algorithm for plasma particle-in-cell simulation codes
International Nuclear Information System (INIS)
Liewer, P.C.; Decyk, V.K.
1989-01-01
We have developed a new algorithm for implementing plasma particle-in-cell (PIC) simulation codes on concurrent processors with distributed memory. This algorithm, named the general concurrent PIC algorithm (GCPIC), has been used to implement an electrostatic PIC code on the 33-node JPL Mark III Hypercube parallel computer. To decompose at PIC code using the GCPIC algorithm, the physical domain of the particle simulation is divided into sub-domains, equal in number to the number of processors, such that all sub-domains have roughly equal numbers of particles. For problems with non-uniform particle densities, these sub-domains will be of unequal physical size. Each processor is assigned a sub-domain and is responsible for updating the particles in its sub-domain. This algorithm has led to a a very efficient parallel implementation of a well-benchmarked 1-dimensional PIC code. The dominant portion of the code, updating the particle positions and velocities, is nearly 100% efficient when the number of particles is increased linearly with the number of hypercube processors used so that the number of particles per processor is constant. For example, the increase in time spent updating particles in going from a problem with 11,264 particles run on 1 processor to 360,448 particles on 32 processors was only 3% (parallel efficiency of 97%). Although implemented on a hypercube concurrent computer, this algorithm should also be efficient for PIC codes on other parallel architectures and for large PIC codes on sequential computers where part of the data must reside on external disks. copyright 1989 Academic Press, Inc
Computer Simulation Performed for Columbia Project Cooling System
Ahmad, Jasim
2005-01-01
This demo shows a high-fidelity simulation of the air flow in the main computer room housing the Columbia (10,024 intel titanium processors) system. The simulation asseses the performance of the cooling system and identified deficiencies, and recommended modifications to eliminate them. It used two in house software packages on NAS supercomputers: Chimera Grid tools to generate a geometric model of the computer room, OVERFLOW-2 code for fluid and thermal simulation. This state-of-the-art technology can be easily extended to provide a general capability for air flow analyses on any modern computer room. Columbia_CFD_black.tiff
International Nuclear Information System (INIS)
Twisdale, L.A.; Dunn, W.L.
1981-08-01
A probabilistic methodology has been developed to predict the probabilities of tornado-propelled missiles impacting and damaging nuclear power plant structures. Mathematical models of each event in the tornado missile hazard have been developed and sequenced to form an integrated, time-history simulation methodology. The models are data based where feasible. The data include documented records of tornado occurrence, field observations of missile transport, results of wind tunnel experiments, and missile impact tests. Probabilistic Monte Carlo techniques are used to estimate the risk probabilities. The methodology has been encoded in the TORMIS computer code to facilitate numerical analysis and plant-specific tornado missile probability assessments. Sensitivity analyses have been performed on both the individual models and the integrated methodology, and risk has been assessed for a hypothetical nuclear power plant design case study
International Nuclear Information System (INIS)
Mur, J.; Larrauri, D.
1998-07-01
Computer simulation of flow in configurations close to pressurized water reactor (PWR) geometry is of great interest for Electricite de France (EDF). Although simulation of the flow through a whole PWR core with an all purpose CFD-code is not yet achievable, such a tool cna be quite useful to perform numerical experiments in order to try and improve the modeling introduced in computer codes devoted to reactor core thermal-hydraulic analysis. Further to simulation in small bare rod bundle configurations, the present study is focused on the simulation, with CFD-code ESTET and PWR core code THYC, of the flow in the experimental configuration VATICAN-1. ESTET simulation results are compared on the one hand to local velocity and concentration measurements, on the other hand with subchannel averaged values calculated by THYC. As far as the comparison with measurements is concerned, ESTET results are quite satisfactory relatively to available experimental data and their uncertainties. The effect of spacer grids and the prediction of the evolution of an unbalanced velocity profile seem to be correctly treated. As far as the comparison with THYC subchannel averaged values is concerned, the difficulty of a direct comparison between subchannel averaged and local values is pointed out. ESTET calculated local values are close to experimental local values. ESTET subchannel averaged values are also close to THYC calculation results. Thus, THYC results are satisfactory whereas their direct comparison to local measurements could show some disagreement. (author)
Simulating Coupling Complexity in Space Plasmas: First Results from a new code
Kryukov, I.; Zank, G. P.; Pogorelov, N. V.; Raeder, J.; Ciardo, G.; Florinski, V. A.; Heerikhuisen, J.; Li, G.; Petrini, F.; Shematovich, V. I.; Winske, D.; Shaikh, D.; Webb, G. M.; Yee, H. M.
2005-12-01
The development of codes that embrace 'coupling complexity' via the self-consistent incorporation of multiple physical scales and multiple physical processes in models has been identified by the NRC Decadal Survey in Solar and Space Physics as a crucial necessary development in simulation/modeling technology for the coming decade. The National Science Foundation, through its Information Technology Research (ITR) Program, is supporting our efforts to develop a new class of computational code for plasmas and neutral gases that integrates multiple scales and multiple physical processes and descriptions. We are developing a highly modular, parallelized, scalable code that incorporates multiple scales by synthesizing 3 simulation technologies: 1) Computational fluid dynamics (hydrodynamics or magneto-hydrodynamics-MHD) for the large-scale plasma; 2) direct Monte Carlo simulation of atoms/neutral gas, and 3) transport code solvers to model highly energetic particle distributions. We are constructing the code so that a fourth simulation technology, hybrid simulations for microscale structures and particle distributions, can be incorporated in future work, but for the present, this aspect will be addressed at a test-particle level. This synthesis we will provide a computational tool that will advance our understanding of the physics of neutral and charged gases enormously. Besides making major advances in basic plasma physics and neutral gas problems, this project will address 3 Grand Challenge space physics problems that reflect our research interests: 1) To develop a temporal global heliospheric model which includes the interaction of solar and interstellar plasma with neutral populations (hydrogen, helium, etc., and dust), test-particle kinetic pickup ion acceleration at the termination shock, anomalous cosmic ray production, interaction with galactic cosmic rays, while incorporating the time variability of the solar wind and the solar cycle. 2) To develop a coronal
Validation and uncertainty analysis of the Athlet thermal-hydraulic computer code
International Nuclear Information System (INIS)
Glaeser, H.
1995-01-01
The computer code ATHLET is being developed by GRS as an advanced best-estimate code for the simulation of breaks and transients in Pressurized Water Reactor (PWRs) and Boiling Water Reactor (BWRs) including beyond design basis accidents. A systematic validation of ATHLET is based on a well balanced set of integral and separate effects tests emphasizing the German combined Emergency Core Cooling (ECC) injection system. When using best estimate codes for predictions of reactor plant states during assumed accidents, qualification of the uncertainty in these calculations is highly desirable. A method for uncertainty and sensitivity evaluation has been developed by GRS where the computational effort is independent of the number of uncertain parameters. (author)
Testing the new stochastic neutronic code ANET in simulating safety important parameters
International Nuclear Information System (INIS)
Xenofontos, T.; Delipei, G.-K.; Savva, P.; Varvayanni, M.; Maillard, J.; Silva, J.; Catsaros, N.
2017-01-01
Highlights: • ANET is a new neutronics stochastic code. • Criticality calculations in both subcritical and critical nuclear systems of conventional design were conducted. • Simulations of thermal, lower epithermal and fast neutron fluence rates were performed. • Axial fission rate distributions in standard and MOX fuel pins were computed. - Abstract: ANET (Advanced Neutronics with Evolution and Thermal hydraulic feedback) is an under development Monte Carlo code for simulating both GEN II/III reactors as well as innovative nuclear reactor designs, based on the high energy physics code GEANT3.21 of CERN. ANET is built through continuous GEANT3.21 applicability amplifications, comprising the simulation of particles’ transport and interaction in low energy along with the accessibility of user-provided libraries and tracking algorithms for energies below 20 MeV, as well as the simulation of elastic and inelastic collision, capture and fission. Successive testing applications performed throughout the ANET development have been utilized to verify the new code capabilities. In this context the ANET reliability in simulating certain reactor parameters important to safety is here examined. More specifically the reactor criticality as well as the neutron fluence and fission rates are benchmarked and validated. The Portuguese Research Reactor (RPI) after its conversion to low enrichment in U-235 and the OECD/NEA VENUS-2 MOX international benchmark were considered appropriate for the present study, the former providing criticality and neutron flux data and the latter reaction rates. Concerning criticality benchmarking, the subcritical, Training Nuclear Reactor of the Aristotle University of Thessaloniki (TNR-AUTh) was also analyzed. The obtained results are compared with experimental data from the critical infrastructures and with computations performed by two different, well established stochastic neutronics codes, i.e. TRIPOLI-4.8 and MCNP5. Satisfactory agreement
Relativistic modeling capabilities in PERSEUS extended MHD simulation code for HED plasmas
Energy Technology Data Exchange (ETDEWEB)
Hamlin, Nathaniel D., E-mail: nh322@cornell.edu [438 Rhodes Hall, Cornell University, Ithaca, NY, 14853 (United States); Seyler, Charles E., E-mail: ces7@cornell.edu [Cornell University, Ithaca, NY, 14853 (United States)
2014-12-15
We discuss the incorporation of relativistic modeling capabilities into the PERSEUS extended MHD simulation code for high-energy-density (HED) plasmas, and present the latest hybrid X-pinch simulation results. The use of fully relativistic equations enables the model to remain self-consistent in simulations of such relativistic phenomena as X-pinches and laser-plasma interactions. By suitable formulation of the relativistic generalized Ohm’s law as an evolution equation, we have reduced the recovery of primitive variables, a major technical challenge in relativistic codes, to a straightforward algebraic computation. Our code recovers expected results in the non-relativistic limit, and reveals new physics in the modeling of electron beam acceleration following an X-pinch. Through the use of a relaxation scheme, relativistic PERSEUS is able to handle nine orders of magnitude in density variation, making it the first fluid code, to our knowledge, that can simulate relativistic HED plasmas.
Independent verification and validation testing of the FLASH computer code, Versiion 3.0
International Nuclear Information System (INIS)
Martian, P.; Chung, J.N.
1992-06-01
Independent testing of the FLASH computer code, Version 3.0, was conducted to determine if the code is ready for use in hydrological and environmental studies at various Department of Energy sites. This report describes the technical basis, approach, and results of this testing. Verification tests, and validation tests, were used to determine the operational status of the FLASH computer code. These tests were specifically designed to test: correctness of the FORTRAN coding, computational accuracy, and suitability to simulating actual hydrologic conditions. This testing was performed using a structured evaluation protocol which consisted of: blind testing, independent applications, and graduated difficulty of test cases. Both quantitative and qualitative testing was performed through evaluating relative root mean square values and graphical comparisons of the numerical, analytical, and experimental data. Four verification test were used to check the computational accuracy and correctness of the FORTRAN coding, and three validation tests were used to check the suitability to simulating actual conditions. These tests cases ranged in complexity from simple 1-D saturated flow to 2-D variably saturated problems. The verification tests showed excellent quantitative agreement between the FLASH results and analytical solutions. The validation tests showed good qualitative agreement with the experimental data. Based on the results of this testing, it was concluded that the FLASH code is a versatile and powerful two-dimensional analysis tool for fluid flow. In conclusion, all aspects of the code that were tested, except for the unit gradient bottom boundary condition, were found to be fully operational and ready for use in hydrological and environmental studies
Alipchenkov, V. M.; Anfimov, A. M.; Afremov, D. A.; Gorbunov, V. S.; Zeigarnik, Yu. A.; Kudryavtsev, A. V.; Osipov, S. L.; Mosunova, N. A.; Strizhov, V. F.; Usov, E. V.
2016-02-01
The conceptual fundamentals of the development of the new-generation system thermal-hydraulic computational HYDRA-IBRAE/LM code are presented. The code is intended to simulate the thermalhydraulic processes that take place in the loops and the heat-exchange equipment of liquid-metal cooled fast reactor systems under normal operation and anticipated operational occurrences and during accidents. The paper provides a brief overview of Russian and foreign system thermal-hydraulic codes for modeling liquid-metal coolants and gives grounds for the necessity of development of a new-generation HYDRA-IBRAE/LM code. Considering the specific engineering features of the nuclear power plants (NPPs) equipped with the BN-1200 and the BREST-OD-300 reactors, the processes and the phenomena are singled out that require a detailed analysis and development of the models to be correctly described by the system thermal-hydraulic code in question. Information on the functionality of the computational code is provided, viz., the thermalhydraulic two-phase model, the properties of the sodium and the lead coolants, the closing equations for simulation of the heat-mass exchange processes, the models to describe the processes that take place during the steam-generator tube rupture, etc. The article gives a brief overview of the usability of the computational code, including a description of the support documentation and the supply package, as well as possibilities of taking advantages of the modern computer technologies, such as parallel computations. The paper shows the current state of verification and validation of the computational code; it also presents information on the principles of constructing of and populating the verification matrices for the BREST-OD-300 and the BN-1200 reactor systems. The prospects are outlined for further development of the HYDRA-IBRAE/LM code, introduction of new models into it, and enhancement of its usability. It is shown that the program of development and
Computational Particle Dynamic Simulations on Multicore Processors (CPDMu) Final Report Phase I
Energy Technology Data Exchange (ETDEWEB)
Schmalz, Mark S
2011-07-24
Statement of Problem - Department of Energy has many legacy codes for simulation of computational particle dynamics and computational fluid dynamics applications that are designed to run on sequential processors and are not easily parallelized. Emerging high-performance computing architectures employ massively parallel multicore architectures (e.g., graphics processing units) to increase throughput. Parallelization of legacy simulation codes is a high priority, to achieve compatibility, efficiency, accuracy, and extensibility. General Statement of Solution - A legacy simulation application designed for implementation on mainly-sequential processors has been represented as a graph G. Mathematical transformations, applied to G, produce a graph representation {und G} for a high-performance architecture. Key computational and data movement kernels of the application were analyzed/optimized for parallel execution using the mapping G {yields} {und G}, which can be performed semi-automatically. This approach is widely applicable to many types of high-performance computing systems, such as graphics processing units or clusters comprised of nodes that contain one or more such units. Phase I Accomplishments - Phase I research decomposed/profiled computational particle dynamics simulation code for rocket fuel combustion into low and high computational cost regions (respectively, mainly sequential and mainly parallel kernels), with analysis of space and time complexity. Using the research team's expertise in algorithm-to-architecture mappings, the high-cost kernels were transformed, parallelized, and implemented on Nvidia Fermi GPUs. Measured speedups (GPU with respect to single-core CPU) were approximately 20-32X for realistic model parameters, without final optimization. Error analysis showed no loss of computational accuracy. Commercial Applications and Other Benefits - The proposed research will constitute a breakthrough in solution of problems related to efficient
Enhanced Verification Test Suite for Physics Simulation Codes
Energy Technology Data Exchange (ETDEWEB)
Kamm, J R; Brock, J S; Brandon, S T; Cotrell, D L; Johnson, B; Knupp, P; Rider, W; Trucano, T; Weirs, V G
2008-10-10
This document discusses problems with which to augment, in quantity and in quality, the existing tri-laboratory suite of verification problems used by Los Alamos National Laboratory (LANL), Lawrence Livermore National Laboratory (LLNL), and Sandia National Laboratories (SNL). The purpose of verification analysis is demonstrate whether the numerical results of the discretization algorithms in physics and engineering simulation codes provide correct solutions of the corresponding continuum equations. The key points of this document are: (1) Verification deals with mathematical correctness of the numerical algorithms in a code, while validation deals with physical correctness of a simulation in a regime of interest. This document is about verification. (2) The current seven-problem Tri-Laboratory Verification Test Suite, which has been used for approximately five years at the DOE WP laboratories, is limited. (3) Both the methodology for and technology used in verification analysis have evolved and been improved since the original test suite was proposed. (4) The proposed test problems are in three basic areas: (a) Hydrodynamics; (b) Transport processes; and (c) Dynamic strength-of-materials. (5) For several of the proposed problems we provide a 'strong sense verification benchmark', consisting of (i) a clear mathematical statement of the problem with sufficient information to run a computer simulation, (ii) an explanation of how the code result and benchmark solution are to be evaluated, and (iii) a description of the acceptance criterion for simulation code results. (6) It is proposed that the set of verification test problems with which any particular code be evaluated include some of the problems described in this document. Analysis of the proposed verification test problems constitutes part of a necessary--but not sufficient--step that builds confidence in physics and engineering simulation codes. More complicated test cases, including physics models of
International Nuclear Information System (INIS)
Fernandes Filho, T.L.
1982-11-01
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.) [pt
Theoretical Atomic Physics code development IV: LINES, A code for computing atomic line spectra
International Nuclear Information System (INIS)
Abdallah, J. Jr.; Clark, R.E.H.
1988-12-01
A new computer program, LINES, has been developed for simulating atomic line emission and absorption spectra using the accurate fine structure energy levels and transition strengths calculated by the (CATS) Cowan Atomic Structure code. Population distributions for the ion stages are obtained in LINES by using the Local Thermodynamic Equilibrium (LTE) model. LINES is also useful for displaying the pertinent atomic data generated by CATS. This report describes the use of LINES. Both CATS and LINES are part of the Theoretical Atomic PhysicS (TAPS) code development effort at Los Alamos. 11 refs., 9 figs., 1 tab
HYDRASTAR - a code for stochastic simulation of groundwater flow
International Nuclear Information System (INIS)
Norman, S.
1992-05-01
The computer code HYDRASTAR was developed as a tool for groundwater flow and transport simulations in the SKB 91 safety analysis project. Its conceptual ideas can be traced back to a report by Shlomo Neuman in 1988, see the reference section. The main idea of the code is the treatment of the rock as a stochastic continuum which separates it from the deterministic methods previously employed by SKB and also from the discrete fracture models. The current report is a comprehensive description of HYDRASTAR including such topics as regularization or upscaling of a hydraulic conductivity field, unconditional and conditional simulation of stochastic processes, numerical solvers for the hydrology and streamline equations and finally some proposals for future developments
Application of parallel computing techniques to a large-scale reservoir simulation
International Nuclear Information System (INIS)
Zhang, Keni; Wu, Yu-Shu; Ding, Chris; Pruess, Karsten
2001-01-01
Even with the continual advances made in both computational algorithms and computer hardware used in reservoir modeling studies, large-scale simulation of fluid and heat flow in heterogeneous reservoirs remains a challenge. The problem commonly arises from intensive computational requirement for detailed modeling investigations of real-world reservoirs. This paper presents the application of a massive parallel-computing version of the TOUGH2 code developed for performing large-scale field simulations. As an application example, the parallelized TOUGH2 code is applied to develop a three-dimensional unsaturated-zone numerical model simulating flow of moisture, gas, and heat in the unsaturated zone of Yucca Mountain, Nevada, a potential repository for high-level radioactive waste. The modeling approach employs refined spatial discretization to represent the heterogeneous fractured tuffs of the system, using more than a million 3-D gridblocks. The problem of two-phase flow and heat transfer within the model domain leads to a total of 3,226,566 linear equations to be solved per Newton iteration. The simulation is conducted on a Cray T3E-900, a distributed-memory massively parallel computer. Simulation results indicate that the parallel computing technique, as implemented in the TOUGH2 code, is very efficient. The reliability and accuracy of the model results have been demonstrated by comparing them to those of small-scale (coarse-grid) models. These comparisons show that simulation results obtained with the refined grid provide more detailed predictions of the future flow conditions at the site, aiding in the assessment of proposed repository performance
Development of a computer code for Dalat research reactor transient analysis
International Nuclear Information System (INIS)
Le Vinh Vinh; Nguyen Thai Sinh; Huynh Ton Nghiem; Luong Ba Vien; Pham Van Lam; Nguyen Kien Cuong
2003-01-01
DRSIM (Dalat Reactor SIMulation) computer code has been developed for Dalat reactor transient analysis. It is basically a coupled neutronics-hydrodynamics-heat transfer code employing point kinetics, one dimensional hydrodynamics and one dimensional heat transfer. The work was financed by VAEC and DNRI in the framework of institutional R and D programme. Some transient problems related to reactivity and loss of coolant flow was carried out by DRSIM using temperature and void coefficients calculated by WIMS and HEXNOD2D codes. (author)
Graphical Visualization on Computational Simulation Using Shared Memory
International Nuclear Information System (INIS)
Lima, A B; Correa, Eberth
2014-01-01
The Shared Memory technique is a powerful tool for parallelizing computer codes. In particular it can be used to visualize the results ''on the fly'' without stop running the simulation. In this presentation we discuss and show how to use the technique conjugated with a visualization code using openGL
V.S.O.P. (99/05) computer code system
International Nuclear Information System (INIS)
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 (∼65000 Fortran statements). (orig.)
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)
Marrel, A.
2008-01-01
In the studies of environmental transfer and risk assessment, numerical models are used to simulate, understand and predict the transfer of pollutant. These computer codes can depend on a high number of uncertain input parameters (geophysical variables, chemical parameters, etc.) and can be often too computer time expensive. To conduct uncertainty propagation studies and to measure the importance of each input on the response variability, the computer code has to be approximated by a meta model which is build on an acceptable number of simulations of the code and requires a negligible calculation time. We focused our research work on the use of Gaussian process meta model to make the sensitivity analysis of the code. We proposed a methodology with estimation and input selection procedures in order to build the meta model in the case of a high number of inputs and with few simulations available. Then, we compared two approaches to compute the sensitivity indices with the meta model and proposed an algorithm to build prediction intervals for these indices. Afterwards, we were interested in the choice of the code simulations. We studied the influence of different sampling strategies on the predictiveness of the Gaussian process meta model. Finally, we extended our statistical tools to a functional output of a computer code. We combined a decomposition on a wavelet basis with the Gaussian process modelling before computing the functional sensitivity indices. All the tools and statistical methodologies that we developed were applied to the real case of a complex hydrogeological computer code, simulating radionuclide transport in groundwater. (author) [fr
The first accident simulation of Angra-1 power plant using the ALMOD computer code
International Nuclear Information System (INIS)
Camargo, C.T.M.
1981-01-01
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. (E.G.) [pt
International Nuclear Information System (INIS)
Silva Filho, E.
1986-01-01
The computer code RELAP5/MOD1 has been utilized to investigate the thermal-hydraulic behaviour of a standard 1300 Mwe pressurized water reactor plant of the KWU design during a station blackout and during a loss-of-coolant accident involving 2% break in the cross-sectional area the cold leg in one of the four loops and located between the pump and the reactor pressure vessel. During the simulations the reactor scram system and the emergency coolant system were considered inactive. (Author) [pt
Verification of thermal-hydraulic computer codes against standard problems for WWER reflooding
International Nuclear Information System (INIS)
Alexander D Efanov; Vladimir N Vinogradov; Victor V Sergeev; Oleg A Sudnitsyn
2005-01-01
Full text of publication follows: The computational assessment of reactor core components behavior under accident conditions is impossible without knowledge of the thermal-hydraulic processes occurring in this case. The adequacy of the results obtained using the computer codes to the real processes is verified by carrying out a number of standard problems. In 2000-2003, the fulfillment of three Russian standard problems on WWER core reflooding was arranged using the experiments on full-height electrically heated WWER 37-rod bundle model cooldown in regimes of bottom (SP-1), top (SP-2) and combined (SP-3) reflooding. The representatives from the eight MINATOM's organizations took part in this work, in the course of which the 'blind' and posttest calculations were performed using various versions of the RELAP5, ATHLET, CATHARE, COBRA-TF, TRAP, KORSAR computer codes. The paper presents a brief description of the test facility, test section, test scenarios and conditions as well as the basic results of computational analysis of the experiments. The analysis of the test data revealed a significantly non-one-dimensional nature of cooldown and rewetting of heater rods heated up to a high temperature in a model bundle. This was most pronounced at top and combined reflooding. The verification of the model reflooding computer codes showed that most of computer codes fairly predict the peak rod temperature and the time of bundle cooldown. The exception is provided by the results of calculations with the ATHLET and CATHARE codes. The nature and rate of rewetting front advance in the lower half of the bundle are fairly predicted practically by all computer codes. The disagreement between the calculations and experimental results for the upper half of the bundle is caused by the difficulties of computational simulation of multidimensional effects by 1-D computer codes. In this regard, a quasi-two-dimensional computer code COBRA-TF offers certain advantages. Overall, the closest
Coding considerations for standalone molecular dynamics simulations of atomistic structures
Ocaya, R. O.; Terblans, J. J.
2017-10-01
The laws of Newtonian mechanics allow ab-initio molecular dynamics to model and simulate particle trajectories in material science by defining a differentiable potential function. This paper discusses some considerations for the coding of ab-initio programs for simulation on a standalone computer and illustrates the approach by C language codes in the context of embedded metallic atoms in the face-centred cubic structure. The algorithms use velocity-time integration to determine particle parameter evolution for up to several thousands of particles in a thermodynamical ensemble. Such functions are reusable and can be placed in a redistributable header library file. While there are both commercial and free packages available, their heuristic nature prevents dissection. In addition, developing own codes has the obvious advantage of teaching techniques applicable to new problems.
Simulation of international standard problem no. 44 open tests using Melcor computer code
International Nuclear Information System (INIS)
Song, Y.M.; Cho, S.W.
2001-01-01
MELCOR 1.8.4 code has been employed to simulate the KAEVER test series of K123/K148/K186/K188 that were proposed as open experiments of International Standard Problem No.44 by OECD-CSNI. The main purpose of this study is to evaluate the accuracy of the MELCOR aerosol model which calculates the aerosol distribution and settlement in a containment. For this, thermal hydraulic conditions are simulated first for the whole test period and then the behavior of hygroscopic CsOH/CsI and unsoluble Ag aerosols, which are predominant activity carriers in a release into the containment, is compared between the experimental results and the code predictions. The calculation results of vessel atmospheric concentration show a good simulation for dry aerosol but show large difference for wet aerosol due to a data mismatch in vessel humidity and the hygroscopicity. (authors)
Python Radiative Transfer Emission code (PyRaTE): non-LTE spectral lines simulations
Tritsis, A.; Yorke, H.; Tassis, K.
2018-05-01
We describe PyRaTE, a new, non-local thermodynamic equilibrium (non-LTE) line radiative transfer code developed specifically for post-processing astrochemical simulations. Population densities are estimated using the escape probability method. When computing the escape probability, the optical depth is calculated towards all directions with density, molecular abundance, temperature and velocity variations all taken into account. A very easy-to-use interface, capable of importing data from simulations outputs performed with all major astrophysical codes, is also developed. The code is written in PYTHON using an "embarrassingly parallel" strategy and can handle all geometries and projection angles. We benchmark the code by comparing our results with those from RADEX (van der Tak et al. 2007) and against analytical solutions and present case studies using hydrochemical simulations. The code will be released for public use.
TEMPEST code modifications and testing for erosion-resisting sludge simulations
International Nuclear Information System (INIS)
Onishi, Y.; Trent, D.S.
1998-01-01
The TEMPEST computer code has been used to address many waste retrieval operational and safety questions regarding waste mobilization, mixing, and gas retention. Because the amount of sludge retrieved from the tank is directly related to the sludge yield strength and the shear stress acting upon it, it is important to incorporate the sludge yield strength into simulations of erosion-resisting tank waste retrieval operations. This report describes current efforts to modify the TEMPEST code to simulate pump jet mixing of erosion-resisting tank wastes and the models used to test for erosion of waste sludge with yield strength. Test results for solid deposition and diluent/slurry jet injection into sludge layers in simplified tank conditions show that the modified TEMPEST code has a basic ability to simulate both the mobility and immobility of the sludges with yield strength. Further testing, modification, calibration, and verification of the sludge mobilization/immobilization model are planned using erosion data as they apply to waste tank sludges
Use of computer codes for system reliability analysis
International Nuclear Information System (INIS)
Sabek, M.; Gaafar, M.; Poucet, A.
1988-01-01
This paper gives a collective summary of the studies performed at the JRC, ISPRA on the use of computer codes for complex systems analysis. The computer codes dealt with are: CAFTS-SALP software package, FRANTIC, FTAP, computer code package RALLY, and BOUNDS codes. Two reference study cases were executed by each code. The results obtained logic/probabilistic analysis as well as computation time are compared
FIRAC: a computer code to predict fire-accident effects in nuclear facilities
International Nuclear Information System (INIS)
Bolstad, J.W.; Krause, F.R.; Tang, P.K.; Andrae, R.W.; Martin, R.A.; Gregory, W.S.
1983-01-01
FIRAC is a medium-sized computer code designed to predict fire-induced flows, temperatures, and material transport within the ventilating systems and other airflow pathways in nuclear-related facilities. The code is designed to analyze the behavior of interconnected networks of rooms and typical ventilation system components. This code is one in a family of computer codes that is designed to provide improved methods of safety analysis for the nuclear industry. The structure of this code closely follows that of the previously developed TVENT and EVENT codes. Because a lumped-parameter formulation is used, this code is particularly suitable for calculating the effects of fires in the far field (that is, in regions removed from the fire compartment), where the fire may be represented parametrically. However, a fire compartment model to simulate conditions in the enclosure is included. This model provides transport source terms to the ventilation system that can affect its operation and in turn affect the fire
International Nuclear Information System (INIS)
Raymond, P.; Caruge, D.; Paik, H.J.
1994-01-01
The French CEA has recently developed a set of new computer codes for reactor physics computations called the Saphir system which includes CRONOS-2, a three-dimensional neutronic code, FLICA-4, a three-dimensional core thermal hydraulic code, and FLICA-S, a primary loops thermal-hydraulic transient computation code, which are coupled and applied to analyze a severe reactivity accident induced by a thermal hydraulic transient: the Steamline Break accident for a pressurized water reactor until soluble boron begins to accumulate in the core. The coupling of these codes has proved to be numerically stable. 15 figs., 7 refs
International Nuclear Information System (INIS)
Gupta, S.C.; Sikka, S.K.; Chidambaram, R.
1979-01-01
An account is given of a one-dimensional spherical symmetric computer code for the numerical simulation of the effects of peaceful underground nuclear explosions in rocks (OCENER). In the code, the nature of the stress field and response of the medium to this field are modelled numerically by finite difference form of the laws of continuum mechanics and the constitutive relations of the rock medium in which the detonation occurs. It enables to approximate well the cavity growth and fracturing of the surrounding rock for contained explosions and the events upto the time the spherical symmetry is valid for cratering-type explosions. (auth.)
CHOLLA: A NEW MASSIVELY PARALLEL HYDRODYNAMICS CODE FOR ASTROPHYSICAL SIMULATION
International Nuclear Information System (INIS)
Schneider, Evan E.; Robertson, Brant E.
2015-01-01
We present Computational Hydrodynamics On ParaLLel Architectures (Cholla ), a new three-dimensional hydrodynamics code that harnesses the power of graphics processing units (GPUs) to accelerate astrophysical simulations. Cholla models the Euler equations on a static mesh using state-of-the-art techniques, including the unsplit Corner Transport Upwind algorithm, a variety of exact and approximate Riemann solvers, and multiple spatial reconstruction techniques including the piecewise parabolic method (PPM). Using GPUs, Cholla evolves the fluid properties of thousands of cells simultaneously and can update over 10 million cells per GPU-second while using an exact Riemann solver and PPM reconstruction. Owing to the massively parallel architecture of GPUs and the design of the Cholla code, astrophysical simulations with physically interesting grid resolutions (≳256 3 ) can easily be computed on a single device. We use the Message Passing Interface library to extend calculations onto multiple devices and demonstrate nearly ideal scaling beyond 64 GPUs. A suite of test problems highlights the physical accuracy of our modeling and provides a useful comparison to other codes. We then use Cholla to simulate the interaction of a shock wave with a gas cloud in the interstellar medium, showing that the evolution of the cloud is highly dependent on its density structure. We reconcile the computed mixing time of a turbulent cloud with a realistic density distribution destroyed by a strong shock with the existing analytic theory for spherical cloud destruction by describing the system in terms of its median gas density
CHOLLA: A NEW MASSIVELY PARALLEL HYDRODYNAMICS CODE FOR ASTROPHYSICAL SIMULATION
Energy Technology Data Exchange (ETDEWEB)
Schneider, Evan E.; Robertson, Brant E. [Steward Observatory, University of Arizona, 933 North Cherry Avenue, Tucson, AZ 85721 (United States)
2015-04-15
We present Computational Hydrodynamics On ParaLLel Architectures (Cholla ), a new three-dimensional hydrodynamics code that harnesses the power of graphics processing units (GPUs) to accelerate astrophysical simulations. Cholla models the Euler equations on a static mesh using state-of-the-art techniques, including the unsplit Corner Transport Upwind algorithm, a variety of exact and approximate Riemann solvers, and multiple spatial reconstruction techniques including the piecewise parabolic method (PPM). Using GPUs, Cholla evolves the fluid properties of thousands of cells simultaneously and can update over 10 million cells per GPU-second while using an exact Riemann solver and PPM reconstruction. Owing to the massively parallel architecture of GPUs and the design of the Cholla code, astrophysical simulations with physically interesting grid resolutions (≳256{sup 3}) can easily be computed on a single device. We use the Message Passing Interface library to extend calculations onto multiple devices and demonstrate nearly ideal scaling beyond 64 GPUs. A suite of test problems highlights the physical accuracy of our modeling and provides a useful comparison to other codes. We then use Cholla to simulate the interaction of a shock wave with a gas cloud in the interstellar medium, showing that the evolution of the cloud is highly dependent on its density structure. We reconcile the computed mixing time of a turbulent cloud with a realistic density distribution destroyed by a strong shock with the existing analytic theory for spherical cloud destruction by describing the system in terms of its median gas density.
Computing element evolution towards Exascale and its impact on legacy simulation codes
International Nuclear Information System (INIS)
Colin de Verdiere, Guillaume J.L.
2015-01-01
In the light of the current race towards the Exascale, this article highlights the main features of the forthcoming computing elements that will be at the core of next generations of supercomputers. The market analysis, underlying this work, shows that computers are facing a major evolution in terms of architecture. As a consequence, it is important to understand the impacts of those evolutions on legacy codes or programming methods. The problems of dissipated power and memory access are discussed and will lead to a vision of what should be an exascale system. To survive, programming languages had to respond to the hardware evolutions either by evolving or with the creation of new ones. From the previous elements, we elaborate why vectorization, multithreading, data locality awareness and hybrid programming will be the key to reach the exascale, implying that it is time to start rewriting codes. (orig.)
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.
Computational simulation of natural circulation and rewetting experiments using the TRAC/PF1 code
International Nuclear Information System (INIS)
Silva, J.D. da.
1994-05-01
In this work the TRAC code was used to simulate experiments of natural circulation performed in the first Brazilian integral test facility at (COPESP), Sao Paulo and a rewetting experiment in a single tube test section carried out at CDTN, Belo Horizonte, Brazil. In the first simulation the loop behavior in two transient conditions with different thermal power, namely 20 k W and 120 k W, was verified in the second one the quench front propagation, the liquid mass collected in the carry over measuring tube and the wall temperature at different elevations during the flooding experiment was measured. A comparative analysis, for code consistency, shows a good agreement between the code results and experimental data, except for the quench from velocity. (author). 15 refs, 19 figs, 12 tabs
Translation of ARAC computer codes
International Nuclear Information System (INIS)
Takahashi, Kunio; Chino, Masamichi; Honma, Toshimitsu; Ishikawa, Hirohiko; Kai, Michiaki; Imai, Kazuhiko; Asai, Kiyoshi
1982-05-01
In 1981 we have translated the famous MATHEW, ADPIC and their auxiliary computer codes for CDC 7600 computer version to FACOM M-200's. The codes consist of a part of the Atmospheric Release Advisory Capability (ARAC) system of Lawrence Livermore National Laboratory (LLNL). The MATHEW is a code for three-dimensional wind field analysis. Using observed data, it calculates the mass-consistent wind field of grid cells by a variational method. The ADPIC is a code for three-dimensional concentration prediction of gases and particulates released to the atmosphere. It calculates concentrations in grid cells by the particle-in-cell method. They are written in LLLTRAN, i.e., LLNL Fortran language and are implemented on the CDC 7600 computers of LLNL. In this report, i) the computational methods of the MATHEW/ADPIC and their auxiliary codes, ii) comparisons of the calculated results with our JAERI particle-in-cell, and gaussian plume models, iii) translation procedures from the CDC version to FACOM M-200's, are described. Under the permission of LLNL G-Division, this report is published to keep the track of the translation procedures and to serve our JAERI researchers for comparisons and references of their works. (author)
The Premar Code for the Monte Carlo Simulation of Radiation Transport In the Atmosphere
International Nuclear Information System (INIS)
Cupini, E.; Borgia, M.G.; Premuda, M.
1997-03-01
The Montecarlo code PREMAR is described, which allows the user to simulate the radiation transport in the atmosphere, in the ultraviolet-infrared frequency interval. A plan multilayer geometry is at present foreseen by the code, witch albedo possibility at the lower boundary surface. For a given monochromatic point source, the main quantities computed by the code are the absorption spatial distributions of aerosol and molecules, together with the related atmospheric transmittances. Moreover, simulation of of Lidar experiments are foreseen by the code, the source and telescope fields of view being assigned. To build-up the appropriate probability distributions, an input data library is assumed to be read by the code. For this purpose the radiance-transmittance LOWTRAN-7 code has been conveniently adapted as a source of the library so as to exploit the richness of information of the code for a large variety of atmospheric simulations. Results of applications of the PREMAR code are finally presented, with special reference to simulations of Lidar system and radiometer experiments carried out at the Brasimone ENEA Centre by the Environment Department
SASSYS-1 computer code verification with EBR-II test data
International Nuclear Information System (INIS)
Warinner, D.K.; Dunn, F.E.
1985-01-01
The EBR-II natural circulation experiment, XX08 Test 8A, is simulated with the SASSYS-1 computer code and the results for the latter are compared with published data taken during the transient at selected points in the core. The SASSYS-1 results provide transient temperature and flow responses for all points of interest simultaneously during one run, once such basic parameters as pipe sizes, initial core flows, and elevations are specified. The SASSYS-1 simulation results for the EBR-II experiment XX08 Test 8A, conducted in March 1979, are within the published plant data uncertainties and, thereby, serve as a partial verification/validation of the SASSYS-1 code
Parallelization of a beam dynamics code and first large scale radio frequency quadrupole simulations
Directory of Open Access Journals (Sweden)
J. Xu
2007-01-01
Full Text Available The design and operation support of hadron (proton and heavy-ion linear accelerators require substantial use of beam dynamics simulation tools. The beam dynamics code TRACK has been originally developed at Argonne National Laboratory (ANL to fulfill the special requirements of the rare isotope accelerator (RIA accelerator systems. From the beginning, the code has been developed to make it useful in the three stages of a linear accelerator project, namely, the design, commissioning, and operation of the machine. To realize this concept, the code has unique features such as end-to-end simulations from the ion source to the final beam destination and automatic procedures for tuning of a multiple charge state heavy-ion beam. The TRACK code has become a general beam dynamics code for hadron linacs and has found wide applications worldwide. Until recently, the code has remained serial except for a simple parallelization used for the simulation of multiple seeds to study the machine errors. To speed up computation, the TRACK Poisson solver has been parallelized. This paper discusses different parallel models for solving the Poisson equation with the primary goal to extend the scalability of the code onto 1024 and more processors of the new generation of supercomputers known as BlueGene (BG/L. Domain decomposition techniques have been adapted and incorporated into the parallel version of the TRACK code. To demonstrate the new capabilities of the parallelized TRACK code, the dynamics of a 45 mA proton beam represented by 10^{8} particles has been simulated through the 325 MHz radio frequency quadrupole and initial accelerator section of the proposed FNAL proton driver. The results show the benefits and advantages of large-scale parallel computing in beam dynamics simulations.
Computer simulation of high energy displacement cascades
International Nuclear Information System (INIS)
Heinisch, H.L.
1990-01-01
A methodology developed for modeling many aspects of high energy displacement cascades with molecular level computer simulations is reviewed. The initial damage state is modeled in the binary collision approximation (using the MARLOWE computer code), and the subsequent disposition of the defects within a cascade is modeled with a Monte Carlo annealing simulation (the ALSOME code). There are few adjustable parameters, and none are set to physically unreasonable values. The basic configurations of the simulated high energy cascades in copper, i.e., the number, size and shape of damage regions, compare well with observations, as do the measured numbers of residual defects and the fractions of freely migrating defects. The success of these simulations is somewhat remarkable, given the relatively simple models of defects and their interactions that are employed. The reason for this success is that the behavior of the defects is very strongly influenced by their initial spatial distributions, which the binary collision approximation adequately models. The MARLOWE/ALSOME system, with input from molecular dynamics and experiments, provides a framework for investigating the influence of high energy cascades on microstructure evolution. (author)
High performance simulation for the Silva project using the tera computer
International Nuclear Information System (INIS)
Bergeaud, V.; La Hargue, J.P.; Mougery, F.; Boulet, M.; Scheurer, B.; Le Fur, J.F.; Comte, M.; Benisti, D.; Lamare, J. de; Petit, A.
2003-01-01
In the context of the SILVA Project (Atomic Vapor Laser Isotope Separation), numerical simulation of the plant scale propagation of laser beams through uranium vapour was a great challenge. The PRODIGE code has been developed to achieve this goal. Here we focus on the task of achieving high performance simulation on the TERA computer. We describe the main issues for optimizing the parallelization of the PRODIGE code on TERA. Thus, we discuss advantages and drawbacks of the implemented diagonal parallelization scheme. As a consequence, it has been found fruitful to fit out the code in three aspects: memory allocation, MPI communications and interconnection network bandwidth usage. We stress out the interest of MPI/IO in this context and the benefit obtained for production computations on TERA. Finally, we shall illustrate our developments. We indicate some performance measurements reflecting the good parallelization properties of PRODIGE on the TERA computer. The code is currently used for demonstrating the feasibility of the laser propagation at a plant enrichment level and for preparing the 2003 Menphis experiment. We conclude by emphasizing the contribution of high performance TERA simulation to the project. (authors)
High performance simulation for the Silva project using the tera computer
Energy Technology Data Exchange (ETDEWEB)
Bergeaud, V.; La Hargue, J.P.; Mougery, F. [CS Communication and Systemes, 92 - Clamart (France); Boulet, M.; Scheurer, B. [CEA Bruyeres-le-Chatel, 91 - Bruyeres-le-Chatel (France); Le Fur, J.F.; Comte, M.; Benisti, D.; Lamare, J. de; Petit, A. [CEA Saclay, 91 - Gif sur Yvette (France)
2003-07-01
In the context of the SILVA Project (Atomic Vapor Laser Isotope Separation), numerical simulation of the plant scale propagation of laser beams through uranium vapour was a great challenge. The PRODIGE code has been developed to achieve this goal. Here we focus on the task of achieving high performance simulation on the TERA computer. We describe the main issues for optimizing the parallelization of the PRODIGE code on TERA. Thus, we discuss advantages and drawbacks of the implemented diagonal parallelization scheme. As a consequence, it has been found fruitful to fit out the code in three aspects: memory allocation, MPI communications and interconnection network bandwidth usage. We stress out the interest of MPI/IO in this context and the benefit obtained for production computations on TERA. Finally, we shall illustrate our developments. We indicate some performance measurements reflecting the good parallelization properties of PRODIGE on the TERA computer. The code is currently used for demonstrating the feasibility of the laser propagation at a plant enrichment level and for preparing the 2003 Menphis experiment. We conclude by emphasizing the contribution of high performance TERA simulation to the project. (authors)
Computer code development plant for SMART design
International Nuclear Information System (INIS)
Bae, Kyoo Hwan; Choi, S.; Cho, B.H.; Kim, K.K.; Lee, J.C.; Kim, J.P.; Kim, J.H.; Chung, M.; Kang, D.J.; Chang, M.H.
1999-03-01
In accordance with the localization plan for the nuclear reactor design driven since the middle of 1980s, various computer codes have been transferred into the korea nuclear industry through the technical transfer program from the worldwide major pressurized water reactor supplier or through the international code development program. These computer codes have been successfully utilized in reactor and reload core design works. As the results, design- related technologies have been satisfactorily accumulated. However, the activities for the native code development activities to substitute the some important computer codes of which usages are limited by the original technique owners have been carried out rather poorly. Thus, it is most preferentially required to secure the native techniques on the computer code package and analysis methodology in order to establish the capability required for the independent design of our own model of reactor. Moreover, differently from the large capacity loop-type commercial reactors, SMART (SYSTEM-integrated Modular Advanced ReacTor) design adopts a single reactor pressure vessel containing the major primary components and has peculiar design characteristics such as self-controlled gas pressurizer, helical steam generator, passive residual heat removal system, etc. Considering those peculiar design characteristics for SMART, part of design can be performed with the computer codes used for the loop-type commercial reactor design. However, most of those computer codes are not directly applicable to the design of an integral reactor such as SMART. Thus, they should be modified to deal with the peculiar design characteristics of SMART. In addition to the modification efforts, various codes should be developed in several design area. Furthermore, modified or newly developed codes should be verified their reliability through the benchmarking or the test for the object design. Thus, it is necessary to proceed the design according to the
Computer code development plant for SMART design
Energy Technology Data Exchange (ETDEWEB)
Bae, Kyoo Hwan; Choi, S.; Cho, B.H.; Kim, K.K.; Lee, J.C.; Kim, J.P.; Kim, J.H.; Chung, M.; Kang, D.J.; Chang, M.H
1999-03-01
In accordance with the localization plan for the nuclear reactor design driven since the middle of 1980s, various computer codes have been transferred into the korea nuclear industry through the technical transfer program from the worldwide major pressurized water reactor supplier or through the international code development program. These computer codes have been successfully utilized in reactor and reload core design works. As the results, design- related technologies have been satisfactorily accumulated. However, the activities for the native code development activities to substitute the some important computer codes of which usages are limited by the original technique owners have been carried out rather poorly. Thus, it is most preferentially required to secure the native techniques on the computer code package and analysis methodology in order to establish the capability required for the independent design of our own model of reactor. Moreover, differently from the large capacity loop-type commercial reactors, SMART (SYSTEM-integrated Modular Advanced ReacTor) design adopts a single reactor pressure vessel containing the major primary components and has peculiar design characteristics such as self-controlled gas pressurizer, helical steam generator, passive residual heat removal system, etc. Considering those peculiar design characteristics for SMART, part of design can be performed with the computer codes used for the loop-type commercial reactor design. However, most of those computer codes are not directly applicable to the design of an integral reactor such as SMART. Thus, they should be modified to deal with the peculiar design characteristics of SMART. In addition to the modification efforts, various codes should be developed in several design area. Furthermore, modified or newly developed codes should be verified their reliability through the benchmarking or the test for the object design. Thus, it is necessary to proceed the design according to the
Srna-Monte Carlo codes for proton transport simulation in combined and voxelized geometries
Ilic, R D; Stankovic, S J
2002-01-01
This paper describes new Monte Carlo codes for proton transport simulations in complex geometrical forms and in materials of different composition. The SRNA codes were developed for three dimensional (3D) dose distribution calculation in proton therapy and dosimetry. The model of these codes is based on the theory of proton multiple scattering and a simple model of compound nucleus decay. The developed package consists of two codes: SRNA-2KG and SRNA-VOX. The first code simulates proton transport in combined geometry that can be described by planes and second order surfaces. The second one uses the voxelized geometry of material zones and is specifically adopted for the application of patient computer tomography data. Transition probabilities for both codes are given by the SRNADAT program. In this paper, we will present the models and algorithms of our programs, as well as the results of the numerical experiments we have carried out applying them, along with the results of proton transport simulation obtaine...
Computer-aided software understanding systems to enhance confidence of scientific codes
International Nuclear Information System (INIS)
Sheng, G.; Oeren, T.I.
1991-01-01
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.)
Fast code for Monte Carlo simulations
International Nuclear Information System (INIS)
Oliveira, P.M.C. de; Penna, T.J.P.
1988-01-01
A computer code to generate the dynamic evolution of the Ising model on a square lattice, following the Metropolis algorithm is presented. The computer time consumption is reduced by a factor of 8 when one compares our code with traditional multiple spin codes. The memory allocation size is also reduced by a factor of 4. The code is easily generalizable for other lattices and models. (author) [pt
International Nuclear Information System (INIS)
Shirakawa, Toshihiko; Hatanaka, Koichiro
2001-11-01
In order to document a basic manual about input data, output data, execution of computer code on groundwater flow and radionuclide transport calculation in heterogeneous porous rock, we investigated the theoretical background about geostatistical computer codes and the user's manual for the computer code on groundwater flow and radionuclide transport which calculates water flow in three dimension, the path of moving radionuclide, and one dimensional radionuclide migration. In this report, based on above investigation we describe the geostatistical background about simulating heterogeneous permeability field. And we describe construction of files, input and output data, a example of calculating of the programs which simulates heterogeneous permeability field, and calculates groundwater flow and radionuclide transport. Therefore, we can document a manual by investigating the theoretical background about geostatistical computer codes and the user's manual for the computer code on groundwater flow and radionuclide transport calculation. And we can model heterogeneous porous rock and analyze groundwater flow and radionuclide transport by utilizing the information from this report. (author)
Polymer Composites Corrosive Degradation: A Computational Simulation
Chamis, Christos C.; Minnetyan, Levon
2007-01-01
A computational simulation of polymer composites corrosive durability is presented. The corrosive environment is assumed to manage the polymer composite degradation on a ply-by-ply basis. The degradation is correlated with a measured pH factor and is represented by voids, temperature and moisture which vary parabolically for voids and linearly for temperature and moisture through the laminate thickness. The simulation is performed by a computational composite mechanics computer code which includes micro, macro, combined stress failure and laminate theories. This accounts for starting the simulation from constitutive material properties and up to the laminate scale which exposes the laminate to the corrosive environment. Results obtained for one laminate indicate that the ply-by-ply degradation degrades the laminate to the last one or the last several plies. Results also demonstrate that the simulation is applicable to other polymer composite systems as well.
The advanced computational testing and simulation toolkit (ACTS)
International Nuclear Information System (INIS)
Drummond, L.A.; Marques, O.
2002-01-01
During the past decades there has been a continuous growth in the number of physical and societal problems that have been successfully studied and solved by means of computational modeling and simulation. Distinctively, a number of these are important scientific problems ranging in scale from the atomic to the cosmic. For example, ionization is a phenomenon as ubiquitous in modern society as the glow of fluorescent lights and the etching on silicon computer chips; but it was not until 1999 that researchers finally achieved a complete numerical solution to the simplest example of ionization, the collision of a hydrogen atom with an electron. On the opposite scale, cosmologists have long wondered whether the expansion of the Universe, which began with the Big Bang, would ever reverse itself, ending the Universe in a Big Crunch. In 2000, analysis of new measurements of the cosmic microwave background radiation showed that the geometry of the Universe is flat, and thus the Universe will continue expanding forever. Both of these discoveries depended on high performance computer simulations that utilized computational tools included in the Advanced Computational Testing and Simulation (ACTS) Toolkit. The ACTS Toolkit is an umbrella project that brought together a number of general purpose computational tool development projects funded and supported by the U.S. Department of Energy (DOE). These tools, which have been developed independently, mainly at DOE laboratories, make it easier for scientific code developers to write high performance applications for parallel computers. They tackle a number of computational issues that are common to a large number of scientific applications, mainly implementation of numerical algorithms, and support for code development, execution and optimization. The ACTS Toolkit Project enables the use of these tools by a much wider community of computational scientists, and promotes code portability, reusability, reduction of duplicate efforts
The advanced computational testing and simulation toolkit (ACTS)
Energy Technology Data Exchange (ETDEWEB)
Drummond, L.A.; Marques, O.
2002-05-21
During the past decades there has been a continuous growth in the number of physical and societal problems that have been successfully studied and solved by means of computational modeling and simulation. Distinctively, a number of these are important scientific problems ranging in scale from the atomic to the cosmic. For example, ionization is a phenomenon as ubiquitous in modern society as the glow of fluorescent lights and the etching on silicon computer chips; but it was not until 1999 that researchers finally achieved a complete numerical solution to the simplest example of ionization, the collision of a hydrogen atom with an electron. On the opposite scale, cosmologists have long wondered whether the expansion of the Universe, which began with the Big Bang, would ever reverse itself, ending the Universe in a Big Crunch. In 2000, analysis of new measurements of the cosmic microwave background radiation showed that the geometry of the Universe is flat, and thus the Universe will continue expanding forever. Both of these discoveries depended on high performance computer simulations that utilized computational tools included in the Advanced Computational Testing and Simulation (ACTS) Toolkit. The ACTS Toolkit is an umbrella project that brought together a number of general purpose computational tool development projects funded and supported by the U.S. Department of Energy (DOE). These tools, which have been developed independently, mainly at DOE laboratories, make it easier for scientific code developers to write high performance applications for parallel computers. They tackle a number of computational issues that are common to a large number of scientific applications, mainly implementation of numerical algorithms, and support for code development, execution and optimization. The ACTS Toolkit Project enables the use of these tools by a much wider community of computational scientists, and promotes code portability, reusability, reduction of duplicate efforts
Advanced computers and simulation
International Nuclear Information System (INIS)
Ryne, R.D.
1993-01-01
Accelerator physicists today have access to computers that are far more powerful than those available just 10 years ago. In the early 1980's, desktop workstations performed less one million floating point operations per second (Mflops), and the realized performance of vector supercomputers was at best a few hundred Mflops. Today vector processing is available on the desktop, providing researchers with performance approaching 100 Mflops at a price that is measured in thousands of dollars. Furthermore, advances in Massively Parallel Processors (MPP) have made performance of over 10 gigaflops a reality, and around mid-decade MPPs are expected to be capable of teraflops performance. Along with advances in MPP hardware, researchers have also made significant progress in developing algorithms and software for MPPS. These changes have had, and will continue to have, a significant impact on the work of computational accelerator physicists. Now, instead of running particle simulations with just a few thousand particles, we can perform desktop simulations with tens of thousands of simulation particles, and calculations with well over 1 million particles are being performed on MPPs. In the area of computational electromagnetics, simulations that used to be performed only on vector supercomputers now run in several hours on desktop workstations, and researchers are hoping to perform simulations with over one billion mesh points on future MPPs. In this paper we will discuss the latest advances, and what can be expected in the near future, in hardware, software and applications codes for advanced simulation of particle accelerators
Waste Evaporator Accident Simulation Using RELAP5 Computer Code
International Nuclear Information System (INIS)
POLIZZI, L.M.
2004-01-01
An evaporator is used on liquid waste from processing facilities to reduce the volume of the waste through heating the waste and allowing some of the water to be separated from the waste through boiling. This separation process allows for more efficient processing and storage of liquid waste. Commonly, the liquid waste consists of an aqueous solution of chemicals that over time could induce corrosion, and in turn weaken the tubes in the steam tube bundle of the waste evaporator that are used to heat the waste. This chemically induced corrosion could escalate into a possible tube leakage and/or the severance of a tube(s) in the tube bundle. In this paper, analyses of a waste evaporator system for the processing of liquid waste containing corrosive chemicals are presented to assess the system response to this accident scenario. This accident scenario is evaluated since its consequences can propagate to a release of hazardous material to the outside environment. It is therefore important to ensure that the evaporator system component structural integrity is not compromised, i.e. the design pressure and temperature of the system is not exceeded during the accident transient. The computer code used for the accident simulation is RELAP5-MOD31. The accident scenario analyzed includes a double-ended guillotine break of a tube in the tube bundle of the evaporator. A mitigated scenario is presented to evaluate the excursion of the peak pressure and temperature in the various components of the evaporator system to assess whether the protective actions and controls available are adequate to ensure that the structural integrity of the evaporator system is maintained and that no atmospheric release occurs
An introduction to LIME 1.0 and its use in coupling codes for multiphysics simulations.
Energy Technology Data Exchange (ETDEWEB)
Belcourt, Noel; Pawlowski, Roger Patrick; Schmidt, Rodney Cannon; Hooper, Russell Warren
2011-11-01
LIME is a small software package for creating multiphysics simulation codes. The name was formed as an acronym denoting 'Lightweight Integrating Multiphysics Environment for coupling codes.' LIME is intended to be especially useful when separate computer codes (which may be written in any standard computer language) already exist to solve different parts of a multiphysics problem. LIME provides the key high-level software (written in C++), a well defined approach (with example templates), and interface requirements to enable the assembly of multiple physics codes into a single coupled-multiphysics simulation code. In this report we introduce important software design characteristics of LIME, describe key components of a typical multiphysics application that might be created using LIME, and provide basic examples of its use - including the customized software that must be written by a user. We also describe the types of modifications that may be needed to individual physics codes in order for them to be incorporated into a LIME-based multiphysics application.
Software Development Processes Applied to Computational Icing Simulation
Levinson, Laurie H.; Potapezuk, Mark G.; Mellor, Pamela A.
1999-01-01
The development of computational icing simulation methods is making the transition form the research to common place use in design and certification efforts. As such, standards of code management, design validation, and documentation must be adjusted to accommodate the increased expectations of the user community with respect to accuracy, reliability, capability, and usability. This paper discusses these concepts with regard to current and future icing simulation code development efforts as implemented by the Icing Branch of the NASA Lewis Research Center in collaboration with the NASA Lewis Engineering Design and Analysis Division. With the application of the techniques outlined in this paper, the LEWICE ice accretion code has become a more stable and reliable software product.
Large-scale simulations of error-prone quantum computation devices
International Nuclear Information System (INIS)
Trieu, Doan Binh
2009-01-01
The theoretical concepts of quantum computation in the idealized and undisturbed case are well understood. However, in practice, all quantum computation devices do suffer from decoherence effects as well as from operational imprecisions. This work assesses the power of error-prone quantum computation devices using large-scale numerical simulations on parallel supercomputers. We present the Juelich Massively Parallel Ideal Quantum Computer Simulator (JUMPIQCS), that simulates a generic quantum computer on gate level. It comprises an error model for decoherence and operational errors. The robustness of various algorithms in the presence of noise has been analyzed. The simulation results show that for large system sizes and long computations it is imperative to actively correct errors by means of quantum error correction. We implemented the 5-, 7-, and 9-qubit quantum error correction codes. Our simulations confirm that using error-prone correction circuits with non-fault-tolerant quantum error correction will always fail, because more errors are introduced than being corrected. Fault-tolerant methods can overcome this problem, provided that the single qubit error rate is below a certain threshold. We incorporated fault-tolerant quantum error correction techniques into JUMPIQCS using Steane's 7-qubit code and determined this threshold numerically. Using the depolarizing channel as the source of decoherence, we find a threshold error rate of (5.2±0.2) x 10 -6 . For Gaussian distributed operational over-rotations the threshold lies at a standard deviation of 0.0431±0.0002. We can conclude that quantum error correction is especially well suited for the correction of operational imprecisions and systematic over-rotations. For realistic simulations of specific quantum computation devices we need to extend the generic model to dynamic simulations, i.e. time-dependent Hamiltonian simulations of realistic hardware models. We focus on today's most advanced technology, i
Shrekenhamer, Abraham; Gottesman, Stephen R.
2012-10-01
A novel and memory efficient method for computing diffraction patterns produced on large-scale focal planes by largescale Coded Apertures at wavelengths where diffraction effects are significant has been developed and tested. The scheme, readily implementable on portable computers, overcomes the memory limitations of present state-of-the-art simulation codes such as Zemax. The method consists of first calculating a set of reference complex field (amplitude and phase) patterns on the focal plane produced by a single (reference) central hole, extending to twice the focal plane array size, with one such pattern for each Line-of-Sight (LOS) direction and wavelength in the scene, and with the pattern amplitude corresponding to the square-root of the spectral irradiance from each such LOS direction in the scene at selected wavelengths. Next the set of reference patterns is transformed to generate pattern sets for other holes. The transformation consists of a translational pattern shift corresponding to each hole's position offset and an electrical phase shift corresponding to each hole's position offset and incoming radiance's direction and wavelength. The set of complex patterns for each direction and wavelength is then summed coherently and squared for each detector to yield a set of power patterns unique for each direction and wavelength. Finally the set of power patterns is summed to produce the full waveband diffraction pattern from the scene. With this tool researchers can now efficiently simulate diffraction patterns produced from scenes by large-scale Coded Apertures onto large-scale focal plane arrays to support the development and optimization of coded aperture masks and image reconstruction algorithms.
FLAME: A finite element computer code for contaminant transport n variably-saturated media
International Nuclear Information System (INIS)
Baca, R.G.; Magnuson, S.O.
1992-06-01
A numerical model was developed for use in performance assessment studies at the INEL. The numerical model referred to as the FLAME computer code, is designed to simulate subsurface contaminant transport in a variably-saturated media. The code can be applied to model two-dimensional contaminant transport in an and site vadose zone or in an unconfined aquifer. In addition, the code has the capability to describe transport processes in a porous media with discrete fractures. This report presents the following: description of the conceptual framework and mathematical theory, derivations of the finite element techniques and algorithms, computational examples that illustrate the capability of the code, and input instructions for the general use of the code. The development of the FLAME computer code is aimed at providing environmental scientists at the INEL with a predictive tool for the subsurface water pathway. This numerical model is expected to be widely used in performance assessments for: (1) the Remedial Investigation/Feasibility Study process and (2) compliance studies required by the US Department of energy Order 5820.2A
FLAME: A finite element computer code for contaminant transport n variably-saturated media
Energy Technology Data Exchange (ETDEWEB)
Baca, R.G.; Magnuson, S.O.
1992-06-01
A numerical model was developed for use in performance assessment studies at the INEL. The numerical model referred to as the FLAME computer code, is designed to simulate subsurface contaminant transport in a variably-saturated media. The code can be applied to model two-dimensional contaminant transport in an and site vadose zone or in an unconfined aquifer. In addition, the code has the capability to describe transport processes in a porous media with discrete fractures. This report presents the following: description of the conceptual framework and mathematical theory, derivations of the finite element techniques and algorithms, computational examples that illustrate the capability of the code, and input instructions for the general use of the code. The development of the FLAME computer code is aimed at providing environmental scientists at the INEL with a predictive tool for the subsurface water pathway. This numerical model is expected to be widely used in performance assessments for: (1) the Remedial Investigation/Feasibility Study process and (2) compliance studies required by the US Department of energy Order 5820.2A.
Simulations of Laboratory Astrophysics Experiments using the CRASH code
Trantham, Matthew; Kuranz, Carolyn; Fein, Jeff; Wan, Willow; Young, Rachel; Keiter, Paul; Drake, R. Paul
2015-11-01
Computer simulations can assist in the design and analysis of laboratory astrophysics experiments. The Center for Radiative Shock Hydrodynamics (CRASH) at the University of Michigan developed a code that has been used to design and analyze high-energy-density experiments on OMEGA, NIF, and other large laser facilities. This Eulerian code uses block-adaptive mesh refinement (AMR) with implicit multigroup radiation transport, electron heat conduction and laser ray tracing. This poster will demonstrate some of the experiments the CRASH code has helped design or analyze including: Kelvin-Helmholtz, Rayleigh-Taylor, magnetized flows, jets, and laser-produced plasmas. This work is funded by the following grants: DEFC52-08NA28616, DE-NA0001840, and DE-NA0002032.
Simulation and Noise Analysis of Multimedia Transmission in Optical CDMA Computer Networks
Directory of Open Access Journals (Sweden)
Nasaruddin Nasaruddin
2013-09-01
Full Text Available This paper simulates and analyzes noise of multimedia transmission in a flexible optical code division multiple access (OCDMA computer network with different quality of service (QoS requirements. To achieve multimedia transmission in OCDMA, we have proposed strict variable-weight optical orthogonal codes (VW-OOCs, which can guarantee the smallest correlation value of one by the optimal design. In developing multimedia transmission for computer network, a simulation tool is essential in analyzing the effectiveness of various transmissions of services. In this paper, implementation models are proposed to analyze the multimedia transmission in the representative of OCDMA computer networks by using MATLAB simulink tools. Simulation results of the models are discussed including spectrums outputs of transmitted signals, superimposed signals, received signals, and eye diagrams with and without noise. Using the proposed models, multimedia OCDMA computer network using the strict VW-OOC is practically evaluated. Furthermore, system performance is also evaluated by considering avalanche photodiode (APD noise and thermal noise. The results show that the system performance depends on code weight, received laser power, APD noise, and thermal noise which should be considered as important parameters to design and implement multimedia transmission in OCDMA computer networks.
Simulation and Noise Analysis of Multimedia Transmission in Optical CDMA Computer Networks
Directory of Open Access Journals (Sweden)
Nasaruddin
2009-11-01
Full Text Available This paper simulates and analyzes noise of multimedia transmission in a flexible optical code division multiple access (OCDMA computer network with different quality of service (QoS requirements. To achieve multimedia transmission in OCDMA, we have proposed strict variable-weight optical orthogonal codes (VW-OOCs, which can guarantee the smallest correlation value of one by the optimal design. In developing multimedia transmission for computer network, a simulation tool is essential in analyzing the effectiveness of various transmissions of services. In this paper, implementation models are proposed to analyze the multimedia transmission in the representative of OCDMA computer networks by using MATLAB simulink tools. Simulation results of the models are discussed including spectrums outputs of transmitted signals, superimposed signals, received signals, and eye diagrams with and without noise. Using the proposed models, multimedia OCDMA computer network using the strict VW-OOC is practically evaluated. Furthermore, system performance is also evaluated by considering avalanche photodiode (APD noise and thermal noise. The results show that the system performance depends on code weight, received laser power, APD noise, and thermal noise which should be considered as important parameters to design and implement multimedia transmission in OCDMA computer networks.
GeNN: a code generation framework for accelerated brain simulations
Yavuz, Esin; Turner, James; Nowotny, Thomas
2016-01-01
Large-scale numerical simulations of detailed brain circuit models are important for identifying hypotheses on brain functions and testing their consistency and plausibility. An ongoing challenge for simulating realistic models is, however, computational speed. In this paper, we present the GeNN (GPU-enhanced Neuronal Networks) framework, which aims to facilitate the use of graphics accelerators for computational models of large-scale neuronal networks to address this challenge. GeNN is an open source library that generates code to accelerate the execution of network simulations on NVIDIA GPUs, through a flexible and extensible interface, which does not require in-depth technical knowledge from the users. We present performance benchmarks showing that 200-fold speedup compared to a single core of a CPU can be achieved for a network of one million conductance based Hodgkin-Huxley neurons but that for other models the speedup can differ. GeNN is available for Linux, Mac OS X and Windows platforms. The source code, user manual, tutorials, Wiki, in-depth example projects and all other related information can be found on the project website http://genn-team.github.io/genn/.
Scientific codes developed and used at GRS. Nuclear simulation chain
Energy Technology Data Exchange (ETDEWEB)
Schaffrath, Andreas; Sonnenkalb, Martin; Sievers, Juergen; Luther, Wolfgang; Velkov, Kiril [Gesellschaft fuer Anlagen und Reaktorsicherheit (GRS) gGmbH, Garching/Muenchen (Germany). Forschungszentrum
2016-05-15
Over 60 technical experts of the reactor safety research division of the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH are developing and validating reliable methods and computer codes - summarized under the term nuclear simulation chain - for the safety-related assessment for all types of nuclear power plants (NPP) and other nuclear facilities considering the current state of science and technology. This nuclear simulation chain has to be able to simulate and assess all relevant physical processes and phenomena for all operating states and (severe) accidents. In the present contribution, the nuclear simulation chain developed and applied by GRS as well as selected examples of its application are presented. The latter demonstrate impressively the width of its scope and its performance. The GRS codes can be passed on request to other (national as well as international) organizations. This contributes to a worldwide increase of the nuclear safety standards. The code transfer is especially important for developing and emerging countries lacking the financial means and/or the necessary know-how for this purpose. At the end of this contribution, the respective course of action is described.
Computer codes in nuclear safety, radiation transport and dosimetry
International Nuclear Information System (INIS)
Bordy, J.M.; Kodeli, I.; Menard, St.; Bouchet, J.L.; Renard, F.; Martin, E.; Blazy, L.; Voros, S.; Bochud, F.; Laedermann, J.P.; Beaugelin, K.; Makovicka, L.; Quiot, A.; Vermeersch, F.; Roche, H.; Perrin, M.C.; Laye, F.; Bardies, M.; Struelens, L.; Vanhavere, F.; Gschwind, R.; Fernandez, F.; Quesne, B.; Fritsch, P.; Lamart, St.; Crovisier, Ph.; Leservot, A.; Antoni, R.; Huet, Ch.; Thiam, Ch.; Donadille, L.; Monfort, M.; Diop, Ch.; Ricard, M.
2006-01-01
The purpose of this conference was to describe the present state of computer codes dedicated to radiation transport or radiation source assessment or dosimetry. The presentations have been parted into 2 sessions: 1) methodology and 2) uses in industrial or medical or research domains. It appears that 2 different calculation strategies are prevailing, both are based on preliminary Monte-Carlo calculations with data storage. First, quick simulations made from a database of particle histories built though a previous Monte-Carlo simulation and secondly, a neuronal approach involving a learning platform generated through a previous Monte-Carlo simulation. This document gathers the slides of the presentations
STADIC: a computer code for combining probability distributions
International Nuclear Information System (INIS)
Cairns, J.J.; Fleming, K.N.
1977-03-01
The STADIC computer code uses a Monte Carlo simulation technique for combining probability distributions. The specific function for combination of the input distribution is defined by the user by introducing the appropriate FORTRAN statements to the appropriate subroutine. The code generates a Monte Carlo sampling from each of the input distributions and combines these according to the user-supplied function to provide, in essence, a random sampling of the combined distribution. When the desired number of samples is obtained, the output routine calculates the mean, standard deviation, and confidence limits for the resultant distribution. This method of combining probability distributions is particularly useful in cases where analytical approaches are either too difficult or undefined
Simulation of hydrogen deflagration experiment – Benchmark exercise with lumped-parameter codes
Energy Technology Data Exchange (ETDEWEB)
Kljenak, Ivo, E-mail: ivo.kljenak@ijs.si [Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Kuznetsov, Mikhail, E-mail: mike.kuznetsov@kit.edu [Karlsruhe Institute of Technology, Kaiserstraße 12, 76131 Karlsruhe (Germany); Kostka, Pal, E-mail: kostka@nubiki.hu [NUBIKI Nuclear Safety Research Institute, Konkoly-Thege Miklós út 29-33, 1121 Budapest (Hungary); Kubišova, Lubica, E-mail: lubica.kubisova@ujd.gov.sk [Nuclear Regulatory Authority of the Slovak Republic, Bajkalská 27, 82007 Bratislava (Slovakia); Maltsev, Mikhail, E-mail: maltsev_MB@aep.ru [JSC Atomenergoproekt, 1, st. Podolskykh Kursantov, Moscow (Russian Federation); Manzini, Giovanni, E-mail: giovanni.manzini@rse-web.it [Ricerca sul Sistema Energetico, Via Rubattino 54, 20134 Milano (Italy); Povilaitis, Mantas, E-mail: mantas.p@mail.lei.lt [Lithuania Energy Institute, Breslaujos g.3, 44403 Kaunas (Lithuania)
2015-03-15
Highlights: • Blind and open simulations of hydrogen combustion experiment in large-scale containment-like facility with different lumped-parameter codes. • Simulation of axial as well as radial flame propagation. • Confirmation of adequacy of lumped-parameter codes for safety analyses of actual nuclear power plants. - Abstract: An experiment on hydrogen deflagration (Upward Flame Propagation Experiment – UFPE) was proposed by the Jozef Stefan Institute (Slovenia) and performed in the HYKA A2 facility at the Karlsruhe Institute of Technology (Germany). The experimental results were used to organize a benchmark exercise for lumped-parameter codes. Six organizations (JSI, AEP, LEI, NUBIKI, RSE and UJD SR) participated in the benchmark exercise, using altogether four different computer codes: ANGAR, ASTEC, COCOSYS and ECART. Both blind and open simulations were performed. In general, all the codes provided satisfactory results of the pressure increase, whereas the results of the temperature show a wider dispersal. Concerning the flame axial and radial velocities, the results may be considered satisfactory, given the inherent simplification of the lumped-parameter description compared to the local instantaneous description.
Simulation of hydrogen deflagration experiment – Benchmark exercise with lumped-parameter codes
International Nuclear Information System (INIS)
Kljenak, Ivo; Kuznetsov, Mikhail; Kostka, Pal; Kubišova, Lubica; Maltsev, Mikhail; Manzini, Giovanni; Povilaitis, Mantas
2015-01-01
Highlights: • Blind and open simulations of hydrogen combustion experiment in large-scale containment-like facility with different lumped-parameter codes. • Simulation of axial as well as radial flame propagation. • Confirmation of adequacy of lumped-parameter codes for safety analyses of actual nuclear power plants. - Abstract: An experiment on hydrogen deflagration (Upward Flame Propagation Experiment – UFPE) was proposed by the Jozef Stefan Institute (Slovenia) and performed in the HYKA A2 facility at the Karlsruhe Institute of Technology (Germany). The experimental results were used to organize a benchmark exercise for lumped-parameter codes. Six organizations (JSI, AEP, LEI, NUBIKI, RSE and UJD SR) participated in the benchmark exercise, using altogether four different computer codes: ANGAR, ASTEC, COCOSYS and ECART. Both blind and open simulations were performed. In general, all the codes provided satisfactory results of the pressure increase, whereas the results of the temperature show a wider dispersal. Concerning the flame axial and radial velocities, the results may be considered satisfactory, given the inherent simplification of the lumped-parameter description compared to the local instantaneous description
Adaptation of HAMMER computer code to CYBER 170/750 computer
International Nuclear Information System (INIS)
Pinheiro, A.M.B.S.; Nair, R.P.K.
1982-01-01
The adaptation of HAMMER computer code to CYBER 170/750 computer is presented. The HAMMER code calculates cell parameters by multigroup transport theory and reactor parameters by few group diffusion theory. The auxiliary programs, the carried out modifications and the use of HAMMER system adapted to CYBER 170/750 computer are described. (M.C.K.) [pt
Generating performance portable geoscientific simulation code with Firedrake (Invited)
Ham, D. A.; Bercea, G.; Cotter, C. J.; Kelly, P. H.; Loriant, N.; Luporini, F.; McRae, A. T.; Mitchell, L.; Rathgeber, F.
2013-12-01
This presentation will demonstrate how a change in simulation programming paradigm can be exploited to deliver sophisticated simulation capability which is far easier to programme than are conventional models, is capable of exploiting different emerging parallel hardware, and is tailored to the specific needs of geoscientific simulation. Geoscientific simulation represents a grand challenge computational task: many of the largest computers in the world are tasked with this field, and the requirements of resolution and complexity of scientists in this field are far from being sated. However, single thread performance has stalled, even sometimes decreased, over the last decade, and has been replaced by ever more parallel systems: both as conventional multicore CPUs and in the emerging world of accelerators. At the same time, the needs of scientists to couple ever-more complex dynamics and parametrisations into their models makes the model development task vastly more complex. The conventional approach of writing code in low level languages such as Fortran or C/C++ and then hand-coding parallelism for different platforms by adding library calls and directives forces the intermingling of the numerical code with its implementation. This results in an almost impossible set of skill requirements for developers, who must simultaneously be domain science experts, numericists, software engineers and parallelisation specialists. Even more critically, it requires code to be essentially rewritten for each emerging hardware platform. Since new platforms are emerging constantly, and since code owners do not usually control the procurement of the supercomputers on which they must run, this represents an unsustainable development load. The Firedrake system, conversely, offers the developer the opportunity to write PDE discretisations in the high-level mathematical language UFL from the FEniCS project (http://fenicsproject.org). Non-PDE model components, such as parametrisations
SIMIFR: A code to simulate material movement in the Integral Fast Reactor
International Nuclear Information System (INIS)
White, A.M.; Orechwa, Yuri.
1991-01-01
The SIMIFR code has been written to simulate the movement of material through a process. This code can be used to investigate inventory differences in material balances, assist in process design, and to produce operational scheduling. The particular process that is of concern to the authors is that centered around Argonne National Laboratory's Integral Fast Reactor. This is a process which involves the irradiation of fissile material for power production, and the recycling of the irradiated reactor fuel pins into fresh fuel elements. To adequately simulate this process it is necessary to allow for locations which can contain either discrete items or homogeneous mixtures. It is also necessary to allow for a very flexible process control algorithm. Further, the code must have the capability of transmuting isotopic compositions and computing internally the fraction of material from a process ending up in a given location. The SIMIFR code has been developed to perform all of these tasks. In addition to simulating the process, the code is capable of generating random measurement values and sampling errors for all locations, and of producing a restart deck so that terminated problems may be continued. In this paper the authors first familiarize the reader with the IFR fuel cycle. The different capabilities of the SIMIFR code are described. Finally, the simulation of the IFR fuel cycle using the SIMIFR code is discussed. 4 figs
Computer code for thermal-hydraulic simulation of heat pressurizer tanks operation (Simterm-H)
International Nuclear Information System (INIS)
Sellos, R.F.
1987-01-01
It is presented the Simtherm-H computer code, developed for calculating the thermodynamic properties of the high pressure heating system and the feedwater tank in transient state for PWR nuclear power plants (1300 MWe). (E.G.) [pt
VAMPIR - A two-group two-dimensional diffusion computer code for burnup calculation
International Nuclear Information System (INIS)
Zmijarevic, I.; Petrovic, I.
1985-01-01
VAMPIR is a computer code which simulates the burnup within a reactor coe. It computes the neutron flux, power distribution and burnup taking into account spatial variations of temperature and xenon poisoning. Its overall reactor calculation uses diffusion theory with finite differences approximation in X-Y or R-Z geometry. Two-group macroscopic cross section data are prepared by the lattice cell code WIMS-D4 and stored in the library form of multi entry tabulation against the various parameters that significantly affect the physical conditions in the reactor core. herein, the main features of the program are presented. (author)
Fluid Dynamics Theory, Computation, and Numerical Simulation
Pozrikidis, Constantine
2009-01-01
Fluid Dynamics: Theory, Computation, and Numerical Simulation is the only available book that extends the classical field of fluid dynamics into the realm of scientific computing in a way that is both comprehensive and accessible to the beginner. The theory of fluid dynamics, and the implementation of solution procedures into numerical algorithms, are discussed hand-in-hand and with reference to computer programming. This book is an accessible introduction to theoretical and computational fluid dynamics (CFD), written from a modern perspective that unifies theory and numerical practice. There are several additions and subject expansions in the Second Edition of Fluid Dynamics, including new Matlab and FORTRAN codes. Two distinguishing features of the discourse are: solution procedures and algorithms are developed immediately after problem formulations are presented, and numerical methods are introduced on a need-to-know basis and in increasing order of difficulty. Matlab codes are presented and discussed for ...
User's manual for the G.T.M.-1 computer code
International Nuclear Information System (INIS)
Prado-Herrero, P.
1992-01-01
This document describes the GTM-1 ( Geosphere Transport Model, release-1) computer code and is intended to provide the reader with enough detailed information in order to use the code. GTM-1 was developed for the assessment of radionuclide migration by the ground water through geologic deposits whose properties can change along the pathway.GTM-1 solves the transport equation by the finite differences method ( Crank-Nicolson scheme ). It was developped for specific use within Probabilistic System Assessment (PSA) Monte Carlo Method codes; in this context the first application of GTM-1 was within the LISA (Long Term Isolation System Assessment) code. GTM-1 is also available as an independent model, which includes various submodels simulating a multi-barrier disposal system. The code has been tested with the PSACOIN ( Probabilistic System Assessment Codes intercomparison) benchmarks exercises from PSAC User Group (OECD/NEA). 10 refs., 6 Annex., 2 tabs
Methodology of modeling and measuring computer architectures for plasma simulations
Wang, L. P. T.
1977-01-01
A brief introduction to plasma simulation using computers and the difficulties on currently available computers is given. Through the use of an analyzing and measuring methodology - SARA, the control flow and data flow of a particle simulation model REM2-1/2D are exemplified. After recursive refinements the total execution time may be greatly shortened and a fully parallel data flow can be obtained. From this data flow, a matched computer architecture or organization could be configured to achieve the computation bound of an application problem. A sequential type simulation model, an array/pipeline type simulation model, and a fully parallel simulation model of a code REM2-1/2D are proposed and analyzed. This methodology can be applied to other application problems which have implicitly parallel nature.
International Nuclear Information System (INIS)
Soares, P.A.; Sirimarco, L.F.
1984-01-01
SACI-2 is a computer code created to study the dynamic behaviour of a PWR nuclear power plant. To evaluate the quality of its results, SACI-2 was used to recalculate commissioning tests done in BIBLIS-A nuclear power plant and to calculate postulated transients for Angra-2 reactor. The results of SACI-2 computer code from BIBLIS-A showed as much good agreement as those calculated with the KWU Loop 7 computer code for Angra-2. (E.G.) [pt
Cluster computing for lattice QCD simulations
International Nuclear Information System (INIS)
Coddington, P.D.; Williams, A.G.
2000-01-01
Full text: Simulations of lattice quantum chromodynamics (QCD) require enormous amounts of compute power. In the past, this has usually involved sharing time on large, expensive machines at supercomputing centres. Over the past few years, clusters of networked computers have become very popular as a low-cost alternative to traditional supercomputers. The dramatic improvements in performance (and more importantly, the ratio of price/performance) of commodity PCs, workstations, and networks have made clusters of off-the-shelf computers an attractive option for low-cost, high-performance computing. A major advantage of clusters is that since they can have any number of processors, they can be purchased using any sized budget, allowing research groups to install a cluster for their own dedicated use, and to scale up to more processors if additional funds become available. Clusters are now being built for high-energy physics simulations. Wuppertal has recently installed ALiCE, a cluster of 128 Alpha workstations running Linux, with a peak performance of 158 G flops. The Jefferson Laboratory in the US has a 16 node Alpha cluster and plans to upgrade to a 256 processor machine. In Australia, several large clusters have recently been installed. Swinburne University of Technology has a cluster of 64 Compaq Alpha workstations used for astrophysics simulations. Early this year our DHPC group constructed a cluster of 116 dual Pentium PCs (i.e. 232 processors) connected by a Fast Ethernet network, which is used by chemists at Adelaide University and Flinders University to run computational chemistry codes. The Australian National University has recently installed a similar PC cluster with 192 processors. The Centre for the Subatomic Structure of Matter (CSSM) undertakes large-scale high-energy physics calculations, mainly lattice QCD simulations. The choice of the computer and network hardware for a cluster depends on the particular applications to be run on the machine. Our
APC: A new code for Atmospheric Polarization Computations
International Nuclear Information System (INIS)
Korkin, Sergey V.; Lyapustin, Alexei I.; Rozanov, Vladimir V.
2013-01-01
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. -- Highlights: •A new code, APC, has been developed. •The code was validated against well-known codes. •The BPDF for an arbitrary Mueller matrix is computed
3-D electromagnetic plasma particle simulations on the Intel Delta parallel computer
International Nuclear Information System (INIS)
Wang, J.; Liewer, P.C.
1994-01-01
A three-dimensional electromagnetic PIC code has been developed on the 512 node Intel Touchstone Delta MIMD parallel computer. This code is based on the General Concurrent PIC algorithm which uses a domain decomposition to divide the computation among the processors. The 3D simulation domain can be partitioned into 1-, 2-, or 3-dimensional sub-domains. Particles must be exchanged between processors as they move among the subdomains. The Intel Delta allows one to use this code for very-large-scale simulations (i.e. over 10 8 particles and 10 6 grid cells). The parallel efficiency of this code is measured, and the overall code performance on the Delta is compared with that on Cray supercomputers. It is shown that their code runs with a high parallel efficiency of ≥ 95% for large size problems. The particle push time achieved is 115 nsecs/particle/time step for 162 million particles on 512 nodes. Comparing with the performance on a single processor Cray C90, this represents a factor of 58 speedup. The code uses a finite-difference leap frog method for field solve which is significantly more efficient than fast fourier transforms on parallel computers. The performance of this code on the 128 node Cray T3D will also be discussed
Reactor Simulations for Safeguards with the MCNP Utility for Reactor Evolution Code
International Nuclear Information System (INIS)
Shiba, T.; Fallot, M.
2015-01-01
To tackle nuclear material proliferation, we conducted several proliferation scenarios using the MURE (MCNP Utility for Reactor Evolution) code. The MURE code, developed by CNRS laboratories, is a precision, open-source code written in C++ that automates the preparation and computation of successive MCNP (Monte Carlo N-Particle) calculations and solves the Bateman equations in between, for burnup or thermal-hydraulics purposes. In addition, MURE has been completed recently with a module for the CHaracterization of Radioactive Sources, called CHARS, which computes the emitted gamma, beta and alpha rays associated to any fuel composition. Reactor simulations could allow knowing how plutonium or other material generation evolves inside reactors in terms of time and amount. The MURE code is appropriate for this purpose and can also provide knowledge on associated particle emissions. Using MURE, we have both developed a cell simulation of a typical CANDU reactor and a detailed model of light water PWR core, which could be used to analyze the composition of fuel assemblies as a function of time or burnup. MURE is also able to provide, thanks to its extension MURE-CHARTS, the emitted gamma rays from fuel assemblies unloaded from the core at any burnup. Diversion cases of Generation IV reactors have been also developed; a design of Very High Temperature Reactor (a Pebble Bed Reactor (PBR), loaded with UOx, PuOx and ThUOx fuels), and a Na-cooled Fast Breeder Reactor (FBR) (with depleted Uranium or Minor Actinides in the blanket). The loading of Protected Plutonium Production (P3) in the FBR was simulated. The simulations of various reactor designs taking into account reactor physics constraints may bring valuable information to inspectors. At this symposium, we propose to show the results of these reactor simulations as examples of the potentiality of reactor simulations for safeguards. (author)
Verification of the network flow and transport/distributed velocity (NWFT/DVM) computer code
International Nuclear Information System (INIS)
Duda, L.E.
1984-05-01
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
GEANT4 simulations for Proton computed tomography applications
International Nuclear Information System (INIS)
Yevseyeva, Olga; Assis, Joaquim T. de; Evseev, Ivan; Schelin, Hugo R.; Shtejer Diaz, Katherin; Lopes, Ricardo T.
2011-01-01
Proton radiation therapy is a highly precise form of cancer treatment. In existing proton treatment centers, dose calculations are performed based on X-ray computed tomography (CT). Alternatively, one could image the tumor directly with proton CT (pCT). Proton beams in medical applications deal with relatively thick targets like the human head or trunk. Thus, the fidelity of proton computed tomography (pCT) simulations as a tool for proton therapy planning depends in the general case on the accuracy of results obtained for the proton interaction with thick absorbers. GEANT4 simulations of proton energy spectra after passing thick absorbers do not agree well with existing experimental data, as showed previously. The spectra simulated for the Bethe-Bloch domain showed an unexpected sensitivity to the choice of low-energy electromagnetic models during the code execution. These observations were done with the GEANT4 version 8.2 during our simulations for pCT. This work describes in more details the simulations of the proton passage through gold absorbers with varied thickness. The simulations were done by modifying only the geometry in the Hadron therapy Example, and for all available choices of the Electromagnetic Physics Models. As the most probable reasons for these effects is some specific feature in the code or some specific implicit parameters in the GEANT4 manual, we continued our study with version 9.2 of the code. Some improvements in comparison with our previous results were obtained. The simulations were performed considering further applications for pCT development. The authors want to thank CNPq, CAPES and 'Fundacao Araucaria' for financial support of this work. (Author)
Availability of fusion plants employing a Monte Carlo simulation computer code
International Nuclear Information System (INIS)
Musicki, Z.
1984-01-01
The fusion facilities being built or designed will have availability problems due to their complexity and employment of not yet fully developed technologies. Low availability of test facilities will have an adverse impact on the learning time and will therefore push back the commercialization date of fusion. Low availability of commercial electric power plants will increase the cost of electricity and make fusion a less-attractive power source. Thus, the time to study the availability problems of fusion plants and suggest improvements is now, before costly mistakes are committed. This study is an initial effort in the area and is an attempt to develop methods for calculation of system's performance, specifically its availability, start collecting necessary data and identify the areas where data are lacking, as well as to point out the subsystems where resources need to be applied in order to bring about an acceptable system performance. The method used to study availability is a simulation computer code based on the Monte Carlo process and developed by the author. The fusion systems analyzed were TASKA (a tandem mirror test facility design) and MARS (a tandem mirror power plant design). The model and available data were employed to find that the most critical subsystems needing further work are the neutral beams, RF heating subsystems, direct convertor, and certain magnets
International Nuclear Information System (INIS)
Paulsen, M.P.; McFadden, J.H.; Peterson, C.E.; McClure, J.A.; Gose, G.C.; Jensen, P.J.
1991-01-01
The RETRAN-03 code development effort is designed to overcome the major theoretical and practical limitations associated with the RETRAN-02 computer code. The major objectives of the development program are to extend the range of analyses that can be performed with RETRAN, to make the code more dependable and faster running, and to have a more transportable code. The first two objectives are accomplished by developing new models and adding other models to the RETRAN-02 base code. The major model additions for RETRAN-03 are as follows: implicit solution methods for the steady-state and transient forms of the field equations; additional options for the velocity difference equation; a new steady-state initialization option for computer low-power steam generator initial conditions; models for nonequilibrium thermodynamic conditions; and several special-purpose models. The source code and the environmental library for RETRAN-03 are written in standard FORTRAN 77, which allows the last objective to be fulfilled. Some models in RETRAN-02 have been deleted in RETRAN-03. In this paper the changes between RETRAN-02 and RETRAN-03 are reviewed
Energy Technology Data Exchange (ETDEWEB)
White, M.J.; Iskander, M.F. [Univ. of Utah, Salt Lake City, UT (United States). Electrical Engineering Dept.; Kimrey, H.D. [Oak Ridge National Lab., TN (United States)
1996-12-31
The Finite-Difference Time-Domain (FDTD) code available at the University of Utah has been used to simulate sintering of ceramics in single and multimode cavities, and many useful results have been reported in literature. More detailed and accurate results, specifically around and including the ceramic sample, are often desired to help evaluate the adequacy of the heating procedure. In electrically large multimode cavities, however, computer memory requirements limit the number of the mathematical cells, and the desired resolution is impractical to achieve due to limited computer resources. Therefore, an FDTD algorithm which incorporates multiple-grid regions with variable-grid sizes is required to adequately perform the desired simulations. In this paper the authors describe the development of a three-dimensional multi-grid FDTD code to help focus a large number of cells around the desired region. Test geometries were solved using a uniform-grid and the developed multi-grid code to help validate the results from the developed code. Results from these comparisons, as well as the results of comparisons between the developed FDTD code and other available variable-grid codes are presented. In addition, results from the simulation of realistic microwave sintering experiments showed improved resolution in critical sites inside the three-dimensional sintering cavity. With the validation of the FDTD code, simulations were performed for electrically large, multimode, microwave sintering cavities to fully demonstrate the advantages of the developed multi-grid FDTD code.
Probabilistic evaluations for CANTUP computer code analysis improvement
International Nuclear Information System (INIS)
Florea, S.; Pavelescu, M.
2004-01-01
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
Implatation of MC2 computer code
International Nuclear Information System (INIS)
Seehusen, J.; Nair, R.P.K.; Becceneri, J.C.
1981-01-01
The implantation of MC2 computer code in the CDC system is presented. The MC2 computer code calculates multigroup cross sections for tipical compositions of fast reactors. The multigroup constants are calculated using solutions of PI or BI approximations for determined buckling value as weighting function. (M.C.K.) [pt
Cloud Computing for Complex Performance Codes.
Energy Technology Data Exchange (ETDEWEB)
Appel, Gordon John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Klein, Brandon Thorin [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Miner, John Gifford [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
2017-02-01
This report describes the use of cloud computing services for running complex public domain performance assessment problems. The work consisted of two phases: Phase 1 was to demonstrate complex codes, on several differently configured servers, could run and compute trivial small scale problems in a commercial cloud infrastructure. Phase 2 focused on proving non-trivial large scale problems could be computed in the commercial cloud environment. The cloud computing effort was successfully applied using codes of interest to the geohydrology and nuclear waste disposal modeling community.
Computer codes for safety analysis
International Nuclear Information System (INIS)
Holland, D.F.
1986-11-01
Computer codes for fusion safety analysis have been under development in the United States for about a decade. This paper will discuss five codes that are currently under development by the Fusion Safety Program. The purpose and capability of each code will be presented, a sample given, followed by a discussion of the present status and future development plans
Steam generator transient studies using a simplified two-fluid computer code
International Nuclear Information System (INIS)
Munshi, P.; Bhatnagar, R.; Ram, K.S.
1985-01-01
A simplified two-fluid computer code has been used to simulate reactor-side (or primary-side) transients in a PWR steam generator. The disturbances are modelled as ramp inputs for pressure, internal energy and mass flow-rate for the primary fluid. The CPU time for a transient duration of 4 s is approx. 10 min on a DEC-1090 computer system. The results are thermodynamically consistent and encouraging for further studies. (author)
Use of computer codes for system reliability analysis
International Nuclear Information System (INIS)
Sabek, M.; Gaafar, M.; Poucet, A.
1989-01-01
This paper gives a summary of studies performed at the JRC, ISPRA on the use of computer codes for complex systems analysis. The computer codes dealt with are: CAFTS-SALP software package, FRACTIC, FTAP, computer code package RALLY, and BOUNDS. Two reference case studies were executed by each code. The probabilistic results obtained, as well as the computation times are compared. The two cases studied are the auxiliary feedwater system of a 1300 MW PWR reactor and the emergency electrical power supply system. (author)
Use of computer codes for system reliability analysis
Energy Technology Data Exchange (ETDEWEB)
Sabek, M.; Gaafar, M. (Nuclear Regulatory and Safety Centre, Atomic Energy Authority, Cairo (Egypt)); Poucet, A. (Commission of the European Communities, Ispra (Italy). Joint Research Centre)
1989-01-01
This paper gives a summary of studies performed at the JRC, ISPRA on the use of computer codes for complex systems analysis. The computer codes dealt with are: CAFTS-SALP software package, FRACTIC, FTAP, computer code package RALLY, and BOUNDS. Two reference case studies were executed by each code. The probabilistic results obtained, as well as the computation times are compared. The two cases studied are the auxiliary feedwater system of a 1300 MW PWR reactor and the emergency electrical power supply system. (author).
Large-scale simulations of error-prone quantum computation devices
Energy Technology Data Exchange (ETDEWEB)
Trieu, Doan Binh
2009-07-01
The theoretical concepts of quantum computation in the idealized and undisturbed case are well understood. However, in practice, all quantum computation devices do suffer from decoherence effects as well as from operational imprecisions. This work assesses the power of error-prone quantum computation devices using large-scale numerical simulations on parallel supercomputers. We present the Juelich Massively Parallel Ideal Quantum Computer Simulator (JUMPIQCS), that simulates a generic quantum computer on gate level. It comprises an error model for decoherence and operational errors. The robustness of various algorithms in the presence of noise has been analyzed. The simulation results show that for large system sizes and long computations it is imperative to actively correct errors by means of quantum error correction. We implemented the 5-, 7-, and 9-qubit quantum error correction codes. Our simulations confirm that using error-prone correction circuits with non-fault-tolerant quantum error correction will always fail, because more errors are introduced than being corrected. Fault-tolerant methods can overcome this problem, provided that the single qubit error rate is below a certain threshold. We incorporated fault-tolerant quantum error correction techniques into JUMPIQCS using Steane's 7-qubit code and determined this threshold numerically. Using the depolarizing channel as the source of decoherence, we find a threshold error rate of (5.2{+-}0.2) x 10{sup -6}. For Gaussian distributed operational over-rotations the threshold lies at a standard deviation of 0.0431{+-}0.0002. We can conclude that quantum error correction is especially well suited for the correction of operational imprecisions and systematic over-rotations. For realistic simulations of specific quantum computation devices we need to extend the generic model to dynamic simulations, i.e. time-dependent Hamiltonian simulations of realistic hardware models. We focus on today's most advanced
Interactive computer codes for education and training on nuclear safety and radioprotection
International Nuclear Information System (INIS)
Leszczynski, Francisco
2008-01-01
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)
ASAS: Computational code for Analysis and Simulation of Atomic Spectra
Directory of Open Access Journals (Sweden)
Jhonatha R. dos Santos
2017-01-01
Full Text Available The laser isotopic separation process is based on the selective photoionization principle and, because of this, it is necessary to know the absorption spectrum of the desired atom. Computational resource has become indispensable for the planning of experiments and analysis of the acquired data. The ASAS (Analysis and Simulation of Atomic Spectra software presented here is a helpful tool to be used in studies involving atomic spectroscopy. The input for the simulations is friendly and essentially needs a database containing the energy levels and spectral lines of the atoms subjected to be studied.
PHoToNs–A parallel heterogeneous and threads oriented code for cosmological N-body simulation
Wang, Qiao; Cao, Zong-Yan; Gao, Liang; Chi, Xue-Bin; Meng, Chen; Wang, Jie; Wang, Long
2018-06-01
We introduce a new code for cosmological simulations, PHoToNs, which incorporates features for performing massive cosmological simulations on heterogeneous high performance computer (HPC) systems and threads oriented programming. PHoToNs adopts a hybrid scheme to compute gravitational force, with the conventional Particle-Mesh (PM) algorithm to compute the long-range force, the Tree algorithm to compute the short range force and the direct summation Particle-Particle (PP) algorithm to compute gravity from very close particles. A self-similar space filling a Peano-Hilbert curve is used to decompose the computing domain. Threads programming is advantageously used to more flexibly manage the domain communication, PM calculation and synchronization, as well as Dual Tree Traversal on the CPU+MIC platform. PHoToNs scales well and efficiency of the PP kernel achieves 68.6% of peak performance on MIC and 74.4% on CPU platforms. We also test the accuracy of the code against the much used Gadget-2 in the community and found excellent agreement.
Application of the TEMPEST computer code to canister-filling heat transfer problems
International Nuclear Information System (INIS)
Farnsworth, R.K.; Faletti, D.W.; Budden, M.J.
1988-03-01
Pacific Northwest Laboratory (PNL) researchers used the TEMPEST computer code to simulate thermal cooldown behavior of nuclear waste glass after it was poured into steel canisters for long-term storage. The objective of this work was to determine the accuracy and applicability of the TEMPEST code when used to compute canister thermal histories. First, experimental data were obtained to provide the basis for comparing TEMPEST-generated predictions. Five canisters were instrumented with appropriately located radial and axial thermocouples. The canister were filled using the pilot-scale ceramic melter (PSCM) at PNL. Each canister was filled in either a continous or a batch filling mode. One of the canisters was also filled within a turntable simulant (a group of cylindrical shells with heat transfer resistances similar to those in an actual melter turntable). This was necessary to provide a basis for assessing the ability of the TEMPEST code to also model the transient cooling of canisters in a melter turntable. The continous-fill model, Version M, was found to predict temperatures with more accuracy. The turntable simulant experiment demonstrated that TEMPEST can adequately model the asymmetric temperature field caused by the turntable geometry. Further, TEMPEST can acceptably predict the canister cooling history within a turntable, despite code limitations in computing simultaneous radiation and convection heat transfer between shells, along with uncertainty in stainless-steel surface emissivities. Based on the successful performance of TEMPEST Version M, development was initiated to incorporate 1) full viscous glass convection, 2) a dynamically adaptive grid that automatically follows the glass/air interface throughout the transient, and 3) a full enclosure radiation model to allow radiation heat transfer to non-nearest neighbor cells. 5 refs., 47 figs., 17 tabs
Fluid dynamics theory, computation, and numerical simulation
Pozrikidis, C
2001-01-01
Fluid Dynamics Theory, Computation, and Numerical Simulation is the only available book that extends the classical field of fluid dynamics into the realm of scientific computing in a way that is both comprehensive and accessible to the beginner The theory of fluid dynamics, and the implementation of solution procedures into numerical algorithms, are discussed hand-in-hand and with reference to computer programming This book is an accessible introduction to theoretical and computational fluid dynamics (CFD), written from a modern perspective that unifies theory and numerical practice There are several additions and subject expansions in the Second Edition of Fluid Dynamics, including new Matlab and FORTRAN codes Two distinguishing features of the discourse are solution procedures and algorithms are developed immediately after problem formulations are presented, and numerical methods are introduced on a need-to-know basis and in increasing order of difficulty Matlab codes are presented and discussed for a broad...
International Nuclear Information System (INIS)
Yamada, Tomonori
2010-01-01
The safety requirement of nuclear power plant attracts much attention nowadays. With the growing computing power, numerical simulation is one of key technologies to meet this safety requirement. Center for Computational Science and e-Systems of Japan Atomic Energy Agency has been developing a finite element analysis code for assembled structure to accurately evaluate the structural integrity of nuclear power plant in its entirety under seismic events. Because nuclear power plant is very huge assembled structure with tens of millions of mechanical components, the finite element model of each component is assembled into one structure and non-conforming meshes of mechanical components are bonded together inside the code. The main technique to bond these mechanical components is triple sparse matrix multiplication with multiple point constrains and global stiffness matrix. In our code, this procedure is conducted in a component by component manner, so that the working memory size and computing time for this multiplication are available on the current computing environment. As an illustrative example, seismic simulation of a real nuclear reactor of High Temperature engineering Test Reactor, which is located at the O-arai research and development center of JAEA, with 80 major mechanical components was conducted. Consequently, our code successfully simulated detailed elasto-plastic deformation of nuclear reactor and its computational performance was investigated. (author)
Computer simulation of ultrasonic waves in solids
International Nuclear Information System (INIS)
Thibault, G.A.; Chaplin, K.
1992-01-01
A computer model that simulates the propagation of ultrasonic waves has been developed at AECL Research, Chalk River Laboratories. This program is called EWE, short for Elastic Wave Equations, the mathematics governing the propagation of ultrasonic waves. This report contains a brief summary of the use of ultrasonic waves in non-destructive testing techniques, a discussion of the EWE simulation code explaining the implementation of the equations and the types of output received from the model, and an example simulation showing the abilities of the model. (author). 2 refs., 2 figs
Electromagnetic simulations of the ASDEX Upgrade ICRF Antenna with the TOPICA code
International Nuclear Information System (INIS)
Krivska, A.; Milanesio, D.; Bobkov, V.; Braun, F.; Noterdaeme, J.-M.
2009-01-01
Accurate and efficient simulation tools are necessary to optimize the ICRF antenna design for a set of operational conditions. The TOPICA code was developed for performance prediction and for the analysis of ICRF antenna systems in the presence of plasma, given realistic antenna geometries. Fully 3D antenna geometries can be adopted in TOPICA, just as in available commercial codes. But while those commercial codes cannot operate with a plasma loading, the TOPICA code correctly accounts for realistic plasma loading conditions, by means of the coupling with 1D FELICE code. This paper presents the evaluation of the electric current distribution on the structure, of the parallel electric field in the region between the straps and the plasma and the computation of sheaths driving RF potentials. Results of TOPICA simulations will help to optimize and re-design the ICRF ASDEX Upgrade antenna in order to reduce tungsten (W) sputtering attributed to the rectified sheath effect during ICRF operation.
International Nuclear Information System (INIS)
Courageot, Estelle
2010-01-01
After a description of the context of radiological accidents (definition, history, context, exposure types, associated clinic symptoms of irradiation and contamination, medical treatment, return on experience) and a presentation of dose assessment in the case of external exposure (clinic, biological and physical dosimetry), this research thesis describes the principles of numerical reconstruction of a radiological accident, presents some computation codes (Monte Carlo code, MCNPX code) and the SESAME tool, and reports an application to an actual case (an accident which occurred in Equator in April 2009). The next part reports the developments performed to modify the posture of voxelized phantoms and the experimental and numerical validations. The last part reports a feasibility study for the reconstruction of radiological accidents occurring in external radiotherapy. This work is based on a Monte Carlo simulation of a linear accelerator, with the aim of identifying the most relevant parameters to be implemented in SESAME in the case of external radiotherapy
Srna - Monte Carlo codes for proton transport simulation in combined and voxelized geometries
Directory of Open Access Journals (Sweden)
Ilić Radovan D.
2002-01-01
Full Text Available This paper describes new Monte Carlo codes for proton transport simulations in complex geometrical forms and in materials of different composition. The SRNA codes were developed for three dimensional (3D dose distribution calculation in proton therapy and dosimetry. The model of these codes is based on the theory of proton multiple scattering and a simple model of compound nucleus decay. The developed package consists of two codes: SRNA-2KG and SRNA-VOX. The first code simulates proton transport in combined geometry that can be described by planes and second order surfaces. The second one uses the voxelized geometry of material zones and is specifically adopted for the application of patient computer tomography data. Transition probabilities for both codes are given by the SRNADAT program. In this paper, we will present the models and algorithms of our programs, as well as the results of the numerical experiments we have carried out applying them, along with the results of proton transport simulation obtained through the PETRA and GEANT programs. The simulation of the proton beam characterization by means of the Multi-Layer Faraday Cup and spatial distribution of positron emitters obtained by our program indicate the imminent application of Monte Carlo techniques in clinical practice.
Application of parallel computing to seismic damage process simulation of an arch dam
International Nuclear Information System (INIS)
Zhong Hong; Lin Gao; Li Jianbo
2010-01-01
The simulation of damage process of high arch dam subjected to strong earthquake shocks is significant to the evaluation of its performance and seismic safety, considering the catastrophic effect of dam failure. However, such numerical simulation requires rigorous computational capacity. Conventional serial computing falls short of that and parallel computing is a fairly promising solution to this problem. The parallel finite element code PDPAD was developed for the damage prediction of arch dams utilizing the damage model with inheterogeneity of concrete considered. Developed with programming language Fortran, the code uses a master/slave mode for programming, domain decomposition method for allocation of tasks, MPI (Message Passing Interface) for communication and solvers from AZTEC library for solution of large-scale equations. Speedup test showed that the performance of PDPAD was quite satisfactory. The code was employed to study the damage process of a being-built arch dam on a 4-node PC Cluster, with more than one million degrees of freedom considered. The obtained damage mode was quite similar to that of shaking table test, indicating that the proposed procedure and parallel code PDPAD has a good potential in simulating seismic damage mode of arch dams. With the rapidly growing need for massive computation emerged from engineering problems, parallel computing will find more and more applications in pertinent areas.
International Nuclear Information System (INIS)
Nanty, Simon
2015-01-01
This work relates to the framework of uncertainty quantification for numerical simulators, and more precisely studies two industrial applications linked to the safety studies of nuclear plants. These two applications have several common features. The first one is that the computer code inputs are functional and scalar variables, functional ones being dependent. The second feature is that the probability distribution of functional variables is known only through a sample of their realizations. The third feature, relative to only one of the two applications, is the high computational cost of the code, which limits the number of possible simulations. The main objective of this work was to propose a complete methodology for the uncertainty analysis of numerical simulators for the two considered cases. First, we have proposed a methodology to quantify the uncertainties of dependent functional random variables from a sample of their realizations. This methodology enables to both model the dependency between variables and their link to another variable, called co-variate, which could be, for instance, the output of the considered code. Then, we have developed an adaptation of a visualization tool for functional data, which enables to simultaneously visualize the uncertainties and features of dependent functional variables. Second, a method to perform the global sensitivity analysis of the codes used in the two studied cases has been proposed. In the case of a computationally demanding code, the direct use of quantitative global sensitivity analysis methods is intractable. To overcome this issue, the retained solution consists in building a surrogate model or meta model, a fast-running model approximating the computationally expensive code. An optimized uniform sampling strategy for scalar and functional variables has been developed to build a learning basis for the meta model. Finally, a new approximation approach for expensive codes with functional outputs has been
Computer codes for RF cavity design
International Nuclear Information System (INIS)
Ko, K.
1992-08-01
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
International Nuclear Information System (INIS)
BEEBE - WANG, J.; LUCCIO, A.U.; D IMPERIO, N.; MACHIDA, S.
2002-01-01
Space charge in high intensity beams is an important issue in accelerator physics. Due to the complicity of the problems, the most effective way of investigating its effect is by computer simulations. In the resent years, many space charge simulation methods have been developed and incorporated in various 2D or 3D multi-particle-tracking codes. It has becoming necessary to benchmark these methods against each other, and against experimental results. As a part of global effort, we present our initial comparison of the space charge methods incorporated in simulation codes ORBIT++, ORBIT and SIMPSONS. In this paper, the methods included in these codes are overviewed. The simulation results are presented and compared. Finally, from this study, the advantages and disadvantages of each method are discussed
Energy Technology Data Exchange (ETDEWEB)
BEEBE - WANG,J.; LUCCIO,A.U.; D IMPERIO,N.; MACHIDA,S.
2002-06-03
Space charge in high intensity beams is an important issue in accelerator physics. Due to the complicity of the problems, the most effective way of investigating its effect is by computer simulations. In the resent years, many space charge simulation methods have been developed and incorporated in various 2D or 3D multi-particle-tracking codes. It has becoming necessary to benchmark these methods against each other, and against experimental results. As a part of global effort, we present our initial comparison of the space charge methods incorporated in simulation codes ORBIT++, ORBIT and SIMPSONS. In this paper, the methods included in these codes are overviewed. The simulation results are presented and compared. Finally, from this study, the advantages and disadvantages of each method are discussed.
Computational Simulation of a Water-Cooled Heat Pump
Bozarth, Duane
2008-01-01
A Fortran-language computer program for simulating the operation of a water-cooled vapor-compression heat pump in any orientation with respect to gravity has been developed by modifying a prior general-purpose heat-pump design code used at Oak Ridge National Laboratory (ORNL).
Computer codes for the analysis of flask impact problems
International Nuclear Information System (INIS)
Neilson, A.J.
1984-09-01
This review identifies typical features of the design of transportation flasks and considers some of the analytical tools required for the analysis of impact events. Because of the complexity of the physical problem, it is unlikely that a single code will adequately deal with all the aspects of the impact incident. Candidate codes are identified on the basis of current understanding of their strengths and limitations. It is concluded that the HONDO-II, DYNA3D AND ABAQUS codes which ar already mounted on UKAEA computers will be suitable tools for use in the analysis of experiments conducted in the proposed AEEW programme and of general flask impact problems. Initial attention should be directed at the DYNA3D and ABAQUS codes with HONDO-II being reserved for situations where the three-dimensional elements of DYNA3D may provide uneconomic simulations in planar or axisymmetric geometries. Attention is drawn to the importance of access to suitable mesh generators to create the nodal coordinate and element topology data required by these structural analysis codes. (author)
ELEGANT: A flexible SDDS-compliant code for accelerator simulation
International Nuclear Information System (INIS)
Borland, M.
2000-01-01
ELEGANT (ELEctron Generation ANd Tracking) is the principle accelerator simulation code used at the Advanced Photon Source (APS) for circular and one-pass machines. Capabilities include 6-D tracking using matrices up to third order, canonical integration, and numerical integration. Standard beamline elements are supported, as well as coherent synchrotron radiation, wakefields, rf elements, kickers, apertures, scattering, and more. In addition to tracking with and without errors, ELEGANT performs optimization of tracked properties, as well as computation and optimization of Twiss parameters, radiation integrals, matrices, and floor coordinates. Orbit/trajectory, tune, and chromaticity correction are supported. ELEGANT is fully compliant with the Self Describing Data Sets (SDDS) file protocol, and hence uses the SDDS Toolkit for pre- and post-processing. This permits users to prepare scripts to run the code in a flexible and automated fashion. It is particularly well suited to multistage simulation and concurrent simulation on many workstations. Several examples of complex projects performed with ELEGANT are given, including top-up safety analysis of the APS and design of the APS bunch compressor
HYDRA-II: A hydrothermal analysis computer code: Volume 1, Equations and numerics
International Nuclear Information System (INIS)
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
Computer codes for RF cavity design
International Nuclear Information System (INIS)
Ko, K.
1992-01-01
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 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.
Microgravity computing codes. User's guide
1982-01-01
Codes used in microgravity experiments to compute fluid parameters and to obtain data graphically are introduced. The computer programs are stored on two diskettes, compatible with the floppy disk drives of the Apple 2. Two versions of both disks are available (DOS-2 and DOS-3). The codes are written in BASIC and are structured as interactive programs. Interaction takes place through the keyboard of any Apple 2-48K standard system with single floppy disk drive. The programs are protected against wrong commands given by the operator. The programs are described step by step in the same order as the instructions displayed on the monitor. Most of these instructions are shown, with samples of computation and of graphics.
A conceptual design of multidisciplinary-integrated C.F.D. simulation on parallel computers
International Nuclear Information System (INIS)
Onishi, Ryoichi; Ohta, Takashi; Kimura, Toshiya.
1996-11-01
A design of a parallel aeroelastic code for aircraft integrated simulations is conducted. The method for integrating aerodynamics and structural dynamics software on parallel computers is devised by using the Euler/Navier-Stokes equations coupled with wing-box finite element structures. A synthesis of modern aircraft requires the optimizations of aerodynamics, structures, controls, operabilities, or other design disciplines, and the R and D efforts to implement Multidisciplinary Design Optimization environments using high performance computers are made especially among the U.S. aerospace industries. This report describes a Multiple Program Multiple Data (MPMD) parallelization of aerodynamics and structural dynamics codes with a dynamic deformation grid. A three-dimensional computation of a flowfield with dynamic deformation caused by a structural deformation is performed, and a pressure data calculated is used for a computation of the structural deformation which is input again to a fluid dynamics code. This process is repeated exchanging the computed data of pressures and deformations between flowfield grids and structural elements. It enables to simulate the structure movements which take into account of the interaction of fluid and structure. The conceptual design for achieving the aforementioned various functions is reported. Also the future extensions to incorporate control systems, which enable to simulate a realistic aircraft configuration to be a major tool for Aircraft Integrated Simulation, are investigated. (author)
International Nuclear Information System (INIS)
Delene, J.
1984-01-01
CONCEPT is a computer code that will provide conceptual capital investment cost estimates for nuclear and coal-fired power plants. The code can develop an estimate for construction at any point in time. Any unit size within the range of about 400 to 1300 MW electric may be selected. Any of 23 reference site locations across the United States and Canada may be selected. PWR, BWR, and coal-fired plants burning high-sulfur and low-sulfur coal can be estimated. Multiple-unit plants can be estimated. Costs due to escalation/inflation and interest during construction are calculated
Simulation of biological ion channels with technology computer-aided design.
Pandey, Santosh; Bortei-Doku, Akwete; White, Marvin H
2007-01-01
Computer simulations of realistic ion channel structures have always been challenging and a subject of rigorous study. Simulations based on continuum electrostatics have proven to be computationally cheap and reasonably accurate in predicting a channel's behavior. In this paper we discuss the use of a device simulator, SILVACO, to build a solid-state model for KcsA channel and study its steady-state response. SILVACO is a well-established program, typically used by electrical engineers to simulate the process flow and electrical characteristics of solid-state devices. By employing this simulation program, we have presented an alternative computing platform for performing ion channel simulations, besides the known methods of writing codes in programming languages. With the ease of varying the different parameters in the channel's vestibule and the ability of incorporating surface charges, we have shown the wide-ranging possibilities of using a device simulator for ion channel simulations. Our simulated results closely agree with the experimental data, validating our model.
Directory of Open Access Journals (Sweden)
Leonardo da Silva Boia
2014-03-01
Full Text Available Purpose: A computational system was developed for this paper in the C++ programming language, to create a 125I radioactive seed entry file, based on the positioning of a virtual grid (template in voxel geometries, with the purpose of performing prostate cancer treatment simulations using the MCNPX code.Methods: The system is fed with information from the planning system with regard to each seed’s location and its depth, and an entry file is automatically created with all the cards (instructions for each seed regarding their cell blocks and surfaces spread out spatially in the 3D environment. The system provides with precision a reproduction of the clinical scenario for the MCNPX code’s simulation environment, thereby allowing the technique’s in-depth study.Results and Conclusion: The preliminary results from this study showed that the lateral penumbra of uniform scanning proton beams was less sensitive In order to validate the computational system, an entry file was created with 88 125I seeds that were inserted in the phantom’s MAX06 prostate region with initial activity determined for the seeds at the 0.27 mCi value. Isodose curves were obtained in all the prostate slices in 5 mm steps in the 7 to 10 cm interval, totaling 7 slices. Variance reduction techniques were applied in order to optimize computational time and the reduction of uncertainties such as photon and electron energy interruptions in 4 keV and forced collisions regarding cells of interest. Through the acquisition of isodose curves, the results obtained show that hot spots have values above 300 Gy, as anticipated in literature, stressing the importance of the sources’ correct positioning, in which the computational system developed provides, in order not to release excessive doses in adjacent risk organs. The 144 Gy prescription curve showed in the validation process that it covers perfectly a large percentage of the volume, at the same time that it demonstrates a large
40 CFR 194.23 - Models and computer codes.
2010-07-01
... 40 Protection of Environment 24 2010-07-01 2010-07-01 false Models and computer codes. 194.23... General Requirements § 194.23 Models and computer codes. (a) Any compliance application shall include: (1... obtain stable solutions; (iv) Computer models accurately implement the numerical models; i.e., computer...
Computational strategies for three-dimensional flow simulations on distributed computer systems
Sankar, Lakshmi N.; Weed, Richard A.
1995-08-01
This research effort is directed towards an examination of issues involved in porting large computational fluid dynamics codes in use within the industry to a distributed computing environment. This effort addresses strategies for implementing the distributed computing in a device independent fashion and load balancing. A flow solver called TEAM presently in use at Lockheed Aeronautical Systems Company was acquired to start this effort. The following tasks were completed: (1) The TEAM code was ported to a number of distributed computing platforms including a cluster of HP workstations located in the School of Aerospace Engineering at Georgia Tech; a cluster of DEC Alpha Workstations in the Graphics visualization lab located at Georgia Tech; a cluster of SGI workstations located at NASA Ames Research Center; and an IBM SP-2 system located at NASA ARC. (2) A number of communication strategies were implemented. Specifically, the manager-worker strategy and the worker-worker strategy were tested. (3) A variety of load balancing strategies were investigated. Specifically, the static load balancing, task queue balancing and the Crutchfield algorithm were coded and evaluated. (4) The classical explicit Runge-Kutta scheme in the TEAM solver was replaced with an LU implicit scheme. And (5) the implicit TEAM-PVM solver was extensively validated through studies of unsteady transonic flow over an F-5 wing, undergoing combined bending and torsional motion. These investigations are documented in extensive detail in the dissertation, 'Computational Strategies for Three-Dimensional Flow Simulations on Distributed Computing Systems', enclosed as an appendix.
Computational strategies for three-dimensional flow simulations on distributed computer systems
Sankar, Lakshmi N.; Weed, Richard A.
1995-01-01
This research effort is directed towards an examination of issues involved in porting large computational fluid dynamics codes in use within the industry to a distributed computing environment. This effort addresses strategies for implementing the distributed computing in a device independent fashion and load balancing. A flow solver called TEAM presently in use at Lockheed Aeronautical Systems Company was acquired to start this effort. The following tasks were completed: (1) The TEAM code was ported to a number of distributed computing platforms including a cluster of HP workstations located in the School of Aerospace Engineering at Georgia Tech; a cluster of DEC Alpha Workstations in the Graphics visualization lab located at Georgia Tech; a cluster of SGI workstations located at NASA Ames Research Center; and an IBM SP-2 system located at NASA ARC. (2) A number of communication strategies were implemented. Specifically, the manager-worker strategy and the worker-worker strategy were tested. (3) A variety of load balancing strategies were investigated. Specifically, the static load balancing, task queue balancing and the Crutchfield algorithm were coded and evaluated. (4) The classical explicit Runge-Kutta scheme in the TEAM solver was replaced with an LU implicit scheme. And (5) the implicit TEAM-PVM solver was extensively validated through studies of unsteady transonic flow over an F-5 wing, undergoing combined bending and torsional motion. These investigations are documented in extensive detail in the dissertation, 'Computational Strategies for Three-Dimensional Flow Simulations on Distributed Computing Systems', enclosed as an appendix.
Warren, Gary
1988-01-01
The SOS code is used to compute the resonance modes (frequency-domain information) of sample devices and separately to compute the transient behavior of the same devices. A code, DOT, is created to compute appropriate dot products of the time-domain and frequency-domain results. The transient behavior of individual modes in the device is then plotted. Modes in a coupled-cavity traveling-wave tube (CCTWT) section excited beam in separate simulations are analyzed. Mode energy vs. time and mode phase vs. time are computed and it is determined whether the transient waves are forward or backward waves for each case. Finally, the hot-test mode frequencies of the CCTWT section are computed.
Multilevel criticality computations in AREVA NP'S core simulation code artemis - 195
International Nuclear Information System (INIS)
Van Geemert, R.
2010-01-01
This paper discusses the multi-level critical boron iteration approach that is applied per default in AREVA NP's whole-core neutronics and thermal hydraulics core simulation program ARTEMIS. This multi-level approach is characterized by the projection of variational boron concentration adjustments to the coarser mesh levels in a multi-level re-balancing hierarchy that is associated with the nodal flux equations to be solved in steady-state core simulation. At each individual re-balancing mesh level, optimized variational criticality tuning formulas are applied. The latter drive the core model to a numerically highly accurate self-sustaining state (i.e. with the neutronic eigenvalue being 1 up to a very high numerical precision) by continuous adjustment of the boron concentration as a system-wide scalar criticality parameter. Due to the default application of this approach in ARTEMIS reactor cycle simulations, an accuracy of all critical boron concentration estimates better than 0.001 ppm is enabled for all burnup time steps in a computationally efficient way. This high accuracy is relevant for precision optimization in industrial core simulation as well as for enabling accurate reactivity perturbation assessments. The developed approach is presented from a numerical methodology point of view with an emphasis on the multi-grid aspect of the concept. Furthermore, an application-relevant verification is presented in terms of achieved coupled iteration convergence efficiency for an application-representative industrial core cycle computation. (authors)
Validation and testing of the VAM2D computer code
International Nuclear Information System (INIS)
Kool, J.B.; Wu, Y.S.
1991-10-01
This document describes two modeling studies conducted by HydroGeoLogic, Inc. for the US NRC under contract no. NRC-04089-090, entitled, ''Validation and Testing of the VAM2D Computer Code.'' VAM2D is a two-dimensional, variably saturated flow and transport code, with applications for performance assessment of nuclear waste disposal. The computer code itself is documented in a separate NUREG document (NUREG/CR-5352, 1989). The studies presented in this report involve application of the VAM2D code to two diverse subsurface modeling problems. The first one involves modeling of infiltration and redistribution of water and solutes in an initially dry, heterogeneous field soil. This application involves detailed modeling over a relatively short, 9-month time period. The second problem pertains to the application of VAM2D to the modeling of a waste disposal facility in a fractured clay, over much larger space and time scales and with particular emphasis on the applicability and reliability of using equivalent porous medium approach for simulating flow and transport in fractured geologic media. Reflecting the separate and distinct nature of the two problems studied, this report is organized in two separate parts. 61 refs., 31 figs., 9 tabs
Computational Simulation on Electrowinning for Used LiCl-KCl salts
Energy Technology Data Exchange (ETDEWEB)
Sohn, Sung June; Kim, Pyeong Hwa; Hwang, Il Soon [KAERI, Daejeon (Korea, Republic of); Park, Jae Yeong [Korea Institute of Nuclear Safety, Daejoen (Korea, Republic of)
2016-05-15
That purification is consisted of electrowinning with liquid metal cathode and selective oxidation with chemical equilibrium by using metal chloride as an oxidizing agent. Actinides and rare earth elements are deposited to liquid cathode in electrowinning and rare earth elements are selectively extracted to molten salt, however, code posited Li react to oxidizing agent prior to rare earth elements which are intended to react in selective oxidation. Also if termination point of actinides deposition in electrowinning is clearly known, we would decrease amount of reacting rare earth elements as well as Li and throughput could be enhanced. For pyroprocess research computational simulation is important to save limited resources and research environment. This study shows computational modeling on electrowinning with Bi cathode by using electrochemical simulation code REFIN. This study shows that it is possible to simulate electrochemical behaviors of at least seven elements (excluding electrode and electrolyte materials) according to real time. In order to enhance accuracy of simulation results, it is suggested that combination of REFIN and CFD modeling on two immiscible liquid to calculate diffusion boundary layer thickness as well.
Benchmark Simulation for the Development of the Regulatory Audit Subchannel Analysis Code
Energy Technology Data Exchange (ETDEWEB)
Lee, G. H.; Song, C.; Woo, S. W. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)
2012-05-15
For the safe and reliable operation of a reactor, it is important to predict accurately the flow and temperature distributions in the thermal-hydraulic design of a reactor core. A subchannel approach can give the reasonable flow and temperature distributions with the short computing time. Korea Institute of Nuclear Safety (KINS) is presently reviewing new subchannel code, THALES, which will substitute for both THINC-IV and TORC code. To assess the prediction performance of THALES, KINS is developing the subchannel analysis code for the independent audit calculation. The code is based on workstation version of COBRA-IV-I. The main objective of the present study is to assess the performance of COBRA-IV-I code by comparing the simulation results with experimental ones for the sample problems
International Nuclear Information System (INIS)
Royl, P.
1981-01-01
PBDOWN is a computer code that simulates the blowdown of confined boiling materials ('pools') into a colder upper coolant plenum as time dependent ejection and expansion with consideration of a few selected exchange processes. Its application is restricted to situations resulting from hypothetical loss of flow (LOF) accidents in LMFBR's, where enough voiding has occured, that in core sodium vapor pressures become negligible. PBDOWN considers one working fluid for the discharge process (either fuel or steel) and a maximum of two working fluids (either fuel and sodium or steel and sodium) for the expansion process in the upper coolant plenum. Entrainment of sodium at the accelerated bubble liquid interfaces is mechanistically calculated by a Taylor instability entrainment model. Simulation of a hemispherical expansion form together with this mechanistic entrainment model gives a new integrated calculation of the time dependent sodium mass in the bubble. The paper summarizes the basic equations and assumptions of this computer model. Sample results compare different heat transfer and Na entrainment models during steel and fuel driven discharge processes. Mechanistic sodium entrainment simulation for SNR-type reactors coupled with a realistic heat transfer model is shown to reduce the integral mechanical work potential by a factor of 1.3 to 2.0 over the isentropic energy of the discharge working fluids. (orig.)
CEDNBR: a computer code for transient thermal margin analysis of a reactor core
International Nuclear Information System (INIS)
Shesler, A.T.; Lehmann, C.R.
1976-09-01
The report describes the CEDNBR computer code. This code was developed for the transient thermal analysis of a pressurized water reactor core or a critical heat flux test. Included are the code structure, conservation equations, and correlations utilized by CEDNBR. The methods of modelling a reactor core and hot channel and a CHF test are presented. Comparisons of CEDNBR calculations are made with both empirical pressure loss data and simulated loss of flow test data. The code solves the one-dimensional conservation of mass, energy, and momentum equations and the equation of state for the fluid for either steady-state or transient conditions. Tabular time dependent functions of inlet temperatures, pressure, mass velocity, axial heat flux distributions, normalized heat flux, radial peaking factors, and incremental mixing factors are required input to the code. Transient effects are included in the calculation of enthalpy rise and fluid properties. The Departure from Nucleate Boiling Ratio (DNBR) is calculated by applying a Critical Heat Flux (CHF) correlation to the computed local fluid properties. A code user's guide is provided for preparing input to the code. In addition, descriptions of the sub-routines used by CEDNBR are given
Benchmark testing and independent verification of the VS2DT computer code
International Nuclear Information System (INIS)
McCord, J.T.
1994-11-01
The finite difference flow and transport simulator VS2DT was benchmark tested against several other codes which solve the same equations (Richards equation for flow and the Advection-Dispersion equation for transport). The benchmark problems investigated transient two-dimensional flow in a heterogeneous soil profile with a localized water source at the ground surface. The VS2DT code performed as well as or better than all other codes when considering mass balance characteristics and computational speed. It was also rated highly relative to the other codes with regard to ease-of-use. Following the benchmark study, the code was verified against two analytical solutions, one for two-dimensional flow and one for two-dimensional transport. These independent verifications show reasonable agreement with the analytical solutions, and complement the one-dimensional verification problems published in the code's original documentation
International Nuclear Information System (INIS)
Chun, Moon-Hyun; Jeong, Eun-Soo
1983-01-01
A new computer code entitled KREWET has been developed in an effort to improve the accuracy and applicability of the existing reflood heat transfer simulation computer code. Sample calculations for temperature histories and heat transfer coefficient are made using KREWET code and the results are compared with the predictions of REFLUX, QUEN1D, and the PWR-FLECHT data for various conditions. These show favourable agreement in terms of clad temperature versus time. For high flooding rates (5-15cm/sec) and high pressure (∼413 Kpa), reflood predictions are reasonably well predicted by KREWET code as well as with other codes. For low flooding rates (less than ∼4cm/sec) and low pressure (∼138Kpa), predictions show considerable error in evaluating the rewet position versus time. This observation is common to all the codes examined in the present work
International Nuclear Information System (INIS)
Chun, M.-H.; Jeong, E.-S.
1983-01-01
A new computer code entitled KREWET has been developed in an effort to improve the accuracy and applicability of the existing reflood heat transfer simulation computer code. Sample calculations for temperature histories and heat transfer coefficient are made using KREWET code and the results are compared with the predictions of REFLUX, QUENID, and the PWR-FLECHT data for various conditions. These show favorable agreement in terms of clad temperature versus time. For high flooding rates (5-15cm/sec) and high pressure (approx. =413 Kpa), reflood predictions are reasonably well predicted by KREWET code as well as with other codes. For low flooding rates (less than approx. =4cm/sec) and low pressure (approx. =138 Kpa), predictions show considerable error in evaluating the rewet position versus time. This observation is common to all the codes examined in the present work
Simulation of radioactive waste transmutation on the t.node parallel computer
International Nuclear Information System (INIS)
Bacha, F.; Maillard, J.; Silva, J.
1995-01-01
Before any experiment on reactor driven by an accelerator, computer simulation supplies tools for optimization. Some of the key parameters are neutron production on a heavy target and neutronic distribution flux in the core. During two code benchmarks organized by the NEA-OECD, simulations of energetic incident proton collisions on a thin lead target for the first one, on a thick lead target for the second one, are described. One validation of the numeric codes is based on these results. A preliminary design of a burning waste system using benchmark result analysis and fission focused simulations is proposed
Simulation of radioactive waste transmutation on the T. Node parallel computer
International Nuclear Information System (INIS)
Bacha, F.; Maillard, J.; Silva, J.
1995-01-01
Before any experiment on reactor driven by an accelerator, computer simulation supplies tools for optimization. Some of the key parameters are neutron production on a heavy target and neutronic distribution flux in the core. During two code benchmarks organized by the NEA-OECD, simulations of energetic incident proton collisions on a thin lead target for the first one, on a thick lead target for the second one, are described. One validation of our numeric codes is based on these results. A preliminary design of a burning waste system using benchmark result analysis and fission focused simulations is proposed
Simulation of radioactive waste transmutation on the t.node parallel computer
Energy Technology Data Exchange (ETDEWEB)
Bacha, F.; Maillard, J.; Silva, J. [LPC College de France, Paris (France)
1995-10-01
Before any experiment on reactor driven by an accelerator, computer simulation supplies tools for optimization. Some of the key parameters are neutron production on a heavy target and neutronic distribution flux in the core. During two code benchmarks organized by the NEA-OECD, simulations of energetic incident proton collisions on a thin lead target for the first one, on a thick lead target for the second one, are described. One validation of the numeric codes is based on these results. A preliminary design of a burning waste system using benchmark result analysis and fission focused simulations is proposed.
Computer code MLCOSP for multiple-correlation and spectrum analysis with a hybrid computer
International Nuclear Information System (INIS)
Oguma, Ritsuo; Fujii, Yoshio; Usui, Hozumi; Watanabe, Koichi
1975-10-01
Usage of the computer code MLCOSP(Multiple Correlation and Spectrum) developed is described for a hybrid computer installed in JAERI Functions of the hybrid computer and its terminal devices are utilized ingeniously in the code to reduce complexity of the data handling which occurrs in analysis of the multivariable experimental data and to perform the analysis in perspective. Features of the code are as follows; Experimental data can be fed to the digital computer through the analog part of the hybrid computer by connecting with a data recorder. The computed results are displayed in figures, and hardcopies are taken when necessary. Series-messages to the code are shown on the terminal, so man-machine communication is possible. And further the data can be put in through a keyboard, so case study according to the results of analysis is possible. (auth.)
International Nuclear Information System (INIS)
Ford, R.L.; Nelson, W.R.
1978-06-01
A code to simulate almost any electron--photon transport problem conceivable is described. The report begins with a lengthy historical introduction and a description of the shower generation process. Then the detailed physics of the shower processes and the methods used to simulate them are presented. Ideas of sampling theory, transport techniques, particle interactions in general, and programing details are discussed. Next, EGS calculations and various experiments and other Monte Carlo results are compared. The remainder of the report consists of user manuals for EGS, PEGS, and TESTSR codes; options, input specifications, and typical output are included. 38 figures, 12 tables
Galerkin algorithm for multidimensional plasma simulation codes. Informal report
International Nuclear Information System (INIS)
Godfrey, B.B.
1979-03-01
A Galerkin finite element differencing scheme has been developed for a computer simulation of plasmas. The new difference equations identically satisfy an equation of continuity. Thus, the usual current correction procedure, involving inversion of Poisson's equation, is unnecessary. The algorithm is free of many numerical Cherenkov instabilities. This differencing scheme has been implemented in CCUBE, an already existing relativistic, electromagnetic, two-dimensional PIC code in arbitrary separable, orthogonal coordinates. The separability constraint is eliminated by the new algorithm. The new version of CCUBE exhibits good stability and accuracy with reduced computer memory and time requirements. Details of the algorithm and its implementation are presented
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.
FIRAC - a computer code to predict fire accident effects in nuclear facilities
International Nuclear Information System (INIS)
Bolstad, J.W.; Foster, R.D.; Gregory, W.S.
1983-01-01
FIRAC is a medium-sized computer code designed to predict fire-induced flows, temperatures, and material transport within the ventilating systems and other airflow pathways in nuclear-related facilities. The code is designed to analyze the behavior of interconnected networks of rooms and typical ventilation system components. This code is one in a family of computer codes that is designed to provide improved methods of safety analysis for the nuclear industry. The structure of this code closely follows that of the previously developed TVENT and EVENT codes. Because a lumped-parameter formulation is used, this code is particularly suitable for calculating the effects of fires in the far field (that is, in regions removed from the fire compartment), where the fire may be represented parametrically. However, a fire compartment model to simulate conditions in the enclosure is included. This model provides transport source terms to the ventilation system that can affect its operation and in turn affect the fire. A basic material transport capability that features the effects of convection, deposition, entrainment, and filtration of material is included. The interrelated effects of filter plugging, heat transfer, gas dynamics, and material transport are taken into account. In this paper the authors summarize the physical models used to describe the gas dynamics, material transport, and heat transfer processes. They also illustrate how a typical facility is modeled using the code
Turbo Pascal Computer Code for PIXE Analysis
International Nuclear Information System (INIS)
Darsono
2002-01-01
To optimal utilization of the 150 kV ion accelerator facilities and to govern the analysis technique using ion accelerator, the research and development of low energy PIXE technology has been done. The R and D for hardware of the low energy PIXE installation in P3TM have been carried on since year 2000. To support the R and D of PIXE accelerator facilities in harmonize with the R and D of the PIXE hardware, the development of PIXE software for analysis is also needed. The development of database of PIXE software for analysis using turbo Pascal computer code is reported in this paper. This computer code computes the ionization cross-section, the fluorescence yield, and the stopping power of elements also it computes the coefficient attenuation of X- rays energy. The computer code is named PIXEDASIS and it is part of big computer code planed for PIXE analysis that will be constructed in the near future. PIXEDASIS is designed to be communicative with the user. It has the input from the keyboard. The output shows in the PC monitor, which also can be printed. The performance test of the PIXEDASIS shows that it can be operated well and it can provide data agreement with data form other literatures. (author)
Users guide for NRC145-2 accident assessment computer code
International Nuclear Information System (INIS)
Pendergast, M.M.
1982-08-01
An accident assessment computer code has been developed for use at the Savannah River Plant. This computer code is based upon NRC Regulatory Guide 1.145 which provides guidence for accident assessements for power reactors. The code contains many options so that the user may utilize the code for many different assessments. For example the code can be used for non-nuclear assessments such as Sulpher Dioxide which may be required by the EPA. A discription of the code is contained in DP-1646. This document is a compilation of step-by-step instructions on how to use the code on the SRP IBM 3308 computer. This document consists of a number of tables which contain copies of computer listings. Some of the computer listings are copies of input; other listings give examples of computer output
International Nuclear Information System (INIS)
Shida, M.; Naito, M.; Suto, T.; Omori, E.; Nojiri, T.
2001-01-01
MIXSET is a FORTRAN code developed to simulate the Purex solvent extraction system using mixer-settler extractors. Japan Nuclear Cycle Development Institute (JNC) has been developing the MIXSET code since the years 1970 to analyze the behavior of nuclides in the solvent extraction processes in Tokai Reprocessing Plant (TRP). This paper describes the history of MIXSET code development, the features of the latest version, called MIXSET-X and the application of the code for safety evaluation work. (author)
SimProp: a simulation code for ultra high energy cosmic ray propagation
International Nuclear Information System (INIS)
Aloisio, R.; Grillo, A.F.; Boncioli, D.; Petrera, S.; Salamida, F.
2012-01-01
A new Monte Carlo simulation code for the propagation of Ultra High Energy Cosmic Rays is presented. The results of this simulation scheme are tested by comparison with results of another Monte Carlo computation as well as with the results obtained by directly solving the kinetic equation for the propagation of Ultra High Energy Cosmic Rays. A short comparison with the latest flux published by the Pierre Auger collaboration is also presented
Development of a dose assessment computer code for the NPP severe accident
International Nuclear Information System (INIS)
Cheong, Jae Hak
1993-02-01
A real-time emergency dose assessment computer code called KEDA (KAIST NPP Emergency Dose Assessment) has been developed for the NPP severe accident. A new mathematical model which can calculate cloud shine has been developed and implemented in the code. KEDA considers the specific Korean situations(complex topography, orientals' thyroid metabolism, continuous washout, etc.), and provides functions of dose-monitoring and automatic decision-making. To verify the code results, KEDA has been compared with an NRC officially certified code, RASCAL, for eight hypertical accident scenarios. Through the comparison, KEDA has been proved to provide reasonable results. Qualitative sensitivity analysis also the been performed for potentially important six input parameters, and the trends of the dose v.s. down-wind distance curve have been analyzed comparing with the physical phenomena occurred in the real atmosphere. The source term and meteorological conditions are turned out to be the most important input parameters. KEDA also has been applied to simulate Kori site and a hyperthetical accident with semi-real meteorological data has been simulated and analyzed
Computer simulation of sputtering: A review
International Nuclear Information System (INIS)
Robinson, M.T.; Hou, M.
1992-08-01
In 1986, H. H. Andersen reviewed attempts to understand sputtering by computer simulation and identified several areas where further research was needed: potential energy functions for molecular dynamics (MD) modelling; the role of inelastic effects on sputtering, especially near the target surface; the modelling of surface binding in models based on the binary collision approximation (BCA); aspects of cluster emission in MD models; and angular distributions of sputtered particles. To these may be added kinetic energy distributions of sputtered particles and the relationships between MD and BCA models, as well as the development of intermediate models. Many of these topics are discussed. Recent advances in BCA modelling include the explicit evaluation of the time in strict BCA codes and the development of intermediate codes able to simulate certain many-particle problems realistically. Developments in MD modelling include the wide-spread use of many-body potentials in sputtering calculations, inclusion of realistic electron excitation and electron-phonon interactions, and several studies of cluster ion impacts on solid surfaces
Computer codes used in particle accelerator design: First edition
International Nuclear Information System (INIS)
1987-01-01
This paper contains a listing of more than 150 programs that have been used in the design and analysis of accelerators. Given on each citation are person to contact, classification of the computer code, publications describing the code, computer and language runned on, and a short description of the code. Codes are indexed by subject, person to contact, and code acronym
MISER-I: a computer code for JOYO fuel management
International Nuclear Information System (INIS)
Yamashita, Yoshioki
1976-06-01
A computer code ''MISER-I'' is for a nuclear fuel management of Japan Experimental Fast Breeder Reactor JOYO. The nuclear fuel management in JOYO can be regarded as a fuel assembly management because a handling unit of fuel in JOYO plant is a fuel subassembly (core and blanket subassembly), and so the recording of material balance in computer code is made with each subassembly. The input information into computer code is given with each subassembly for a transfer operation, or with one reactor cycle and every one month for a burn-up in reactor core. The output information of MISER-I code is the fuel assembly storage record, fuel storage weight record in each material balance subarea at any specified day, and fuel subassembly transfer history record. Change of nuclear fuel composition and weight due to a burn-up is calculated with JOYO-Monitoring Code by off-line computation system. MISER-I code is written in FORTRAN-IV language for FACOM 230-48 computer. (auth.)
Massively parallel computation of PARASOL code on the Origin 3800 system
International Nuclear Information System (INIS)
Hosokawa, Masanari; Takizuka, Tomonori
2001-10-01
The divertor particle simulation code named PARASOL simulates open-field plasmas between divertor walls self-consistently by using an electrostatic PIC method and a binary collision Monte Carlo model. The PARASOL parallelized with MPI-1.1 for scalar parallel computer worked on Intel Paragon XP/S system. A system SGI Origin 3800 was newly installed (May, 2001). The parallel programming was improved at this switchover. As a result of the high-performance new hardware and this improvement, the PARASOL is speeded up by about 60 times with the same number of processors. (author)
3D code for simulations of fluid flows
International Nuclear Information System (INIS)
Skandera, D.
2004-01-01
In this paper, a present status in the development of the new numerical code is reported. The code is considered for simulations of fluid flows. The finite volume approach is adopted for solving standard fluid equations. They are treated in a conservative form to ensure a correct conservation of fluid quantities. Thus, a nonlinear hyperbolic system of conservation laws is numerically solved. The code uses the Eulerian description of the fluid and is designed as a high order central numerical scheme. The central approach employs no (approximate) Riemann solver and is less computational expensive. The high order WENO strategy is adopted in the reconstruction step to achieve results comparable with more accurate Riemann solvers. A combination of the central approach with an iterative solving of a local Riemann problem is tested and behaviour of such numerical flux is reported. An extension to three dimensions is implemented using a dimension by dimension approach, hence, no complicated dimensional splitting need to be introduced. The code is fully parallelized with the MPI library. Several standard hydrodynamic tests in one, two and three dimensions were performed and their results are presented. (author)
Hamor-2: a computer code for LWR inventory calculation
International Nuclear Information System (INIS)
Guimaraes, L.N.F.; Marzo, M.A.S.
1985-01-01
A method for calculating the accuracy inventory of LWR reactors is presented. This method uses the Hamor-2 computer code. Hamor-2 is obtained from the coupling of two other computer codes Hammer-Techion and Origen-2 for testing Hamor-2, its results were compared to concentration values measured from activides of two PWR reactors; Kernkraftwerk Obrighein (KWO) and H.B. Robinson (HBR). These actinides are U 235 , U 236 , U 238 , Pu 239 , Pu 241 and PU 242 . The computer code Hammor-2 shows better results than the computer code Origem-2, when both are compared with experimental results. (E.G.) [pt
Simulation of small break loss of coolant accident using relap 5/ MOD 2 computer code
International Nuclear Information System (INIS)
Megahed, M.M.
1992-01-01
An assessment of relap 5 / MOD 2/Cycle 36.05 best estimate computer code capabilities in predicting the thermohydraulic response of a PWR following a small break loss of coolant accident is presented. The experimental data base for the evaluation is the results of Test S-N H-3 performed in the semi scale MOD-2 c Test facility which modeled a 0.5% small break loss of coolant accident with an accompanying failure of the high pressure injection emergency core cooling system. A conclusion was reached that the code is capable of making small break loss of coolant accident calculations efficiently. However, some of the small break loss of coolant accident related phenomena were not properly predicted by the code, suggesting a need for code improvement.9 fig., 3 tab
Schnek: A C++ library for the development of parallel simulation codes on regular grids
Schmitz, Holger
2018-05-01
A large number of algorithms across the field of computational physics are formulated on grids with a regular topology. We present Schnek, a library that enables fast development of parallel simulations on regular grids. Schnek contains a number of easy-to-use modules that greatly reduce the amount of administrative code for large-scale simulation codes. The library provides an interface for reading simulation setup files with a hierarchical structure. The structure of the setup file is translated into a hierarchy of simulation modules that the developer can specify. The reader parses and evaluates mathematical expressions and initialises variables or grid data. This enables developers to write modular and flexible simulation codes with minimal effort. Regular grids of arbitrary dimension are defined as well as mechanisms for defining physical domain sizes, grid staggering, and ghost cells on these grids. Ghost cells can be exchanged between neighbouring processes using MPI with a simple interface. The grid data can easily be written into HDF5 files using serial or parallel I/O.
Implementation of a pressurized water reactor simulator for teaching on a mini-computer
International Nuclear Information System (INIS)
Tallec, Michele.
1982-06-01
This paper presents the design of a pressurized water reactor power plant simulator using a mini-computer. This simulator is oriented towards teaching. It operates real-time simulations and many parameters can be changed by the student during execution of the digital code. First, a state variable model of the dynamic behavior of the plant is derived from the physical laws. The second part presents the problems associated with the use of a mini-computer for the resolution of a large differential system, notably the problems of memory-space availability, execution time and numerical integration. Finally, it contains the description of the control deck outlay used to interfer with the digital code, and of the the conditions that can be changed during an excution [fr
Computer codes for level 1 probabilistic safety assessment
International Nuclear Information System (INIS)
1990-06-01
Probabilistic Safety Assessment (PSA) entails several laborious tasks suitable for computer codes assistance. This guide identifies these tasks, presents guidelines for selecting and utilizing computer codes in the conduct of the PSA tasks and for the use of PSA results in safety management and provides information on available codes suggested or applied in performing PSA in nuclear power plants. The guidance is intended for use by nuclear power plant system engineers, safety and operating personnel, and regulators. Large efforts are made today to provide PC-based software systems and PSA processed information in a way to enable their use as a safety management tool by the nuclear power plant overall management. Guidelines on the characteristics of software needed for management to prepare a software that meets their specific needs are also provided. Most of these computer codes are also applicable for PSA of other industrial facilities. The scope of this document is limited to computer codes used for the treatment of internal events. It does not address other codes available mainly for the analysis of external events (e.g. seismic analysis) flood and fire analysis. Codes discussed in the document are those used for probabilistic rather than for phenomenological modelling. It should be also appreciated that these guidelines are not intended to lead the user to selection of one specific code. They provide simply criteria for the selection. Refs and tabs
Comparison of TITAN hybrid deterministic transport code and MCNP5 for simulation of SPECT
International Nuclear Information System (INIS)
Royston, K.; Haghighat, A.; Yi, C.
2010-01-01
Traditionally, Single Photon Emission Computed Tomography (SPECT) simulations use Monte Carlo methods. The hybrid deterministic transport code TITAN has recently been applied to the simulation of a SPECT myocardial perfusion study. The TITAN SPECT simulation uses the discrete ordinates formulation in the phantom region and a simplified ray-tracing formulation outside of the phantom. A SPECT model has been created in the Monte Carlo Neutral particle (MCNP)5 Monte Carlo code for comparison. In MCNP5 the collimator is directly modeled, but TITAN instead simulates the effect of collimator blur using a circular ordinate splitting technique. Projection images created using the TITAN code are compared to results using MCNP5 for three collimator acceptance angles. Normalized projection images for 2.97 deg, 1.42 deg and 0.98 deg collimator acceptance angles had maximum relative differences of 21.3%, 11.9% and 8.3%, respectively. Visually the images are in good agreement. Profiles through the projection images were plotted to find that the TITAN results followed the shape of the MCNP5 results with some differences in magnitude. A timing comparison on 16 processors found that the TITAN code completed the calculation 382 to 2787 times faster than MCNP5. Both codes exhibit good parallel performance. (author)
A study on the nuclear computer code maintenance and management system
International Nuclear Information System (INIS)
Kim, Yeon Seung; Huh, Young Hwan; Lee, Jong Bok; Choi, Young Gil; Suh, Soong Hyok; Kang, Byong Heon; Kim, Hee Kyung; Kim, Ko Ryeo; Park, Soo Jin
1990-12-01
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)
Simulation-based computation of dose to humans in radiological environments
International Nuclear Information System (INIS)
Breazeal, N.L.; Davis, K.R.; Watson, R.A.; Vickers, D.S.; Ford, M.S.
1996-03-01
The Radiological Environment Modeling System (REMS) quantifies dose to humans working in radiological environments using the IGRIP (Interactive Graphical Robot Instruction Program) and Deneb/ERGO simulation software. These commercially available products are augmented with custom C code to provide radiation exposure information to, and collect radiation dose information from, workcell simulations. Through the use of any radiation transport code or measured data, a radiation exposure input database may be formulated. User-specified IGRIP simulations utilize these databases to compute and accumulate dose to programmable human models operating around radiation sources. Timing, distances, shielding, and human activity may be modeled accurately in the simulations. The accumulated dose is recorded in output files, and the user is able to process and view this output. The entire REMS capability can be operated from a single graphical user interface
Simulation-based computation of dose to humans in radiological environments
Energy Technology Data Exchange (ETDEWEB)
Breazeal, N.L. [Sandia National Labs., Livermore, CA (United States); Davis, K.R.; Watson, R.A. [Sandia National Labs., Albuquerque, NM (United States); Vickers, D.S. [Brigham Young Univ., Provo, UT (United States). Dept. of Electrical and Computer Engineering; Ford, M.S. [Battelle Pantex, Amarillo, TX (United States). Dept. of Radiation Safety
1996-03-01
The Radiological Environment Modeling System (REMS) quantifies dose to humans working in radiological environments using the IGRIP (Interactive Graphical Robot Instruction Program) and Deneb/ERGO simulation software. These commercially available products are augmented with custom C code to provide radiation exposure information to, and collect radiation dose information from, workcell simulations. Through the use of any radiation transport code or measured data, a radiation exposure input database may be formulated. User-specified IGRIP simulations utilize these databases to compute and accumulate dose to programmable human models operating around radiation sources. Timing, distances, shielding, and human activity may be modeled accurately in the simulations. The accumulated dose is recorded in output files, and the user is able to process and view this output. The entire REMS capability can be operated from a single graphical user interface.
Annealing simulation of cascade damage using MARLOWE-DAIQUIRI codes
International Nuclear Information System (INIS)
Muroga, Takeo
1984-01-01
The localization effect of the defects generated by the cascade damage on the properties of solids was studied by using a computer code. The code is based on the two-body collision approximation method and the Monte Carlo method. The MARLOWE and DAIQUIRI codes were partly improved to fit the present calculation of the annealing of cascade damage. The purpose of this study is to investigate the behavior of defects under the simulated reactive and irradiation condition. Calculation was made for alpha iron (BCC), and the threshold energy was set at 40 eV. The temperature dependence of annealing and the growth of a cluster were studied. The overlapping effect of cascade was studied. At first, the extreme case of overlapping was studied, then the practical cases were estimated by interpolation. The state of overlapping of cascade corresponded to the irradiation speed. The interaction between cascade and dislocations was studied, and the calculation of the annealing of primary knock-out atoms (PKA) in alpha iron was performed. At low temperature, the effect of dislocations was large, but the growth of vacancy was not seen. At high temperature, the effect of dislocations was small. The evaluation of the simulation of various ion irradiation and the growth efficiency of defects were performed. (Kato, T.)
International Nuclear Information System (INIS)
Bourauel, Peter; Nabbi, Rahim; Biel, Wolfgang; Forrest, Robin
2009-01-01
The MCNP 3D Monte Carlo computer code is used not only for criticality calculations of nuclear systems but also to simulate transports of radiation and particles. The findings so obtained about neutron flux distribution and the associated spectra allow information about materials activation, nuclear heating, and radiation damage to be obtained by means of activation codes such as FISPACT. The stochastic character of particle and radiation transport processes normally links findings to the materials cells making up the geometry model of MCNP. Where high spatial resolution is required for the activation calculations with FISPACT, fine segmentation of the MCNP geometry becomes compulsory, which implies considerable expense for the modeling process. For this reason, an alternative simulation technique has been developed in an effort to automate and optimize data transfer between MCNP and FISPACT. (orig.)
HUDU: The Hanford Unified Dose Utility computer code
International Nuclear Information System (INIS)
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
Coded aperture optimization using Monte Carlo simulations
International Nuclear Information System (INIS)
Martineau, A.; Rocchisani, J.M.; Moretti, J.L.
2010-01-01
Coded apertures using Uniformly Redundant Arrays (URA) have been unsuccessfully evaluated for two-dimensional and three-dimensional imaging in Nuclear Medicine. The images reconstructed from coded projections contain artifacts and suffer from poor spatial resolution in the longitudinal direction. We introduce a Maximum-Likelihood Expectation-Maximization (MLEM) algorithm for three-dimensional coded aperture imaging which uses a projection matrix calculated by Monte Carlo simulations. The aim of the algorithm is to reduce artifacts and improve the three-dimensional spatial resolution in the reconstructed images. Firstly, we present the validation of GATE (Geant4 Application for Emission Tomography) for Monte Carlo simulations of a coded mask installed on a clinical gamma camera. The coded mask modelling was validated by comparison between experimental and simulated data in terms of energy spectra, sensitivity and spatial resolution. In the second part of the study, we use the validated model to calculate the projection matrix with Monte Carlo simulations. A three-dimensional thyroid phantom study was performed to compare the performance of the three-dimensional MLEM reconstruction with conventional correlation method. The results indicate that the artifacts are reduced and three-dimensional spatial resolution is improved with the Monte Carlo-based MLEM reconstruction.
DEFF Research Database (Denmark)
Sessarego, Matias; Ramos García, Néstor; Sørensen, Jens Nørkær
2017-01-01
Aerodynamic and structural dynamic performance analysis of modern wind turbines are routinely estimated in the wind energy field using computational tools known as aeroelastic codes. Most aeroelastic codes use the blade element momentum (BEM) technique to model the rotor aerodynamics and a modal......, multi-body or the finite-element approach to model the turbine structural dynamics. The present work describes the development of a novel aeroelastic code that combines a three-dimensional viscous–inviscid interactive method, method for interactive rotor aerodynamic simulations (MIRAS...... Code Comparison Collaboration Project. Simulation tests consist of steady wind inflow conditions with different combinations of yaw error, wind shear, tower shadow and turbine-elastic modeling. Turbulent inflow created by using a Mann box is also considered. MIRAS-FLEX results, such as blade tip...
International Nuclear Information System (INIS)
2013-03-01
It is fundamental to the future of nuclear power that reactors can be run safely and economically to compete with other forms of power generation. As a consequence, it is essential to develop the understanding of fuel performance and to embody that knowledge in codes to provide best estimate predictions of fuel behaviour. This in turn leads to a better understanding of fuel performance, a reduction in operating margins, flexibility in fuel management and improved operating economics. The IAEA has therefore embarked on a series of programmes addressing different aspects of fuel behaviour modelling with the following objectives: - To assess the maturity and prediction capabilities of fuel performance codes, and to support interaction and information exchange between countries with code development and application needs (FUMEX series); - To build a database of well defined experiments suitable for code validation in association with the OECD Nuclear Energy Agency (OECD/NEA); - To transfer a mature fuel modelling code to developing countries, to support teams in these countries in their efforts to adapt the code to the requirements of particular reactors, and to provide guidance on applying the code to reactor operation and safety assessments; - To provide guidelines for code quality assurance, code licensing and code application to fuel licensing. This report describes the results of the coordinated research project on the ''Improvement of computer codes used for fuel behaviour simulation (FUMEX-III)''. This programme was initiated in 2008 and completed in 2012. It followed previous programmes on fuel modelling: D-COM 1982-1984, FUMEX 1993-1996 and FUMEX-II 2002-2006. The participants used a mixture of data derived from commercial and experimental irradiation histories, in particular data designed to investigate the mechanical interactions occurring in fuel during normal, transient and severe transient operation. All participants carried out calculations on priority
Evaluation of SPACE code for simulation of inadvertent opening of spray valve in Shin Kori unit 1
International Nuclear Information System (INIS)
Kim, Seyun; Youn, Bumsoo
2013-01-01
SPACE code is expected to be applied to the safety analysis for LOCA (Loss of Coolant Accident) and Non-LOCA scenarios. SPACE code solves two-fluid, three-field governing equations and programmed with C++ computer language using object-oriented concepts. To evaluate the analysis capability for the transient phenomena in the actual nuclear power plant, an inadvertent opening of spray valve in startup test phase of Shin Kori unit 1 was simulated with SPACE code. To evaluate the analysis capability for the transient phenomena in the actual nuclear power plant, an inadvertent opening of spray valve in startup test phase of Shin Kori unit 1 was simulated with SPACE code
Comparisons of 'Identical' Simulations by the Eulerian Gyrokinetic Codes GS2 and GYRO
Bravenec, R. V.; Ross, D. W.; Candy, J.; Dorland, W.; McKee, G. R.
2003-10-01
A major goal of the fusion program is to be able to predict tokamak transport from first-principles theory. To this end, the Eulerian gyrokinetic code GS2 was developed years ago and continues to be improved [1]. Recently, the Eulerian code GYRO was developed [2]. These codes are not subject to the statistical noise inherent to particle-in-cell (PIC) codes, and have been very successful in treating electromagnetic fluctuations. GS2 is fully spectral in the radial coordinate while GYRO uses finite-differences and ``banded" spectral schemes. To gain confidence in nonlinear simulations of experiment with these codes, ``apples-to-apples" comparisons (identical profile inputs, flux-tube geometry, two species, etc.) are first performed. We report on a series of linear and nonlinear comparisons (with overall agreement) including kinetic electrons, collisions, and shaped flux surfaces. We also compare nonlinear simulations of a DIII-D discharge to measurements of not only the fluxes but also the turbulence parameters. [1] F. Jenko, et al., Phys. Plasmas 7, 1904 (2000) and refs. therein. [2] J. Candy, J. Comput. Phys. 186, 545 (2003).
An improved UO2 thermal conductivity model in the ELESTRES computer code
International Nuclear Information System (INIS)
Chassie, G.G.; Tochaie, M.; Xu, Z.
2010-01-01
This paper describes the improved UO 2 thermal conductivity model for use in the ELESTRES (ELEment Simulation and sTRESses) computer code. The ELESTRES computer code models the thermal, mechanical and microstructural behaviour of a CANDU® fuel element under normal operating conditions. The main purpose of the code is to calculate fuel temperatures, fission gas release, internal gas pressure, fuel pellet deformation, and fuel sheath strains for fuel element design and assessment. It is also used to provide initial conditions for evaluating fuel behaviour during high temperature transients. The thermal conductivity of UO 2 fuel is one of the key parameters that affect ELESTRES calculations. The existing ELESTRES thermal conductivity model has been assessed and improved based on a large amount of thermal conductivity data from measurements of irradiated and un-irradiated UO 2 fuel with different densities. The UO 2 thermal conductivity data cover 90% to 99% theoretical density of UO 2 , temperature up to 3027 K, and burnup up to 1224 MW·h/kg U. The improved thermal conductivity model, which is recommended for a full implementation in the ELESTRES computer code, has reduced the ELESTRES code prediction biases of temperature, fission gas release, and fuel sheath strains when compared with the available experimental data. This improved thermal conductivity model has also been checked with a test version of ELESTRES over the full ranges of fuel temperature, fuel burnup, and fuel density expected in CANDU fuel. (author)
Computer simulation of gain fluctuations in proportional counters
International Nuclear Information System (INIS)
Demir, Nelgun; Tapan, . Ilhan
2004-01-01
A computer simulation code has been developed in order to examine the fluctuation in gas amplification in wire proportional counters which are common in detector applications in particle physics experiments. The magnitude of the variance in the gain dominates the statistical portion of the energy resolution. In order to compare simulation and experimental results, the gain and its variation has been calculated numerically for the well known Aleph Inner Tracking Detector geometry. The results show that the bias voltage has a strong influence on the variance in the gain. The simulation calculations are in good agreement with experimental results. (authors)
SIMULATE-3 K coupled code applications
Energy Technology Data Exchange (ETDEWEB)
Joensson, Christian [Studsvik Scandpower AB, Vaesteraas (Sweden); Grandi, Gerardo; Judd, Jerry [Studsvik Scandpower Inc., Idaho Falls, ID (United States)
2017-07-15
This paper describes the coupled code system TRACE/SIMULATE-3 K/VIPRE and the application of this code system to the OECD PWR Main Steam Line Break. A short description is given for the application of the coupled system to analyze DNBR and the flexibility the system creates for the user. This includes the possibility to compare and evaluate the result with the TRACE/SIMULATE-3K (S3K) coupled code, the S3K standalone code (core calculation) as well as performing single-channel calculations with S3K and VIPRE. This is the typical separate-effect-analyses required for advanced calculations in order to develop methodologies to be used for safety analyses in general. The models and methods of the code systems are presented. The outline represents the analysis approach starting with the coupled code system, reactor and core model calculation (TRACE/S3K). This is followed by a more detailed core evaluation (S3K standalone) and finally a very detailed thermal-hydraulic investigation of the hot pin condition (VIPRE).
Comparative simulation of Stirling and Sibling cycle cryocoolers with two codes
International Nuclear Information System (INIS)
Mitchell, M.P.; Wilson, K.J.; Bauwens, L.
1989-01-01
The authors present a comparative analysis of Stirling and Sibling Cycle cryocoolers conducted with two different computer simulation codes. One code (CRYOWEISS) performs an initial analysis on the assumption of isothermal conditions in the machines and adjusts that result with decoupled loss calculations. The other code (MS*2) models fluid flows and heat transfers more realistically but ignores significant loss mechanisms, including flow friction and heat conduction through the metal of the machines. Surprisingly, MS*2 is less optimistic about performance of all machines even though it ignores losses that are modelled by CRYOWEISS. Comparison between constant-bore Stirling and Sibling machines shows that their performance is generally comparable over a range of temperatures, pressures and operating speeds. No machine was consistently superior or inferior according to both codes over the whole range of conditions studied
Parallelization of a Monte Carlo particle transport simulation code
Hadjidoukas, P.; Bousis, C.; Emfietzoglou, D.
2010-05-01
We have developed a high performance version of the Monte Carlo particle transport simulation code MC4. The original application code, developed in Visual Basic for Applications (VBA) for Microsoft Excel, was first rewritten in the C programming language for improving code portability. Several pseudo-random number generators have been also integrated and studied. The new MC4 version was then parallelized for shared and distributed-memory multiprocessor systems using the Message Passing Interface. Two parallel pseudo-random number generator libraries (SPRNG and DCMT) have been seamlessly integrated. The performance speedup of parallel MC4 has been studied on a variety of parallel computing architectures including an Intel Xeon server with 4 dual-core processors, a Sun cluster consisting of 16 nodes of 2 dual-core AMD Opteron processors and a 200 dual-processor HP cluster. For large problem size, which is limited only by the physical memory of the multiprocessor server, the speedup results are almost linear on all systems. We have validated the parallel implementation against the serial VBA and C implementations using the same random number generator. Our experimental results on the transport and energy loss of electrons in a water medium show that the serial and parallel codes are equivalent in accuracy. The present improvements allow for studying of higher particle energies with the use of more accurate physical models, and improve statistics as more particles tracks can be simulated in low response time.
Flight code validation simulator
Sims, Brent A.
1996-05-01
An End-To-End Simulation capability for software development and validation of missile flight software on the actual embedded computer has been developed utilizing a 486 PC, i860 DSP coprocessor, embedded flight computer and custom dual port memory interface hardware. This system allows real-time interrupt driven embedded flight software development and checkout. The flight software runs in a Sandia Digital Airborne Computer and reads and writes actual hardware sensor locations in which Inertial Measurement Unit data resides. The simulator provides six degree of freedom real-time dynamic simulation, accurate real-time discrete sensor data and acts on commands and discretes from the flight computer. This system was utilized in the development and validation of the successful premier flight of the Digital Miniature Attitude Reference System in January of 1995 at the White Sands Missile Range on a two stage attitude controlled sounding rocket.
Nonuniform code concatenation for universal fault-tolerant quantum computing
Nikahd, Eesa; Sedighi, Mehdi; Saheb Zamani, Morteza
2017-09-01
Using transversal gates is a straightforward and efficient technique for fault-tolerant quantum computing. Since transversal gates alone cannot be computationally universal, they must be combined with other approaches such as magic state distillation, code switching, or code concatenation to achieve universality. In this paper we propose an alternative approach for universal fault-tolerant quantum computing, mainly based on the code concatenation approach proposed in [T. Jochym-O'Connor and R. Laflamme, Phys. Rev. Lett. 112, 010505 (2014), 10.1103/PhysRevLett.112.010505], but in a nonuniform fashion. The proposed approach is described based on nonuniform concatenation of the 7-qubit Steane code with the 15-qubit Reed-Muller code, as well as the 5-qubit code with the 15-qubit Reed-Muller code, which lead to two 49-qubit and 47-qubit codes, respectively. These codes can correct any arbitrary single physical error with the ability to perform a universal set of fault-tolerant gates, without using magic state distillation.
Computer simulation of replacement sequences in copper
International Nuclear Information System (INIS)
Schiffgens, J.O.; Schwartz, D.W.; Ariyasu, R.G.; Cascadden, S.E.
1978-01-01
Results of computer simulations of , , and replacement sequences in copper are presented, including displacement thresholds, focusing energies, energy losses per replacement, and replacement sequence lengths. These parameters are tabulated for six interatomic potentials and shown to vary in a systematic way with potential stiffness and range. Comparisons of results from calculations made with ADDES, a quasi-dynamical code, and COMENT, a dynamical code, show excellent agreement, demonstrating that the former can be calibrated and used satisfactorily in the analysis of low energy displacement cascades. Upper limits on , , and replacement sequences were found to be approximately 10, approximately 30, and approximately 14 replacements, respectively. (author)
Computer codes for ventilation in nuclear facilities
International Nuclear Information System (INIS)
Mulcey, P.
1987-01-01
In this paper the authors present some computer codes, developed in the last years, for ventilation and radioprotection. These codes are used for safety analysis in the conception, exploitation and dismantlement of nuclear facilities. The authors present particularly: DACC1 code used for aerosol deposit in sampling circuit of radiation monitors; PIAF code used for modelization of complex ventilation system; CLIMAT 6 code used for optimization of air conditioning system [fr
The GBS code for tokamak scrape-off layer simulations
International Nuclear Information System (INIS)
Halpern, F.D.; Ricci, P.; Jolliet, S.; Loizu, J.; Morales, J.; Mosetto, A.; Musil, F.; Riva, F.; Tran, T.M.; Wersal, C.
2016-01-01
We describe a new version of GBS, a 3D global, flux-driven plasma turbulence code to simulate the turbulent dynamics in the tokamak scrape-off layer (SOL), superseding the code presented by Ricci et al. (2012) [14]. The present work is driven by the objective of studying SOL turbulent dynamics in medium size tokamaks and beyond with a high-fidelity physics model. We emphasize an intertwining framework of improved physics models and the computational improvements that allow them. The model extensions include neutral atom physics, finite ion temperature, the addition of a closed field line region, and a non-Boussinesq treatment of the polarization drift. GBS has been completely refactored with the introduction of a 3-D Cartesian communicator and a scalable parallel multigrid solver. We report dramatically enhanced parallel scalability, with the possibility of treating electromagnetic fluctuations very efficiently. The method of manufactured solutions as a verification process has been carried out for this new code version, demonstrating the correct implementation of the physical model.
International Nuclear Information System (INIS)
Al-Ghorabie, F.H.H.
2003-01-01
In this paper a computer model based on the use of the well-known Monte Carlo simulation code EGS4 was developed to simulate the scattering of polyenergetic X-ray beams through some materials. These materials are: lucite, polyethylene, polypropylene and aluminium. In particular, the ratio of the scattered to total X-ray fluence (scatter fraction) has been calculated for X-ray beams in the energy region 30-120 keV. In addition scatter fractions have been determined experimentally using a polyenergetic superficial X-ray unit. Comparison of the measured and the calculated results has been performed. The Monte Carlo calculations have also been carried out for water, bakelite and bone to examine the dependence of scatter fraction on the density of the scatterer. Good agreement (estimated statistical error < 5%) was obtained between the measured and the calculated values of the scatter fractions for materials with Z < 20 that were studied in this paper. Copyright (2003) Australasian College of Physical Scientists and Engineers in Medicine
Automated uncertainty analysis methods in the FRAP computer codes
International Nuclear Information System (INIS)
Peck, S.O.
1980-01-01
A user oriented, automated uncertainty analysis capability has been incorporated in the Fuel Rod Analysis Program (FRAP) computer codes. The FRAP codes have been developed for the analysis of Light Water Reactor fuel rod behavior during steady state (FRAPCON) and transient (FRAP-T) conditions as part of the United States Nuclear Regulatory Commission's Water Reactor Safety Research Program. The objective of uncertainty analysis of these codes is to obtain estimates of the uncertainty in computed outputs of the codes is to obtain estimates of the uncertainty in computed outputs of the codes as a function of known uncertainties in input variables. This paper presents the methods used to generate an uncertainty analysis of a large computer code, discusses the assumptions that are made, and shows techniques for testing them. An uncertainty analysis of FRAP-T calculated fuel rod behavior during a hypothetical loss-of-coolant transient is presented as an example and carried through the discussion to illustrate the various concepts
International Nuclear Information System (INIS)
Raisali, G.R.
1992-01-01
A series of computer codes based on point kernel technique and also Monte Carlo method have been developed. These codes perform radiation transport calculations for irradiator systems having cartesian, cylindrical and mixed geometries. The monte Carlo calculations, the computer code 'EGS4' has been applied to a radiation processing type problem. This code has been acompanied by a specific user code. The set of codes developed include: GCELLS, DOSMAPM, DOSMAPC2 which simulate the radiation transport in gamma irradiator systems having cylinderical, cartesian, and mixed geometries, respectively. The program 'DOSMAP3' based on point kernel technique, has been also developed for dose rate mapping calculations in carrier type gamma irradiators. Another computer program 'CYLDETM' as a user code for EGS4 has been also developed to simulate dose variations near the interface of heterogeneous media in gamma irradiator systems. In addition a system of computer codes 'PRODMIX' has been developed which calculates the absorbed dose in the products with different densities. validation studies of the calculated results versus experimental dosimetry has been performed and good agreement has been obtained
Study of nuclear computer code maintenance and management system
International Nuclear Information System (INIS)
Ryu, Chang Mo; Kim, Yeon Seung; Eom, Heung Seop; Lee, Jong Bok; Kim, Ho Joon; Choi, Young Gil; Kim, Ko Ryeo
1989-01-01
Software maintenance is one of the most important problems since late 1970's.We wish to develop a nuclear computer code system to maintenance and manage KAERI's nuclear software. As a part of this system, we have developed three code management programs for use on CYBER and PC systems. They are used in systematic management of computer code in KAERI. The first program is embodied on the CYBER system to rapidly provide information on nuclear codes to the users. The second and the third programs were embodied on the PC system for the code manager and for the management of data in korean language, respectively. In the requirement analysis, we defined each code, magnetic tape, manual and abstract information data. In the conceptual design, we designed retrieval, update, and output functions. In the implementation design, we described the technical considerations of database programs, utilities, and directions for the use of databases. As a result of this research, we compiled the status of nuclear computer codes which belonged KAERI until September, 1988. Thus, by using these three database programs, we could provide the nuclear computer code information to the users more rapidly. (Author)
Implantation of FRAPCON-2 code in HB computer
International Nuclear Information System (INIS)
Silva, C.F. da.
1987-05-01
The modifications carried out for implanting FRAPCON-2 computer code in the HB DPS-T7 computer are presented. The FRAPCON-2 code calculates thermo-mechanical response during long period of burnup in stationary state for fuel rods of PWR type reactors. (M.C.K.)
Energy Technology Data Exchange (ETDEWEB)
Ferrouk, M. [Laboratoire du Genie Physique des Hydrocarbures, University of Boumerdes, Boumerdes 35000 (Algeria)], E-mail: m_ferrouk@yahoo.fr; Aissani, S. [Laboratoire du Genie Physique des Hydrocarbures, University of Boumerdes, Boumerdes 35000 (Algeria); D' Auria, F.; DelNevo, A.; Salah, A. Bousbia [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Universita di Pisa (Italy)
2008-10-15
The present article covers the evaluation of the performance of twelve critical heat flux methods/correlations published in the open literature. The study concerns the simulation of an axially non-uniform heat flux distribution with the RELAP5 computer code in a single boiling water reactor channel benchmark problem. The nodalization scheme employed for the considered particular geometry, as modelled in RELAP5 code, is described. For this purpose a review of critical heat flux models/correlations applicable to non-uniform axial heat profile is provided. Simulation results using the RELAP5 code and those obtained from our computer program, based on three type predictions methods such as local conditions, F-factor and boiling length average approaches were compared.
International Nuclear Information System (INIS)
Eslinger, Paul W.; Arimescu, Carmen; Kanyid, Beverly A.; Miley, Terri B.
2001-01-01
One activity of the Department of Energy?s Groundwater/Vadose Zone Integration Project is an assessment of cumulative impacts from Hanford Site wastes on the subsurface environment and the Columbia River. Through the application of a system assessment capability (SAC), decisions for each cleanup and disposal action will be able to take into account the composite effect of other cleanup and disposal actions. The SAC has developed a suite of computer programs to simulate the migration of contaminants (analytes) present on the Hanford Site and to assess the potential impacts of the analytes, including dose to humans, socio-cultural impacts, economic impacts, and ecological impacts. The general approach to handling uncertainty in the SAC computer codes is a Monte Carlo approach. Conceptually, one generates a value for every stochastic parameter in the code (the entire sequence of modules from inventory through transport and impacts) and then executes the simulation, obtaining an output value, or result. This document provides user instructions for the SAC codes that generate human, ecological, economic, and cultural impacts
Development of steam explosion simulation code JASMINE
Energy Technology Data Exchange (ETDEWEB)
Moriyama, Kiyofumi; Yamano, Norihiro; Maruyama, Yu; Kudo, Tamotsu; Sugimoto, Jun [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nagano, Katsuhiro; Araki, Kazuhiro
1995-11-01
A steam explosion is considered as a phenomenon which possibly threatens the integrity of the containment vessel of a nuclear power plant in a severe accident condition. A numerical calculation code JASMINE (JAeri Simulator for Multiphase INteraction and Explosion) purposed to simulate the whole process of steam explosions has been developed. The premixing model is based on a multiphase flow simulation code MISTRAL by Fuji Research Institute Co. In JASMINE code, the constitutive equations and the flow regime map are modified for the simulation of premixing related phenomena. The numerical solution method of the original code is succeeded, i.e. the basic equations are discretized semi-implicitly, BCGSTAB method is used for the matrix solver to improve the stability and convergence, also TVD scheme is applied to capture a steep phase distribution accurately. Test calculations have been performed for the conditions correspond to the experiments by Gilbertson et al. and Angelini et al. in which mixing of solid particles and water were observed in iso-thermal condition and with boiling, respectively. (author).
Development of steam explosion simulation code JASMINE
International Nuclear Information System (INIS)
Moriyama, Kiyofumi; Yamano, Norihiro; Maruyama, Yu; Kudo, Tamotsu; Sugimoto, Jun; Nagano, Katsuhiro; Araki, Kazuhiro.
1995-11-01
A steam explosion is considered as a phenomenon which possibly threatens the integrity of the containment vessel of a nuclear power plant in a severe accident condition. A numerical calculation code JASMINE (JAeri Simulator for Multiphase INteraction and Explosion) purposed to simulate the whole process of steam explosions has been developed. The premixing model is based on a multiphase flow simulation code MISTRAL by Fuji Research Institute Co. In JASMINE code, the constitutive equations and the flow regime map are modified for the simulation of premixing related phenomena. The numerical solution method of the original code is succeeded, i.e. the basic equations are discretized semi-implicitly, BCGSTAB method is used for the matrix solver to improve the stability and convergence, also TVD scheme is applied to capture a steep phase distribution accurately. Test calculations have been performed for the conditions correspond to the experiments by Gilbertson et al. and Angelini et al. in which mixing of solid particles and water were observed in iso-thermal condition and with boiling, respectively. (author)
International Nuclear Information System (INIS)
Sharma, Manish; Pilkhwal, D.S.; Vijayan, P.K.; Saha, D.; Sinha, R.K.
2002-06-01
Occurrence of instability in natural circulation loops can lead to problems in control and occurrence of critical heat flux (CHF) during low flow periods. Remaining within an identified stable zone operation is therefore desirable. Natural circulation loops can pass through an unstable zone during start-up and power raising. In the present work RELAPS / MOD 3.2 computer code has been used to simulate the unstable oscillatory behavior observed in a rectangular natural circulation loop having horizontal heater and horizontal cooler (HHHC) orientation. The results were compared with the experimental data. This report describes the nodalization scheme adopted tor this work and results of the analysis in detail. (author)
The community project COSA: comparison of geo-mechanical computer codes for salt
International Nuclear Information System (INIS)
Lowe, M.J.S.; Knowles, N.C.
1986-01-01
Two benchmark problems related to waste disposal in salt were tackled by ten European organisations using twelve rock-mechanics finite element computer codes. The two problems represented increasing complexity with first a hypothetical verification and then the simulation of a laboratory experiment. The project allowed to ascertain a shapshot of the current combined expertise of European organisations in the modelling of salt behaviour
International Nuclear Information System (INIS)
Rohmann, D.; Koehler, T.
1987-02-01
This is a description of the computer code FIT, written in FORTRAN-77 for a PDP 11/34. FIT is an interactive program to decude position, width and intensity of lines of X-ray spectra (max. length of 4K channels). The lines (max. 30 lines per fit) may have Gauss- or Voigt-profile, as well as exponential tails. Spectrum and fit can be displayed on a Tektronix terminal. (orig.) [de
Auxiliary plasma heating and fueling models for use in particle simulation codes
International Nuclear Information System (INIS)
Procassini, R.J.; Cohen, B.I.
1989-01-01
Computational models of a radiofrequency (RF) heating system and neutral-beam injector are presented. These physics packages, when incorporated into a particle simulation code allow one to simulate the auxiliary heating and fueling of fusion plasmas. The RF-heating package is based upon a quasilinear diffusion equation which describes the slow evolution of the heated particle distribution. The neutral-beam injector package models the charge exchange and impact ionization processes which transfer energy and particles from the beam to the background plasma. Particle simulations of an RF-heated and a neutral-beam-heated simple-mirror plasma are presented. 8 refs., 5 figs
Computer simulation of driven Alfven waves
International Nuclear Information System (INIS)
Geary, J.L. Jr.
1986-01-01
The first particle simulation study of shear Alfven wave resonance heating is presented. Particle simulation codes self-consistently follow the time evolution of the individual and collective aspects of particle dynamics as well as wave dynamics in a fully nonlinear fashion. Alfven wave heating is a possible means of increasing the temperature of magnetized plasmas. A new particle simulation model was developed for this application that incorporates Darwin's formulation of the electromagnetic fields with a guiding center approximation for electron motion perpendicular to the ambient magnetic field. The implementation of this model and the examination of its theoretical and computational properties are presented. With this model, several cases of Alfven wave heating is examined in both uniform and nonuniform simulation systems in a two dimensional slab. For the inhomogeneous case studies, the kinetic Alfven wave develops in the vicinity of the shear Alfven resonance region
Energy Technology Data Exchange (ETDEWEB)
Kang, D.W.; Chi, J.H.; Goh, E.O. [Korea Electric Power Research Institute, Taejon (Korea)
2001-07-01
A computer program, RASSAY was developed to evaluate accurately the activities of various nuclides in the rad-waste container for Ulchin units 3 and 4. This is the final report of the project, {sup D}evelopment of the Computer Code to Determine an Individual Radionuclides in the Rad-wastes Container for Ulchin Units 3 and 4 and includes the followings; 1) Structure of the computer code, RASSAY 2) An example of surface dose calculation by computer simulation using MCNP code 3) Methods of sampling and activity measurement of various Rad-wastes. (author). 21 refs., 35 figs., 6 tabs.
Comparison of SISEC code simulations with earthquake data of ordinary and base-isolated buildings
International Nuclear Information System (INIS)
Wang, C.Y.; Gvildys, J.
1991-01-01
At Argonne National Laboratory (ANL), a 3-D computer program SISEC (Seismic Isolation System Evaluation Code) is being developed for simulating the system response of isolated and ordinary structures (Wang et al. 1991). This paper describes comparison of SISEC code simulations with building response data of actual earthquakes. To ensure the accuracy of analytical simulations, recorded data of full-size reinforced concrete structures located in Sendai, Japan are used in this benchmark comparison. The test structures consist of two three-story buildings, one base-isolated and the other one ordinary founded. They were constructed side by side to investigate the effect of base isolation on the acceleration response. Among 20 earthquakes observed since April 1989, complete records of three representative earthquakes, no.2, no.6, and no.17, are used for the code validation presented in this paper. Correlations of observed and calculated accelerations at all instrument locations are made. Also, relative response characteristics of ordinary and isolated building structures are investigated. (J.P.N.)
Overview of the Tusas Code for Simulation of Dendritic Solidification
Energy Technology Data Exchange (ETDEWEB)
Trainer, Amelia J. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Newman, Christopher Kyle [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Francois, Marianne M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2016-01-07
The aim of this project is to conduct a parametric investigation into the modeling of two dimensional dendrite solidification, using the phase field model. Specifically, we use the Tusas code, which is for coupled heat and phase-field simulation of dendritic solidification. Dendritic solidification, which may occur in the presence of an unstable solidification interface, results in treelike microstructures that often grow perpendicular to the rest of the growth front. The interface may become unstable if the enthalpy of the solid material is less than that of the liquid material, or if the solute is less soluble in solid than it is in liquid, potentially causing a partition [1]. A key motivation behind this research is that a broadened understanding of phase-field formulation and microstructural developments can be utilized for macroscopic simulations of phase change. This may be directly implemented as a part of the Telluride project at Los Alamos National Laboratory (LANL), through which a computational additive manufacturing simulation tool is being developed, ultimately to become part of the Advanced Simulation and Computing Program within the U.S. Department of Energy [2].
International Nuclear Information System (INIS)
2004-08-01
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
International Nuclear Information System (INIS)
Lee, Joo Hee; Kim, Tae Han; Choi, Hae Yun; Lee, Kwang Won; Chung, Chang Kyu
2007-01-01
Safety Injection Tanks (SITs) with the Fluidic Device (FD) of APR1400 provides a means of rapid reflooding of the core following a large break Loss Of Coolant Accident (LOCA), and keeping it covered until flow from the Safety Injection Pump (SIP) becomes available. A passive FD can provide two operation stages of a safety water injection into the RCS and allow more effective use of borated water in case of LOCA. Once a large break LOCA occurs, the system will deliver a high flow rate of cooling water for a certain period of time, and thereafter, the flow rate is reduced to a lower flow rate. The conventional computer code 'TURTLE' used to simulate the blowdown of OPR1000 SIT can not be directly applied to simulate a blowdown process of the SIT with FD. A new computer code is needed to be developed for the blowdown test evaluation of the APR1400 SIT with FD. Korea Power Engineering Company (KOPEC) has developed a new computer code to analyze the characteristics of the SIT with FD and validated the code through the comparison of the calculation results with the test results obtained by Ulchin 5 and 6 units pre-operational test and VAlve Performance Evaluation Rig (VAPER) tests performed by The Korea Atomic Energy Research Institute (KAERI)
Utilization of Relap 5 computer code for analyzing thermohydraulic projects
International Nuclear Information System (INIS)
Silva Filho, E.
1987-01-01
This work deals with the design of a scaled test facility of a typical pressurized water reactor plant of the 1300 MW (electric) class. A station blackout has been choosen to investigate the thermohydraulic behaviour of the the test facility in comparison to the reactor plant. The computer code RELAPS/MOD1 has been utilized to simulate the blackout and to compare the test facility behaviour with the reactor plant one. The results demonstrate similar thermohydraulic behaviours of the two systems. (author) [pt
RIVER-RAD: A computer code for simulating the transport of radionuclides in rivers
International Nuclear Information System (INIS)
Hetrick, D.M.; McDowell-Boyer, L.M.; Sjoreen, A.L.; Thorne, D.J.; Patterson, M.R.
1992-11-01
A screening-level model, RIVER-RAD, has been developed to assess the potential fate of radionuclides released to rivers. The model is simplified in nature and is intended to provide guidance in determining the potential importance of the surface water pathway, relevant transport mechanisms, and key radionuclides in estimating radiological dose to man. The purpose of this report is to provide a description of the model and a user's manual for the FORTRAN computer code
Computer simulation of two-phase flow in nuclear reactors
International Nuclear Information System (INIS)
Wulff, W.
1993-01-01
Two-phase flow models dominate the requirements of economic resources for the development and use of computer codes which serve to analyze thermohydraulic transients in nuclear power plants. An attempt is made to reduce the effort of analyzing reactor transients by combining purpose-oriented modelling with advanced computing techniques. Six principles are presented on mathematical modeling and the selection of numerical methods, along with suggestions on programming and machine selection, all aimed at reducing the cost of analysis. Computer simulation is contrasted with traditional computer calculation. The advantages of run-time interactive access operation in a simulation environment are demonstrated. It is explained that the drift-flux model is better suited than the two-fluid model for the analysis of two-phase flow in nuclear reactors, because of the latter's closure problems. The advantage of analytical over numerical integration is demonstrated. Modeling and programming techniques are presented which minimize the number of needed arithmetical and logical operations and thereby increase the simulation speed, while decreasing the cost. (orig.)
Investigations into radiation damages of reactor materials by computer simulation
International Nuclear Information System (INIS)
Bronnikov, V.A.
2004-01-01
Data on the state of works in European countries in the field of computerized simulation of radiation damages of reactor materials under the context of the international projects ITEM (European Database for Multiscale Modelling) and SIRENA (Simulation of Radiation Effects in Zr-Nb alloys) - computerized simulation of stress corrosion when contact of Zr-Nb alloys with iodine are presented. Computer codes for the simulation of radiation effects in reactor materials were developed. European Database for Multiscale Modelling (EDAM) was organized using the results of the investigations provided in the ITEM project [ru
Development of HTGR plant dynamics simulation code
International Nuclear Information System (INIS)
Ohashi, Kazutaka; Tazawa, Yujiro; Mitake, Susumu; Suzuki, Katsuo.
1987-01-01
Plant dynamics simulation analysis plays an important role in the design work of nuclear power plant especially in the plant safety analysis, control system analysis, and transient condition analysis. The authors have developed the plant dynamics simulation code named VESPER, which is applicable to the design work of High Temperature Engineering Test Reactor, and have been improving the code corresponding to the design changes made in the subsequent design works. This paper describes the outline of VESPER code and shows its sample calculation results selected from the recent design work. (author)
Vrnak, Daniel R.; Stueber, Thomas J.; Le, Dzu K.
2012-01-01
This report presents a method for running a dynamic legacy inlet simulation in concert with another dynamic simulation that uses a graphical interface. The legacy code, NASA's LArge Perturbation INlet (LAPIN) model, was coded using the FORTRAN 77 (The Portland Group, Lake Oswego, OR) programming language to run in a command shell similar to other applications that used the Microsoft Disk Operating System (MS-DOS) (Microsoft Corporation, Redmond, WA). Simulink (MathWorks, Natick, MA) is a dynamic simulation that runs on a modern graphical operating system. The product of this work has both simulations, LAPIN and Simulink, running synchronously on the same computer with periodic data exchanges. Implementing the method described in this paper avoided extensive changes to the legacy code and preserved its basic operating procedure. This paper presents a novel method that promotes inter-task data communication between the synchronously running processes.
Leckey, Cara A C; Wheeler, Kevin R; Hafiychuk, Vasyl N; Hafiychuk, Halyna; Timuçin, Doğan A
2018-03-01
Ultrasonic wave methods constitute the leading physical mechanism for nondestructive evaluation (NDE) and structural health monitoring (SHM) of solid composite materials, such as carbon fiber reinforced polymer (CFRP) laminates. Computational models of ultrasonic wave excitation, propagation, and scattering in CFRP composites can be extremely valuable in designing practicable NDE and SHM hardware, software, and methodologies that accomplish the desired accuracy, reliability, efficiency, and coverage. The development and application of ultrasonic simulation approaches for composite materials is an active area of research in the field of NDE. This paper presents comparisons of guided wave simulations for CFRP composites implemented using four different simulation codes: the commercial finite element modeling (FEM) packages ABAQUS, ANSYS, and COMSOL, and a custom code executing the Elastodynamic Finite Integration Technique (EFIT). Benchmark comparisons are made between the simulation tools and both experimental laser Doppler vibrometry data and theoretical dispersion curves. A pristine and a delamination type case (Teflon insert in the experimental specimen) is studied. A summary is given of the accuracy of simulation results and the respective computational performance of the four different simulation tools. Published by Elsevier B.V.
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.
Quantum computing with Majorana fermion codes
Litinski, Daniel; von Oppen, Felix
2018-05-01
We establish a unified framework for Majorana-based fault-tolerant quantum computation with Majorana surface codes and Majorana color codes. All logical Clifford gates are implemented with zero-time overhead. This is done by introducing a protocol for Pauli product measurements with tetrons and hexons which only requires local 4-Majorana parity measurements. An analogous protocol is used in the fault-tolerant setting, where tetrons and hexons are replaced by Majorana surface code patches, and parity measurements are replaced by lattice surgery, still only requiring local few-Majorana parity measurements. To this end, we discuss twist defects in Majorana fermion surface codes and adapt the technique of twist-based lattice surgery to fermionic codes. Moreover, we propose a family of codes that we refer to as Majorana color codes, which are obtained by concatenating Majorana surface codes with small Majorana fermion codes. Majorana surface and color codes can be used to decrease the space overhead and stabilizer weight compared to their bosonic counterparts.
Two-dimensional color-code quantum computation
International Nuclear Information System (INIS)
Fowler, Austin G.
2011-01-01
We describe in detail how to perform universal fault-tolerant quantum computation on a two-dimensional color code, making use of only nearest neighbor interactions. Three defects (holes) in the code are used to represent logical qubits. Triple-defect logical qubits are deformed into isolated triangular sections of color code to enable transversal implementation of all single logical qubit Clifford group gates. Controlled-NOT (CNOT) is implemented between pairs of triple-defect logical qubits via braiding.
Lightweight computational steering of very large scale molecular dynamics simulations
International Nuclear Information System (INIS)
Beazley, D.M.
1996-01-01
We present a computational steering approach for controlling, analyzing, and visualizing very large scale molecular dynamics simulations involving tens to hundreds of millions of atoms. Our approach relies on extensible scripting languages and an easy to use tool for building extensions and modules. The system is extremely easy to modify, works with existing C code, is memory efficient, and can be used from inexpensive workstations and networks. We demonstrate how we have used this system to manipulate data from production MD simulations involving as many as 104 million atoms running on the CM-5 and Cray T3D. We also show how this approach can be used to build systems that integrate common scripting languages (including Tcl/Tk, Perl, and Python), simulation code, user extensions, and commercial data analysis packages
Computational simulation of Argonauta/IEN nuclear reactor using MCNPX code
International Nuclear Information System (INIS)
Cunha, Victor Lusis Lassance; Silva Junior, Wilson F. Rebello da
2011-01-01
The study consisted of developing a computer simulation of a nuclear research reactor using the MCNPX. The reactor modeled is the Argonauta located at IEN (Rio de Janeiro) designed by Argonne National Laboratory (USA), which is primarily used for non-destructive testing with neutron beam and teaching purposes. It was entirely modeled with geometric fidelity, including detailed material description, shielding and irradiation channels. When available, the model was based on the as-built drawings. Four different simulations were made, the first set of two for criticality calculations and the other set for flux measurement. The first simulation set consisted of estimating the reactors reactivity. The second set consisted of placing detectors on specific places where the reactor is monitored and on the fuel axis covering the multiplicative and non-multiplicative media. Based on this data, the thermal neutron flux profile was plotted. All the outputs were compared with experimental data. Since it is a stochastic method, the statistical convergence was successfully checked for all simulations. The results were in good agreement with the experimental values. For the criticality calculations, the relative error was smaller then 1%. The flux measurements were also very well reproduced. The values were normalized for a reference point and the proportionality between the different spots was respected. The neutron flux profile along the core had the expected shape and values. Based on the good results, it can be said that the model is validated. (author)
High performance stream computing for particle beam transport simulations
International Nuclear Information System (INIS)
Appleby, R; Bailey, D; Higham, J; Salt, M
2008-01-01
Understanding modern particle accelerators requires simulating charged particle transport through the machine elements. These simulations can be very time consuming due to the large number of particles and the need to consider many turns of a circular machine. Stream computing offers an attractive way to dramatically improve the performance of such simulations by calculating the simultaneous transport of many particles using dedicated hardware. Modern Graphics Processing Units (GPUs) are powerful and affordable stream computing devices. The results of simulations of particle transport through the booster-to-storage-ring transfer line of the DIAMOND synchrotron light source using an NVidia GeForce 7900 GPU are compared to the standard transport code MAD. It is found that particle transport calculations are suitable for stream processing and large performance increases are possible. The accuracy and potential speed gains are compared and the prospects for future work in the area are discussed
Energy Technology Data Exchange (ETDEWEB)
Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.; Stuart, D.S.; Thompson, S.L. [Sandia National Labs., Albuquerque, NM (United States); Hodge, S.A.; Hyman, C.R.; Sanders, R.L. [Oak Ridge National Lab., TN (United States)
1995-03-01
MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.
International Nuclear Information System (INIS)
Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.; Stuart, D.S.; Thompson, S.L.; Hodge, S.A.; Hyman, C.R.; Sanders, R.L.
1995-03-01
MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR's phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package
A friend man-machine interface for thermo-hydraulic simulation codes of nuclear installations
International Nuclear Information System (INIS)
Araujo Filho, F. de; Belchior Junior, A.; Barroso, A.C.O.; Gebrim, A.
1994-01-01
This work presents the development of a Man-Machine Interface to the TRAC-PF1 code, a computer program to perform best estimate analysis of transients and accidents at nuclear power plants. The results were considered satisfactory and a considerable productivity gain was achieved in the activity of preparing and analyzing simulations. (author)
The archaeology of computer codes - illustrated on the basis of the code SABINE
International Nuclear Information System (INIS)
Sdouz, G.
1987-02-01
Computer codes used by the physics group of the Institute for Reactor Safety are stored on back-up-tapes. However during the last years both the computer and the system have been changed. For new tasks these programmes have to be available. A new procedure is necessary to find and to activate a stored programme. This procedure is illustrated on the basis of the code SABINE. (Author)
SUMMARY OF GENERAL WORKING GROUP A+B+D: CODES BENCHMARKING.
Energy Technology Data Exchange (ETDEWEB)
WEI, J.; SHAPOSHNIKOVA, E.; ZIMMERMANN, F.; HOFMANN, I.
2006-05-29
Computer simulation is an indispensable tool in assisting the design, construction, and operation of accelerators. In particular, computer simulation complements analytical theories and experimental observations in understanding beam dynamics in accelerators. The ultimate function of computer simulation is to study mechanisms that limit the performance of frontier accelerators. There are four goals for the benchmarking of computer simulation codes, namely debugging, validation, comparison and verification: (1) Debugging--codes should calculate what they are supposed to calculate; (2) Validation--results generated by the codes should agree with established analytical results for specific cases; (3) Comparison--results from two sets of codes should agree with each other if the models used are the same; and (4) Verification--results from the codes should agree with experimental measurements. This is the summary of the joint session among working groups A, B, and D of the HI32006 Workshop on computer codes benchmarking.
Reactor safety computer code development at INEL
International Nuclear Information System (INIS)
Johnsen, G.W.
1985-01-01
This report provides a brief overview of the computer code development programs being conducted at EG and G Idaho, Inc. on behalf of US Nuclear Regulatory Commission and the Department of Energy, Idaho Operations Office. Included are descriptions of the codes being developed, their development status as of the date of this report, and resident code development expertise
Penelope-2006: a code system for Monte Carlo simulation of electron and photon transport
International Nuclear Information System (INIS)
2006-01-01
The computer code system PENELOPE (version 2006) performs Monte Carlo simulation of coupled electron-photon transport in arbitrary materials for a wide energy range, from a few hundred eV to about 1 GeV. Photon transport is simulated by means of the standard, detailed simulation scheme. Electron and positron histories are generated on the basis of a mixed procedure, which combines detailed simulation of hard events with condensed simulation of soft interactions. A geometry package called PENGEOM permits the generation of random electron-photon showers in material systems consisting of homogeneous bodies limited by quadric surfaces, i.e. planes, spheres, cylinders, etc. This report is intended not only to serve as a manual of the PENELOPE code system, but also to provide the user with the necessary information to understand the details of the Monte Carlo algorithm. These proceedings contain the corresponding manual and teaching notes of the PENELOPE-2006 workshop and training course, held on 4-7 July 2006 in Barcelona, Spain. (author)
SKEMA - A computer code to estimate atmospheric dispersion
International Nuclear Information System (INIS)
Sacramento, A.M. do.
1985-01-01
This computer code is a modified version of DWNWND code, developed in Oak Ridge National Laboratory. The Skema code makes an estimative of concentration in air of a material released in atmosphery, by ponctual source. (C.M.) [pt
Fast acceleration of 2D wave propagation simulations using modern computational accelerators.
Directory of Open Access Journals (Sweden)
Wei Wang
Full Text Available Recent developments in modern computational accelerators like Graphics Processing Units (GPUs and coprocessors provide great opportunities for making scientific applications run faster than ever before. However, efficient parallelization of scientific code using new programming tools like CUDA requires a high level of expertise that is not available to many scientists. This, plus the fact that parallelized code is usually not portable to different architectures, creates major challenges for exploiting the full capabilities of modern computational accelerators. In this work, we sought to overcome these challenges by studying how to achieve both automated parallelization using OpenACC and enhanced portability using OpenCL. We applied our parallelization schemes using GPUs as well as Intel Many Integrated Core (MIC coprocessor to reduce the run time of wave propagation simulations. We used a well-established 2D cardiac action potential model as a specific case-study. To the best of our knowledge, we are the first to study auto-parallelization of 2D cardiac wave propagation simulations using OpenACC. Our results identify several approaches that provide substantial speedups. The OpenACC-generated GPU code achieved more than 150x speedup above the sequential implementation and required the addition of only a few OpenACC pragmas to the code. An OpenCL implementation provided speedups on GPUs of at least 200x faster than the sequential implementation and 30x faster than a parallelized OpenMP implementation. An implementation of OpenMP on Intel MIC coprocessor provided speedups of 120x with only a few code changes to the sequential implementation. We highlight that OpenACC provides an automatic, efficient, and portable approach to achieve parallelization of 2D cardiac wave simulations on GPUs. Our approach of using OpenACC, OpenCL, and OpenMP to parallelize this particular model on modern computational accelerators should be applicable to other
International Nuclear Information System (INIS)
1996-05-01
This publication contains the final papers summarizing the validation of the codes on the basis of comparison of observed effects with computer simulated effects on reactor cores from seismic disturbances. Refs, figs tabs
Development of computer code in PNC, 3
International Nuclear Information System (INIS)
Ohtaki, Akira; Ohira, Hiroaki
1990-01-01
Super-COPD, a code which is integrated by calculation modules, has been developed in order to evaluate kinds of dynamics of LMFBR plant by improving COPD. The code involves all models and its advanced models of COPD in module structures. The code makes it possible to simulate the system dynamics of LMFBR plant of any configurations and components. (author)
Multiphase integral reacting flow computer code (ICOMFLO): User`s guide
Energy Technology Data Exchange (ETDEWEB)
Chang, S.L.; Lottes, S.A.; Petrick, M.
1997-11-01
A copyrighted computational fluid dynamics computer code, ICOMFLO, has been developed for the simulation of multiphase reacting flows. The code solves conservation equations for gaseous species and droplets (or solid particles) of various sizes. General conservation laws, expressed by elliptic type partial differential equations, are used in conjunction with rate equations governing the mass, momentum, enthalpy, species, turbulent kinetic energy, and turbulent dissipation. Associated phenomenological submodels of the code include integral combustion, two parameter turbulence, particle evaporation, and interfacial submodels. A newly developed integral combustion submodel replacing an Arrhenius type differential reaction submodel has been implemented to improve numerical convergence and enhance numerical stability. A two parameter turbulence submodel is modified for both gas and solid phases. An evaporation submodel treats not only droplet evaporation but size dispersion. Interfacial submodels use correlations to model interfacial momentum and energy transfer. The ICOMFLO code solves the governing equations in three steps. First, a staggered grid system is constructed in the flow domain. The staggered grid system defines gas velocity components on the surfaces of a control volume, while the other flow properties are defined at the volume center. A blocked cell technique is used to handle complex geometry. Then, the partial differential equations are integrated over each control volume and transformed into discrete difference equations. Finally, the difference equations are solved iteratively by using a modified SIMPLER algorithm. The results of the solution include gas flow properties (pressure, temperature, density, species concentration, velocity, and turbulence parameters) and particle flow properties (number density, temperature, velocity, and void fraction). The code has been used in many engineering applications, such as coal-fired combustors, air
Computer code qualification program for the Advanced CANDU Reactor
International Nuclear Information System (INIS)
Popov, N.K.; Wren, D.J.; Snell, V.G.; White, A.J.; Boczar, P.G.
2003-01-01
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)
Development of NSSS Simulation Engine for SMART Simulator Using the Best Estimate Code, MARS3.1
International Nuclear Information System (INIS)
Kim, K. D.; Lee, S. W.; Lee, Sung Chul; Suh, Yong Suk; Suh, Jae Seung
2011-01-01
Limited computational capability and crude thermalhydraulic modeling in early 1980s forced the use of overly simplified physical models and assumptions for a real-time calculation at the cost of fidelity. Rapid advances in computer technology make it possible to improve the fidelity of the simulator models. These efforts have been made based on RELAP5 in the US, and CATHARE2 in France. The NSSS thermalhydraulic engines adopted in the most domestic fullscope power plant simulators have been replaced with RELAP5 based engines which were provided by US vendors. Since the technology dependency of the NSSS T/H engine by foreign vendors, it may cause difficulties in maintenance and model improvement. KAERI has started to develop a realistic NSSS calculation engine based on the best-estimate code MARS 3.1 for the SMART full-scope simulator. Even though we are developing the NSSS calculation engine for SMART simulator, it can be easily extended to light water reactors and GEN-IV reactors, etc. The verification of the NSSS calculation engine for SMART simulator has been conducted by an integrated test in the simulator environment, Jade 4.0, developed by GSE of Windows 2003. This paper briefly presents our efforts for the NSSS calculation engine for SMART simulator and verification test results of SAT (Site Acceptance Test)
Computer simulation of displacement cascades in copper
International Nuclear Information System (INIS)
Heinisch, H.L.
1983-06-01
More than 500 displacement cascades in copper have been generated with the computer simulation code MARLOWE over an energy range pertinent to both fission and fusion neutron spectra. Three-dimensional graphical depictions of selected cascades, as well as quantitative analysis of cascade shapes and sizes and defect densities, illustrate cascade behavior as a function of energy. With increasing energy, the transition from production of single compact damage regions to widely spaced multiple damage regions is clearly demonstrated
Integration of adaptive process control with computational simulation for spin-forming
International Nuclear Information System (INIS)
Raboin, P. J. LLNL
1998-01-01
Improvements in spin-forming capabilities through upgrades to a metrology and machine control system and advances in numerical simulation techniques were studied in a two year project funded by Laboratory Directed Research and Development (LDRD) at Lawrence Livermore National Laboratory. Numerical analyses were benchmarked with spin-forming experiments and computational speeds increased sufficiently to now permit actual part forming simulations. Extensive modeling activities examined the simulation speeds and capabilities of several metal forming computer codes for modeling flat plate and cylindrical spin-forming geometries. Shape memory research created the first numerical model to describe this highly unusual deformation behavior in Uranium alloys. A spin-forming metrology assessment led to sensor and data acquisition improvements that will facilitate future process accuracy enhancements, such as a metrology frame. Finally, software improvements (SmartCAM) to the manufacturing process numerically integrate the part models to the spin-forming process and to computational simulations
COOL: A code for Dynamic Monte Carlo Simulation of molecular dynamics
Barletta, Paolo
2012-02-01
Cool is a program to simulate evaporative and sympathetic cooling for a mixture of two gases co-trapped in an harmonic potential. The collisions involved are assumed to be exclusively elastic, and losses are due to evaporation from the trap. Each particle is followed individually in its trajectory, consequently properties such as spatial densities or energy distributions can be readily evaluated. The code can be used sequentially, by employing one output as input for another run. The code can be easily generalised to describe more complicated processes, such as the inclusion of inelastic collisions, or the possible presence of more than two species in the trap. New version program summaryProgram title: COOL Catalogue identifier: AEHJ_v2_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEHJ_v2_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 1 097 733 No. of bytes in distributed program, including test data, etc.: 18 425 722 Distribution format: tar.gz Programming language: C++ Computer: Desktop Operating system: Linux RAM: 500 Mbytes Classification: 16.7, 23 Catalogue identifier of previous version: AEHJ_v1_0 Journal reference of previous version: Comput. Phys. Comm. 182 (2011) 388 Does the new version supersede the previous version?: Yes Nature of problem: Simulation of the sympathetic process occurring for two molecular gases co-trapped in a deep optical trap. Solution method: The Direct Simulation Monte Carlo method exploits the decoupling, over a short time period, of the inter-particle interaction from the trapping potential. The particle dynamics is thus exclusively driven by the external optical field. The rare inter-particle collisions are considered with an acceptance/rejection mechanism, that is, by comparing a random number to the collisional probability
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.)
Interface design of VSOP'94 computer code for safety analysis
International Nuclear Information System (INIS)
Natsir, Khairina; Andiwijayakusuma, D.; Wahanani, Nursinta Adi; Yazid, Putranto Ilham
2014-01-01
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
Interface design of VSOP'94 computer code for safety analysis
Natsir, Khairina; Yazid, Putranto Ilham; Andiwijayakusuma, D.; Wahanani, Nursinta Adi
2014-09-01
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
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.)
A computer code for analysis of severe accidents in LWRs
International Nuclear Information System (INIS)
2001-01-01
The ICARE2 computer code, developed and validated since 1988 at IPSN (nuclear safety and protection institute), calculates in a mechanistic way the physical and chemical phenomena involved in the core degradation process during possible severe accidents in LWR's. The coupling between ICARE2 and the best-estimate thermal-hydraulics code CATHARE2 was completed at IPSN and led to the release of a first ICARE/CATHARE V1 version in 1999, followed by 2 successive revisions in 2000 and 2001. This documents gathers all the contributions presented at the first international ICARE/CATHARE users'club seminar that took place in November 2001. This seminar was characterized by a high quality and variety of the presentations, showing an increase of reactor applications and user needs in this area (2D/3D aspects, reflooding, corium slumping into the lower head,...). 2 sessions were organized. The first one was dedicated to the applications of ICARE2 V3mod1 against small-scale experiments such as PHEBUS FPT2 and FPT3 tests, PHEBUS AIC, QUENCH experiments, NRU-FLHT-5 test, ACRR-MP1 and DC1 experiments, CORA-PWR tests, and PBF-SFD1.4 test. The second session involved ICARE/CATHARE V1mod1 reactor applications and users'guidelines. Among reactor applications we found: code applicability to high burn-up fuel rods, simulation of the TMI-2 transient, simulation of a PWR-900 high pressure severe accident sequence, and the simulation of a VVER-1000 large break LOCA scenario. (A.C.)
Neoclassical Simulation of Tokamak Plasmas using Continuum Gyrokinetc Code TEMPEST
International Nuclear Information System (INIS)
Xu, X Q
2007-01-01
We present gyrokinetic neoclassical simulations of tokamak plasmas with self-consistent electric field for the first time using a fully nonlinear (full-f) continuum code TEMPEST in a circular geometry. A set of gyrokinetic equations are discretized on a five dimensional computational grid in phase space. The present implementation is a Method of Lines approach where the phase-space derivatives are discretized with finite differences and implicit backwards differencing formulas are used to advance the system in time. The fully nonlinear Boltzmann model is used for electrons. The neoclassical electric field is obtained by solving gyrokinetic Poisson equation with self-consistent poloidal variation. With our 4D (ψ, θ, ε, μ) version of the TEMPEST code we compute radial particle and heat flux, the Geodesic-Acoustic Mode (GAM), and the development of neoclassical electric field, which we compare with neoclassical theory with a Lorentz collision model. The present work provides a numerical scheme and a new capability for self-consistently studying important aspects of neoclassical transport and rotations in toroidal magnetic fusion devices
Computer simulation of the Blumlein pulse forming network
International Nuclear Information System (INIS)
Edwards, C.B.
1981-03-01
A computer simulation of the Blumlein pulse-forming network is described. The model is able to treat the case of time varying loads, non-zero conductor resistance, and switch closure effects as exhibited by real systems employing non-ohmic loads such as field-emission vacuum diodes in which the impedance is strongly time and voltage dependent. The application of the code to various experimental arrangements is discussed, with particular reference to the prediction of the behaviour of the output circuit of 'ELF', the electron beam generator in operation at the Rutherford Laboratory. The output from the code is compared directly with experimentally obtained voltage waveforms applied to the 'ELF' diode. (author)
Chierici, A.; Chirco, L.; Da Vià, R.; Manservisi, S.; Scardovelli, R.
2017-11-01
Nowadays the rapidly-increasing computational power allows scientists and engineers to perform numerical simulations of complex systems that can involve many scales and several different physical phenomena. In order to perform such simulations, two main strategies can be adopted: one may develop a new numerical code where all the physical phenomena of interest are modelled or one may couple existing validated codes. With the latter option, the creation of a huge and complex numerical code is avoided but efficient methods for data exchange are required since the performance of the simulation is highly influenced by its coupling techniques. In this work we propose a new algorithm that can be used for volume and/or boundary coupling purposes for both multiscale and multiphysics numerical simulations. The proposed algorithm is used for a multiscale simulation involving several CFD domains and monodimensional loops. We adopt the overlapping domain strategy, so the entire flow domain is simulated with the system code. We correct the system code solution by matching averaged inlet and outlet fields located at the boundaries of the CFD domains that overlap parts of the monodimensional loop. In particular we correct pressure losses and enthalpy values with source-sink terms that are imposed in the system code equations. The 1D-CFD coupling is a defective one since the CFD code requires point-wise values on the coupling interfaces and the system code provides only averaged quantities. In particular we impose, as inlet boundary conditions for the CFD domains, the mass flux and the mean enthalpy that are calculated by the system code. With this method the mass balance is preserved at every time step of the simulation. The coupling between consecutive CFD domains is not a defective one since with the proposed algorithm we can interpolate the field solutions on the boundary interfaces. We use the MED data structure as the base structure where all the field operations are
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......-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...
Development of an object-oriented simulation code for repository performance assessment
International Nuclear Information System (INIS)
Tsujimoto, Keiichi; Ahn, J.
1999-01-01
As understanding for mechanisms of radioactivity confinement by a deep geologic repository improves at the individual process level, it has become imperative to evaluate consequences of individual processes to the performance of the whole repository system. For this goal, the authors have developed a model for radionuclide transport in, and release from, the repository region by incorporating multiple-member decay chains and multiple waste canisters. A computer code has been developed with C++, an object-oriented language. By utilizing the feature that a geologic repository consists of thousands of objects of the same kind, such as the waste canister, the repository region is divided into multiple compartments and objects for simulation of radionuclide transport. Massive computational tasks are distributed over, and executed by, multiple networked workstations, with the help of parallel virtual machine (PVM) technology. Temporal change of the mass distribution of 28 radionuclides in the repository region for the time period of 100 million yr has been successfully obtained by the code
Computer code conversion using HISTORIAN
International Nuclear Information System (INIS)
Matsumoto, Kiyoshi; Kumakura, Toshimasa.
1990-09-01
When a computer program written for a computer A is converted for a computer B, in general, the A version source program is rewritten for B version. However, in this way of program conversion, the following inconvenient problems arise. 1) The original statements to be rewritten for B version are lost. 2) If the original statements of the A version rewritten for B version would remain as comment lines, the B version source program becomes quite large. 3) When update directives of the program are mailed from the organization which developed the program or when some modifications are needed for the program, it is difficult to point out the part to be updated or modified in the B version source program. To solve these problems, the conversion method using the general-purpose software management aid system, HISTORIAN, has been introduced. This conversion method makes a large computer code a easy-to-use program for use to update, modify or improve after the conversion. This report describes the planning and procedures of the conversion method and the MELPROG-PWR/MOD1 code conversion from the CRAY version to the JAERI FACOM version as an example. This report would provide useful information for those who develop or introduce large programs. (author)
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.
Further development of the computer code ATHLET-CD
International Nuclear Information System (INIS)
Weber, Sebastian; Austregesilo, Henrique; Bals, Christine; Band, Sebastian; Hollands, Thorsten; Koellein, Carsten; Lovasz, Liviusz; Pandazis, Peter; Schubert, Johann-Dietrich; Sonnenkalb, Martin
2016-10-01
In the framework of the reactor safety research program sponsored by the German Federal Ministry for Economic Affairs and Energy (BMWi), the computer code system ATHLET/ATHLET-CD has been further developed as an analysis tool for the simulation of accidents in nuclear power plants with pressurized and boiling water reactors as well as for the evaluation of accident management procedures. The main objective was to provide a mechanistic analysis tool for best estimate calculations of transients, accidents, and severe accidents with core degradation in light water reactors. With the continued development, the capability of the code system has been largely improved, allowing best estimate calculations of design and beyond design base accidents, and the simulation of advanced core degradation with enhanced model extent in a reasonable calculation time. ATHLET comprises inter alia a 6-equation model, models for the simulation of non-condensable gases and tracking of boron concentration, as well as additional component and process models for the complete system simulation. Among numerous model improvements, the code application has been extended to super critical pressures. The mechanistic description of the dynamic development of flow regimes on the basis of a transport equation for the interface area has been further developed. This ATHLET version is completely integrated in ATHLET-CD. ATHLET-CD further comprises dedicated models for the simulation of fuel and control assembly degradation for both pressurized and boiling water reactors, debris bed with melting in the core region, as well as fission product and aerosol release and transport in the cooling system, inclusive of decay of nuclide inventories and of chemical reactions in the gas phase. The continued development also concerned the modelling of absorber material release, of melting, melt relocation and freezing, and the interaction with the wall of the reactor pressure vessel. The following models were newly
A program code generator for multiphysics biological simulation using markup languages.
Amano, Akira; Kawabata, Masanari; Yamashita, Yoshiharu; Rusty Punzalan, Florencio; Shimayoshi, Takao; Kuwabara, Hiroaki; Kunieda, Yoshitoshi
2012-01-01
To cope with the complexity of the biological function simulation models, model representation with description language is becoming popular. However, simulation software itself becomes complex in these environment, thus, it is difficult to modify the simulation conditions, target computation resources or calculation methods. In the complex biological function simulation software, there are 1) model equations, 2) boundary conditions and 3) calculation schemes. Use of description model file is useful for first point and partly second point, however, third point is difficult to handle for various calculation schemes which is required for simulation models constructed from two or more elementary models. We introduce a simulation software generation system which use description language based description of coupling calculation scheme together with cell model description file. By using this software, we can easily generate biological simulation code with variety of coupling calculation schemes. To show the efficiency of our system, example of coupling calculation scheme with three elementary models are shown.
Holonomic surface codes for fault-tolerant quantum computation
Zhang, Jiang; Devitt, Simon J.; You, J. Q.; Nori, Franco
2018-02-01
Surface codes can protect quantum information stored in qubits from local errors as long as the per-operation error rate is below a certain threshold. Here we propose holonomic surface codes by harnessing the quantum holonomy of the system. In our scheme, the holonomic gates are built via auxiliary qubits rather than the auxiliary levels in multilevel systems used in conventional holonomic quantum computation. The key advantage of our approach is that the auxiliary qubits are in their ground state before and after each gate operation, so they are not involved in the operation cycles of surface codes. This provides an advantageous way to implement surface codes for fault-tolerant quantum computation.
User's manual for a measurement simulation code
International Nuclear Information System (INIS)
Kern, E.A.
1982-07-01
The MEASIM code has been developed primarily for modeling process measurements in materials processing facilities associated with the nuclear fuel cycle. In addition, the code computes materials balances and the summation of materials balances along with associated variances. The code has been used primarily in performance assessment of materials' accounting systems. This report provides the necessary information for a potential user to employ the code in these applications. A number of examples that demonstrate most of the capabilities of the code are provided
International Nuclear Information System (INIS)
Merle-Szeremeta, A.; Thomassin, A.
1999-01-01
The Institute of Protection and Nuclear Safety (I.P.S.N.) has developed a computer code, FOCON96 1.0 to calculate the dosimetric consequences of atmospheric radioactive releases from nuclear installations after several years of usual operation. This communication describes the principal characteristics of FOCON96 1.0 and its functionalities. The principal elements of a comparison between FOCON96 1.0 and PC-CREAM ( European computer code developed by the N.R.P.B. and answering the same criteria) are given here. (N.C.)
Directory of Open Access Journals (Sweden)
Pešić Milan P.
2012-01-01
Full Text Available A numerical simulation of the radiological consequences of the RB reactor reactivity excursion accident, which occurred on October 15, 1958, and an estimation of the total doses received by the operators were run by the MCNP5 computer code. The simulation was carried out under the same assumptions as those used in the 1960 IAEA-organized experimental simulation of the accident: total fission energy of 80 MJ released in the accident and the frozen positions of the operators. The time interval of exposure to high doses received by the operators has been estimated. Data on the RB1/1958 reactor core relevant to the accident are given. A short summary of the accident scenario has been updated. A 3-D model of the reactor room and the RB reactor tank, with all the details of the core, created. For dose determination, 3-D simplified, homogenised, sexless and faceless phantoms, placed inside the reactor room, have been developed. The code was run for a number of neutron histories which have given a dose rate uncertainty of less than 2%. For the determination of radiation spectra escaping the reactor core and radiation interaction in the tissue of the phantoms, the MCNP5 code was run (in the KCODE option and “mode n p e”, with a 55-group neutron spectra, 35-group gamma ray spectra and a 10-group electron spectra. The doses were determined by using the conversion of flux density (obtained by the F4 tally in the phantoms to doses using factors taken from ICRP-74 and from the deposited energy of neutrons and gamma rays (obtained by the F6 tally in the phantoms’ tissue. A rough estimation of the time moment when the odour of ozone was sensed by the operators is estimated for the first time and given in Appendix A.1. Calculated total absorbed and equivalent doses are compared to the previously reported ones and an attempt to understand and explain the reasons for the obtained differences has been made. A Root Cause Analysis of the accident was done and
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...
Computer codes for shaping the magnetic field of the JINR phasotron
International Nuclear Information System (INIS)
Zaplatin, N.L.; Morozov, N.A.
1983-01-01
The computer codes providing for the shaping the magnetic field of the JINR high current phasotron are presented. Using these codes the control for the magnetic field mapping was realized in on- or off-line regimes. Then these field parameters were calculated and ferromagnetic correcting elements and trim coils setting were chosen. Some computer codes were realised for the magnetic field horizontal component measurements. The data are presented on some codes possibilities. The codes were used on the EC-1010 and the CDC-6500 computers
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
Simulation of atmosphere stratification in the HDR test facility with the CONTAIN code
International Nuclear Information System (INIS)
Skerlavaj, A.; Mavko, B.; Kljenak, I.
2001-01-01
The test E11.2 'Hydrogen distribution in loop flow geometry', which was performed in the Heissdampf Reaktor containment test facility in Germany, was simulated with the CONTAIN computer code. The predicted pressure history and thermal stratification are in relatively good agreement with the measurements. The compositional stratification within the containment was qualitatively well predicted, although the degree of the stratification in the dome area was slightly underestimated. The analysis of simulation results enabled a better understanding of the physical phenomena during the test.(author)
COMPUTATION FORMAT computer codes X4TOC4 and PLOTC4. Implementing and Testing on a Personal Computer
International Nuclear Information System (INIS)
McLaughlin, P.K.
1987-05-01
This document describes the contents of the diskette containing the COMPUTATION FORMAT codes X4TOC4 and PLOTC4 by D.E. Cullen, and example data for use in implementing and testing these codes on a Personal Computer of the type IBM-PC/AT. Upon request the codes are available from the IAEA Nuclear Data Section, free of charge, on a single diskette. (author)
International Nuclear Information System (INIS)
Soares, Alexandre de Souza
2014-01-01
The present paper proposes a study of a loss of coolant accident of a PWR reactor through a Full Scope Simulator and computational code RELAP. To this end, it considered a loss of coolant accident with 160 cm 2 breaking area in cold leg of 20 circuit of the reactor cooling system of nuclear power plant Angra 2, with the reactor operating in stationary condition, to 100% power. It considered that occurred at the same time the loss of External Power Supply and the availability of emergency cooling system was not full. The results obtained are quite relevant and with the possibility of being used in the planning of future activities, given that the construction of Angra 3 is underway and resembles the Angra 2. (author)
CloudMC: a cloud computing application for Monte Carlo simulation
International Nuclear Information System (INIS)
Miras, H; Jiménez, R; Miras, C; Gomà, C
2013-01-01
This work presents CloudMC, a cloud computing application—developed in Windows Azure®, the platform of the Microsoft® cloud—for the parallelization of Monte Carlo simulations in a dynamic virtual cluster. CloudMC is a web application designed to be independent of the Monte Carlo code in which the simulations are based—the simulations just need to be of the form: input files → executable → output files. To study the performance of CloudMC in Windows Azure®, Monte Carlo simulations with penelope were performed on different instance (virtual machine) sizes, and for different number of instances. The instance size was found to have no effect on the simulation runtime. It was also found that the decrease in time with the number of instances followed Amdahl's law, with a slight deviation due to the increase in the fraction of non-parallelizable time with increasing number of instances. A simulation that would have required 30 h of CPU on a single instance was completed in 48.6 min when executed on 64 instances in parallel (speedup of 37 ×). Furthermore, the use of cloud computing for parallel computing offers some advantages over conventional clusters: high accessibility, scalability and pay per usage. Therefore, it is strongly believed that cloud computing will play an important role in making Monte Carlo dose calculation a reality in future clinical practice. (note)
CloudMC: a cloud computing application for Monte Carlo simulation.
Miras, H; Jiménez, R; Miras, C; Gomà, C
2013-04-21
This work presents CloudMC, a cloud computing application-developed in Windows Azure®, the platform of the Microsoft® cloud-for the parallelization of Monte Carlo simulations in a dynamic virtual cluster. CloudMC is a web application designed to be independent of the Monte Carlo code in which the simulations are based-the simulations just need to be of the form: input files → executable → output files. To study the performance of CloudMC in Windows Azure®, Monte Carlo simulations with penelope were performed on different instance (virtual machine) sizes, and for different number of instances. The instance size was found to have no effect on the simulation runtime. It was also found that the decrease in time with the number of instances followed Amdahl's law, with a slight deviation due to the increase in the fraction of non-parallelizable time with increasing number of instances. A simulation that would have required 30 h of CPU on a single instance was completed in 48.6 min when executed on 64 instances in parallel (speedup of 37 ×). Furthermore, the use of cloud computing for parallel computing offers some advantages over conventional clusters: high accessibility, scalability and pay per usage. Therefore, it is strongly believed that cloud computing will play an important role in making Monte Carlo dose calculation a reality in future clinical practice.
AX-GADGET: a new code for cosmological simulations of Fuzzy Dark Matter and Axion models
Nori, Matteo; Baldi, Marco
2018-05-01
We present a new module of the parallel N-Body code P-GADGET3 for cosmological simulations of light bosonic non-thermal dark matter, often referred as Fuzzy Dark Matter (FDM). The dynamics of the FDM features a highly non-linear Quantum Potential (QP) that suppresses the growth of structures at small scales. Most of the previous attempts of FDM simulations either evolved suppressed initial conditions, completely neglecting the dynamical effects of QP throughout cosmic evolution, or resorted to numerically challenging full-wave solvers. The code provides an interesting alternative, following the FDM evolution without impairing the overall performance. This is done by computing the QP acceleration through the Smoothed Particle Hydrodynamics (SPH) routines, with improved schemes to ensure precise and stable derivatives. As an extension of the P-GADGET3 code, it inherits all the additional physics modules implemented up to date, opening a wide range of possibilities to constrain FDM models and explore its degeneracies with other physical phenomena. Simulations are compared with analytical predictions and results of other codes, validating the QP as a crucial player in structure formation at small scales.
A compositional reservoir simulator on distributed memory parallel computers
International Nuclear Information System (INIS)
Rame, M.; Delshad, M.
1995-01-01
This paper presents the application of distributed memory parallel computes to field scale reservoir simulations using a parallel version of UTCHEM, The University of Texas Chemical Flooding Simulator. The model is a general purpose highly vectorized chemical compositional simulator that can simulate a wide range of displacement processes at both field and laboratory scales. The original simulator was modified to run on both distributed memory parallel machines (Intel iPSC/960 and Delta, Connection Machine 5, Kendall Square 1 and 2, and CRAY T3D) and a cluster of workstations. A domain decomposition approach has been taken towards parallelization of the code. A portion of the discrete reservoir model is assigned to each processor by a set-up routine that attempts a data layout as even as possible from the load-balance standpoint. Each of these subdomains is extended so that data can be shared between adjacent processors for stencil computation. The added routines that make parallel execution possible are written in a modular fashion that makes the porting to new parallel platforms straight forward. Results of the distributed memory computing performance of Parallel simulator are presented for field scale applications such as tracer flood and polymer flood. A comparison of the wall-clock times for same problems on a vector supercomputer is also presented
Computational simulation of acoustic fatigue for hot composite structures
Singhal, S. N.; Nagpal, V. K.; Murthy, P. L. N.; Chamis, C. C.
1991-01-01
This paper presents predictive methods/codes for computational simulation of acoustic fatigue resistance of hot composite structures subjected to acoustic excitation emanating from an adjacent vibrating component. Select codes developed over the past two decades at the NASA Lewis Research Center are used. The codes include computation of (1) acoustic noise generated from a vibrating component, (2) degradation in material properties of the composite laminate at use temperature, (3) dynamic response of acoustically excited hot multilayered composite structure, (4) degradation in the first-ply strength of the excited structure due to acoustic loading, and (5) acoustic fatigue resistance of the excited structure, including propulsion environment. Effects of the laminate lay-up and environment on the acoustic fatigue life are evaluated. The results show that, by keeping the angled plies on the outer surface of the laminate, a substantial increase in the acoustic fatigue life is obtained. The effect of environment (temperature and moisure) is to relieve the residual stresses leading to an increase in the acoustic fatigue life of the excited panel.
Accelerating Climate Simulations Through Hybrid Computing
Zhou, Shujia; Sinno, Scott; Cruz, Carlos; Purcell, Mark
2009-01-01
Unconventional multi-core processors (e.g., IBM Cell B/E and NYIDIDA GPU) have emerged as accelerators in climate simulation. However, climate models typically run on parallel computers with conventional processors (e.g., Intel and AMD) using MPI. Connecting accelerators to this architecture efficiently and easily becomes a critical issue. When using MPI for connection, we identified two challenges: (1) identical MPI implementation is required in both systems, and; (2) existing MPI code must be modified to accommodate the accelerators. In response, we have extended and deployed IBM Dynamic Application Virtualization (DAV) in a hybrid computing prototype system (one blade with two Intel quad-core processors, two IBM QS22 Cell blades, connected with Infiniband), allowing for seamlessly offloading compute-intensive functions to remote, heterogeneous accelerators in a scalable, load-balanced manner. Currently, a climate solar radiation model running with multiple MPI processes has been offloaded to multiple Cell blades with approx.10% network overhead.
Assessment of the computer code COBRA/CFTL
International Nuclear Information System (INIS)
Baxi, C.B.; Burhop, C.J.
1981-07-01
The COBRA/CFTL code has been developed by Oak Ridge National Laboratory (ORNL) for thermal-hydraulic analysis of simulated gas-cooled fast breeder reactor (GCFR) core assemblies to be tested in the core flow test loop (CFTL). The COBRA/CFTL code was obtained by modifying the General Atomic code COBRA*GCFR. This report discusses these modifications, compares the two code results for three cases which represent conditions from fully rough turbulent flow to laminar flow. Case 1 represented fully rough turbulent flow in the bundle. Cases 2 and 3 represented laminar and transition flow regimes. The required input for the COBRA/CFTL code, a sample problem input/output and the code listing are included in the Appendices
Energy Technology Data Exchange (ETDEWEB)
Grebennikov, A.N.; Zhitnik, A.K.; Zvenigorodskaya, O.A. [and others
1995-12-31
In conformity with the protocol of the Workshop under Contract {open_quotes}Assessment of RBMK reactor safety using modern Western Codes{close_quotes} VNIIEF performed a neutronics computation series to compare western and VNIIEF codes and assess whether VNIIEF codes are suitable for RBMK type reactor safety assessment computation. The work was carried out in close collaboration with M.I. Rozhdestvensky and L.M. Podlazov, NIKIET employees. The effort involved: (1) cell computations with the WIMS, EKRAN codes (improved modification of the LOMA code) and the S-90 code (VNIIEF Monte Carlo). Cell, polycell, burnup computation; (2) 3D computation of static states with the KORAT-3D and NEU codes and comparison with results of computation with the NESTLE code (USA). The computations were performed in the geometry and using the neutron constants presented by the American party; (3) 3D computation of neutron kinetics with the KORAT-3D and NEU codes. These computations were performed in two formulations, both being developed in collaboration with NIKIET. Formulation of the first problem maximally possibly agrees with one of NESTLE problems and imitates gas bubble travel through a core. The second problem is a model of the RBMK as a whole with imitation of control and protection system controls (CPS) movement in a core.
Simulation of ROCOM Experiment using CUPID Code
Energy Technology Data Exchange (ETDEWEB)
Cho, Yun Je; Lee, Jae Ryong; Yoon, Han Young [KAERI, Daejeon (Korea, Republic of)
2016-10-15
KAERI has developed CUPID, which is a three dimensional thermal hydraulics code for the transient analysis of two-phase flows in nuclear reactor components. To validate the capability of CUPID for simulation of multi-dimensional flow mixing behavior, ROCOM (ROssenforf COolant Mixing) test was simulated. ROCOM test has been conducted in the OECD PKL2 Project to investigate in more detail the thermal hydraulic behavior inside the RPV. Thus far, many researchers used the ROCOM data to validate the CFD code capability of thermal mixing behavior. In this study, a hybrid grid was generated using SALOME software and the ROCOM simulation was performed using CUPID. In addition, the effect of turbulence model was also investigated. Test ROCOM 2.1 and 1.2 cases were simulated using the CUPID code. It was shown that CUPID had capabilities to properly simulate the thermal mixing behavior in the case where the cold water is injected asymmetrically. As the result of calculations, it was found that the mixing efficiency in the downcomer and lower plenum was varied according to the turbulence model. In particular, the calculation results showed that the low Reynolds number turbulence model resulted in good agreement with the experimental data. The further works may involve the finer grid generation and the test of other turbulence models.
N-MODY: A Code for Collisionless N-body Simulations in Modified Newtonian Dynamics
Londrillo, Pasquale; Nipoti, Carlo
2011-02-01
N-MODY is a parallel particle-mesh code for collisionless N-body simulations in modified Newtonian dynamics (MOND). N-MODY is based on a numerical potential solver in spherical coordinates that solves the non-linear MOND field equation, and is ideally suited to simulate isolated stellar systems. N-MODY can be used also to compute the MOND potential of arbitrary static density distributions. A few applications of N-MODY indicate that some astrophysically relevant dynamical processes are profoundly different in MOND and in Newtonian gravity with dark matter.
International Nuclear Information System (INIS)
Watanabe, N.; Nishiguchi, R.; Shimomura, Y.
1991-01-01
Spatial distribution of point defects in displacement damage cascades at the early stage of their formation was simulated with the MARLOWE code for primary knock-on atoms which is relevant to D-T neutron irradiation. Calculations were carried out for Au, Ag, Cu, Ni and Al. Computer-simulated results were analyzed with complement of TEM observations of D-T neutron-irradiated metals at low temperature. The spatial configuration of displacement cascades, the size of small vacancy aggregates and the size of displacement damage cascade were examined. Results suggest that most of vacancy clusters which were formed in damage cascades may be as small as below 20 vacancies. The remarkable difference in defect yield of cascade damage in Ni and Cu is due to interstitial cluster formation and main contribution of cascade energy overlapping observed in cryotransfer TEM of D-T neutron-irradiated Au is due to ejected interstitials from cascade cores. (orig.)
Nonlinear simulations with and computational issues for NIMROD
International Nuclear Information System (INIS)
Sovinec, C.R.
1998-01-01
The NIMROD (Non-Ideal Magnetohydrodynamics with Rotation, Open Discussion) code development project was commissioned by the US Department of Energy in February, 1996 to provide the fusion research community with a computational tool for studying low-frequency behavior in experiments. Specific problems of interest include the neoclassical evolution of magnetic islands and the nonlinear behavior of tearing modes in the presence of rotation and nonideal walls in tokamaks; they also include topics relevant to innovative confinement concepts such as magnetic turbulence. Besides having physics models appropriate for these phenomena, an additional requirement is the ability to perform the computations in realistic geometries. The NIMROD Team is using contemporary management and computational methods to develop a computational tool for investigating low-frequency behavior in plasma fusion experiments. The authors intend to make the code freely available, and are taking steps to make it as easy to learn and use as possible. An example application for NIMROD is the nonlinear toroidal RFP simulation--the first in a series to investigate how toroidal geometry affects MHD activity in RFPs. Finally, the most important issue facing the project is execution time, and they are exploring better matrix solvers and a better parallel decomposition to address this
Nonlinear simulations with and computational issues for NIMROD
Energy Technology Data Exchange (ETDEWEB)
Sovinec, C.R. [Los Alamos National Lab., NM (United States)
1998-12-31
The NIMROD (Non-Ideal Magnetohydrodynamics with Rotation, Open Discussion) code development project was commissioned by the US Department of Energy in February, 1996 to provide the fusion research community with a computational tool for studying low-frequency behavior in experiments. Specific problems of interest include the neoclassical evolution of magnetic islands and the nonlinear behavior of tearing modes in the presence of rotation and nonideal walls in tokamaks; they also include topics relevant to innovative confinement concepts such as magnetic turbulence. Besides having physics models appropriate for these phenomena, an additional requirement is the ability to perform the computations in realistic geometries. The NIMROD Team is using contemporary management and computational methods to develop a computational tool for investigating low-frequency behavior in plasma fusion experiments. The authors intend to make the code freely available, and are taking steps to make it as easy to learn and use as possible. An example application for NIMROD is the nonlinear toroidal RFP simulation--the first in a series to investigate how toroidal geometry affects MHD activity in RFPs. Finally, the most important issue facing the project is execution time, and they are exploring better matrix solvers and a better parallel decomposition to address this.
International Nuclear Information System (INIS)
Lee, Y. J.; Bae, S. W.; Chung, B. D.
2003-02-01
This report records the results of the code validation for the one-dimensional module of the MARS 2.1 thermal hydraulics analysis code by means of result-comparison with the RELAP5/MOD3.3 computer code. For the validation calculations, simulations of the RELAP5 code development assessment problem, which consists of 22 simulation problems in 3 categories, have been selected. The results of the 3 categories of simulations demonstrate that the one-dimensional module of the MARS 2.1 code and the RELAP5/MOD3.3 code are essentially the same code. This is expected as the two codes have basically the same set of field equations, constitutive equations and main thermal hydraulic models. The results suggests that the high level of code validity of the RELAP5/MOD3.3 can be directly applied to the MARS one-dimensional module
SPINET: A Parallel Computing Approach to Spine Simulations
Directory of Open Access Journals (Sweden)
Peter G. Kropf
1996-01-01
Full Text Available Research in scientitic programming enables us to realize more and more complex applications, and on the other hand, application-driven demands on computing methods and power are continuously growing. Therefore, interdisciplinary approaches become more widely used. The interdisciplinary SPINET project presented in this article applies modern scientific computing tools to biomechanical simulations: parallel computing and symbolic and modern functional programming. The target application is the human spine. Simulations of the spine help us to investigate and better understand the mechanisms of back pain and spinal injury. Two approaches have been used: the first uses the finite element method for high-performance simulations of static biomechanical models, and the second generates a simulation developmenttool for experimenting with different dynamic models. A finite element program for static analysis has been parallelized for the MUSIC machine. To solve the sparse system of linear equations, a conjugate gradient solver (iterative method and a frontal solver (direct method have been implemented. The preprocessor required for the frontal solver is written in the modern functional programming language SML, the solver itself in C, thus exploiting the characteristic advantages of both functional and imperative programming. The speedup analysis of both solvers show very satisfactory results for this irregular problem. A mixed symbolic-numeric environment for rigid body system simulations is presented. It automatically generates C code from a problem specification expressed by the Lagrange formalism using Maple.
User manual of FRAPCON-I computer code
International Nuclear Information System (INIS)
Chia, C.T.
1985-11-01
The manual for using the FRAPCON-I code implanted by Reactor Department of Brazilian-CNEN to convert IBM FORTRAN in FORTRAN 77 of Honeywell Bull computer is presented. The FRAPCON-I code describes the behaviour of fuel rods of PWR type reactors at stationary state during long periods of burnup. (M.C.K.)
Energy Technology Data Exchange (ETDEWEB)
Hartmann, Christoph Oliver
2016-06-13
Coupled Thermal-hydraulic/Neutron-kinetic (TH/NK) simulations of Boiling Water Reactor transients require well validated and accurate simulation tools. The generation of cross-section (XS) libraries, depending on the individual thermal-hydraulic state parameters, is of paramount importance for coupled simulations. Problem-dependent XS-sets for 3D core simulations are being generated mainly by well validated, fast running commercial and user-friendly lattice codes such as CASMO and HELIOS. In this dissertation a computational route, based on the lattice code SCALE6/TRITON, the cross-section interface GenPMAXS, the best-estimate thermal-hydraulic system code TRACE and the core simulator PARCS, for best-estimate simulations of Boiling Water (BWR) transients has been developed and validated. The computational route has been supplemented by a subsequent uncertainty and sensitivity study based on Monte Carlo sampling and propagation of the uncertainties of input parameters to the output (SUSA code). The analysis of a single BWR fuel assembly depletion problem with PARCS using SCALE/TRITON cross-sections has been shown a good agreement with the results obtained with CASMO cross-section sets. However, to compensate the deficiencies of the interface program GenPMAXS, PYTHON scripts had to be developed to incorporate missing data, as the yields of Iodine, Xenon and Promethium, into the cross-section-data sets (PMAXS-format) generated by GenPMAXS from the SCALE/TRITON output. The results of the depletion analysis of a full BWR core with PARCS have indicated the importance of considering history effects, adequate modeling of the reflector region and the control rods, as the PARCS simulations for depleted fuel and all control rods inserted (ARI) differs significantly at the fuel assembly top and bottom. Systematic investigations with the coupled codes TRACE/PARCS have been performed to analyse the core behaviour at different thermal conditions using nuclear data (XS
Energy Technology Data Exchange (ETDEWEB)
Kurosu, Keita [Department of Radiation Oncology, Indiana University School of Medicine, Indianapolis, IN 46202 (United States); Department of Radiation Oncology, Osaka University Graduate School of Medicine, Suita, Osaka 565-0871 (Japan); Department of Radiology, Osaka University Hospital, Suita, Osaka 565-0871 (Japan); Das, Indra J. [Department of Radiation Oncology, Indiana University School of Medicine, Indianapolis, IN 46202 (United States); Moskvin, Vadim P. [Department of Radiation Oncology, Indiana University School of Medicine, Indianapolis, IN 46202 (United States); Department of Radiation Oncology, St. Jude Children’s Research Hospital, Memphis, TN 38105 (United States)
2016-01-15
Spot scanning, owing to its superior dose-shaping capability, provides unsurpassed dose conformity, in particular for complex targets. However, the robustness of the delivered dose distribution and prescription has to be verified. Monte Carlo (MC) simulation has the potential to generate significant advantages for high-precise particle therapy, especially for medium containing inhomogeneities. However, the inherent choice of computational parameters in MC simulation codes of GATE, PHITS and FLUKA that is observed for uniform scanning proton beam needs to be evaluated. This means that the relationship between the effect of input parameters and the calculation results should be carefully scrutinized. The objective of this study was, therefore, to determine the optimal parameters for the spot scanning proton beam for both GATE and PHITS codes by using data from FLUKA simulation as a reference. The proton beam scanning system of the Indiana University Health Proton Therapy Center was modeled in FLUKA, and the geometry was subsequently and identically transferred to GATE and PHITS. Although the beam transport is managed by spot scanning system, the spot location is always set at the center of a water phantom of 600 × 600 × 300 mm{sup 3}, which is placed after the treatment nozzle. The percentage depth dose (PDD) is computed along the central axis using 0.5 × 0.5 × 0.5 mm{sup 3} voxels in the water phantom. The PDDs and the proton ranges obtained with several computational parameters are then compared to those of FLUKA, and optimal parameters are determined from the accuracy of the proton range, suppressed dose deviation, and computational time minimization. Our results indicate that the optimized parameters are different from those for uniform scanning, suggesting that the gold standard for setting computational parameters for any proton therapy application cannot be determined consistently since the impact of setting parameters depends on the proton irradiation
International Nuclear Information System (INIS)
Kurosu, Keita; Das, Indra J.; Moskvin, Vadim P.
2016-01-01
Spot scanning, owing to its superior dose-shaping capability, provides unsurpassed dose conformity, in particular for complex targets. However, the robustness of the delivered dose distribution and prescription has to be verified. Monte Carlo (MC) simulation has the potential to generate significant advantages for high-precise particle therapy, especially for medium containing inhomogeneities. However, the inherent choice of computational parameters in MC simulation codes of GATE, PHITS and FLUKA that is observed for uniform scanning proton beam needs to be evaluated. This means that the relationship between the effect of input parameters and the calculation results should be carefully scrutinized. The objective of this study was, therefore, to determine the optimal parameters for the spot scanning proton beam for both GATE and PHITS codes by using data from FLUKA simulation as a reference. The proton beam scanning system of the Indiana University Health Proton Therapy Center was modeled in FLUKA, and the geometry was subsequently and identically transferred to GATE and PHITS. Although the beam transport is managed by spot scanning system, the spot location is always set at the center of a water phantom of 600 × 600 × 300 mm 3 , which is placed after the treatment nozzle. The percentage depth dose (PDD) is computed along the central axis using 0.5 × 0.5 × 0.5 mm 3 voxels in the water phantom. The PDDs and the proton ranges obtained with several computational parameters are then compared to those of FLUKA, and optimal parameters are determined from the accuracy of the proton range, suppressed dose deviation, and computational time minimization. Our results indicate that the optimized parameters are different from those for uniform scanning, suggesting that the gold standard for setting computational parameters for any proton therapy application cannot be determined consistently since the impact of setting parameters depends on the proton irradiation technique
Some neutronics and thermal-hydraulics codes for reactor analysis using personal computers
International Nuclear Information System (INIS)
Woodruff, W.L.
1990-01-01
Some neutronics and thermal-hydraulics codes formerly available only for main frame computers may now be run on personal computers. Brief descriptions of the codes are provided. Running times for some of the codes are compared for an assortment of personal and main frame computers. With some limitations in detail, personal computer versions of the codes can be used to solve many problems of interest in reactor analyses at very modest costs. 11 refs., 4 tabs
Scientific and computational challenges of the fusion simulation project (FSP)
International Nuclear Information System (INIS)
Tang, W M
2008-01-01
This paper highlights the scientific and computational challenges facing the Fusion Simulation Project (FSP). The primary objective is to develop advanced software designed to use leadership-class computers for carrying out multiscale physics simulations to provide information vital to delivering a realistic integrated fusion simulation model with unprecedented physics fidelity. This multiphysics capability will be unprecedented in that in the current FES applications domain, the largest-scale codes are used to carry out first-principles simulations of mostly individual phenomena in realistic 3D geometry while the integrated models are much smaller-scale, lower-dimensionality codes with significant empirical elements used for modeling and designing experiments. The FSP is expected to be the most up-to-date embodiment of the theoretical and experimental understanding of magnetically confined thermonuclear plasmas and to provide a living framework for the simulation of such plasmas as the associated physics understanding continues to advance over the next several decades. Substantive progress on answering the outstanding scientific questions in the field will drive the FSP toward its ultimate goal of developing a reliable ability to predict the behavior of plasma discharges in toroidal magnetic fusion devices on all relevant time and space scales. From a computational perspective, the fusion energy science application goal to produce high-fidelity, whole-device modeling capabilities will demand computing resources in the petascale range and beyond, together with the associated multicore algorithmic formulation needed to address burning plasma issues relevant to ITER - a multibillion dollar collaborative device involving seven international partners representing over half the world's population. Even more powerful exascale platforms will be needed to meet the future challenges of designing a demonstration fusion reactor (DEMO). Analogous to other major applied physics
Steady-state and transient simulations of gas cooled reactor with the computer code CATHARE
International Nuclear Information System (INIS)
Tauveron, N.; Saez, M.; Marchand, M.; Chataing, T.; Geffraye, G.; Cherel, J. M.
2003-01-01
This work concerns the design and safety analysis of Gas Cooled Reactors. The CATHARE code is used to test the design and safety of two different concepts, a High Temperature Gas Reactor concept (HTGR) and a Gas Fast Reactor concept (GFR). Relative to the HTGR concept, three transient simulations are performed and described in this paper: loss of electrical load without turbomachine trip, 10 inch cold duct break, 10 inch cold duct break combined with a tube rupture of a cooling exchanger. A second step consists in modelling a GFR concept. A nominal steady state situation at a power of 600 MW is obtained and first transient simulations are carried out to study decay heat removal situations after primary loop depressurisation
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.
Scientific computer simulation review
International Nuclear Information System (INIS)
Kaizer, Joshua S.; Heller, A. Kevin; Oberkampf, William L.
2015-01-01
Before the results of a scientific computer simulation are used for any purpose, it should be determined if those results can be trusted. Answering that question of trust is the domain of scientific computer simulation review. There is limited literature that focuses on simulation review, and most is specific to the review of a particular type of simulation. This work is intended to provide a foundation for a common understanding of simulation review. This is accomplished through three contributions. First, scientific computer simulation review is formally defined. This definition identifies the scope of simulation review and provides the boundaries of the review process. Second, maturity assessment theory is developed. This development clarifies the concepts of maturity criteria, maturity assessment sets, and maturity assessment frameworks, which are essential for performing simulation review. Finally, simulation review is described as the application of a maturity assessment framework. This is illustrated through evaluating a simulation review performed by the U.S. Nuclear Regulatory Commission. In making these contributions, this work provides a means for a more objective assessment of a simulation’s trustworthiness and takes the next step in establishing scientific computer simulation review as its own field. - Highlights: • We define scientific computer simulation review. • We develop maturity assessment theory. • We formally define a maturity assessment framework. • We describe simulation review as the application of a maturity framework. • We provide an example of a simulation review using a maturity framework
MARLOWE: a computer simulation program in the BCA and it's application
International Nuclear Information System (INIS)
Zhou Peng; Pan Zhengying; Huo Yukun
1988-01-01
MARLOWE is a computer program for simulation of atomic-displacement cascades in a variety of solids based upon binary-collision approximation. This paper briefs the program including it's theoretical background, approximation adopted, features and applicable scope. In addition, we outline a number of physical problems we studied with this code
Interface design of VSOP'94 computer code for safety analysis
Energy Technology Data Exchange (ETDEWEB)
Natsir, Khairina, E-mail: yenny@batan.go.id; Andiwijayakusuma, D.; Wahanani, Nursinta Adi [Center for Development of Nuclear Informatics - National Nuclear Energy Agency, PUSPIPTEK, Serpong, Tangerang, Banten (Indonesia); Yazid, Putranto Ilham [Center for Nuclear Technology, Material and Radiometry- National Nuclear Energy Agency, Jl. Tamansari No.71, Bandung 40132 (Indonesia)
2014-09-30
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.
Simulation of containment phenomena during the Phebus FPT1 test with the CONTAIN code
International Nuclear Information System (INIS)
Kljenak, I.; Mavko, B.
2002-01-01
Thermal-hydraulic and aerosol phenomena which occurred in the containment vessel of the Phebus integral experimental facility during the first 30000 s of the Phebus FPT1 test were simulated with the CONTAIN thermal-hydraulic computer code. A single-cell input model of the vessel was developed, and boundary and initial conditions that were determined during the experiment were applied. The comparison of experimental and calculated results shows that, although the atmosphere temperature was well simulated, the calculated condensation rate was apparently too high, resulting in a lower pressure of the containment atmosphere. The aerosol deposition process was well simulated.(author)
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
Two-phase computer codes for zero-gravity applications
International Nuclear Information System (INIS)
Krotiuk, W.J.
1986-10-01
This paper discusses the problems existing in the development of computer codes which can analyze the thermal-hydraulic behavior of two-phase fluids especially in low gravity nuclear reactors. The important phenomenon affecting fluid flow and heat transfer in reduced gravity is discussed. The applicability of using existing computer codes for space applications is assessed. Recommendations regarding the use of existing earth based fluid flow and heat transfer correlations are made and deficiencies in these correlations are identified
MED101: a laser-plasma simulation code. User guide
International Nuclear Information System (INIS)
Rodgers, P.A.; Rose, S.J.; Rogoyski, A.M.
1989-12-01
Complete details for running the 1-D laser-plasma simulation code MED101 are given including: an explanation of the input parameters, instructions for running on the Rutherford Appleton Laboratory IBM, Atlas Centre Cray X-MP and DEC VAX, and information on three new graphics packages. The code, based on the existing MEDUSA code, is capable of simulating a wide range of laser-produced plasma experiments including the calculation of X-ray laser gain. (author)
Parallel Computing Characteristics of CUPID code under MPI and Hybrid environment
Energy Technology Data Exchange (ETDEWEB)
Lee, Jae Ryong; Yoon, Han Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Jeon, Byoung Jin; Choi, Hyoung Gwon [Seoul National Univ. of Science and Technology, Seoul (Korea, Republic of)
2014-05-15
In this paper, a characteristic of parallel algorithm is presented for solving an elliptic type equation of CUPID via domain decomposition method using the MPI and the parallel performance is estimated in terms of a scalability which shows the speedup ratio. In addition, the time-consuming pattern of major subroutines is studied. Two different grid systems are taken into account: 40,000 meshes for coarse system and 320,000 meshes for fine system. Since the matrix of the CUPID code differs according to whether the flow is single-phase or two-phase, the effect of matrix shape is evaluated. Finally, the effect of the preconditioner for matrix solver is also investigated. Finally, the hybrid (OpenMP+MPI) parallel algorithm is introduced and discussed in detail for solving pressure solver. Component-scale thermal-hydraulics code, CUPID has been developed for two-phase flow analysis, which adopts a three-dimensional, transient, three-field model, and parallelized to fulfill a recent demand for long-transient and highly resolved multi-phase flow behavior. In this study, the parallel performance of the CUPID code was investigated in terms of scalability. The CUPID code was parallelized with domain decomposition method. The MPI library was adopted to communicate the information at the neighboring domain. For managing the sparse matrix effectively, the CSR storage format is used. To take into account the characteristics of the pressure matrix which turns to be asymmetric for two-phase flow, both single-phase and two-phase calculations were run. In addition, the effect of the matrix size and preconditioning was also investigated. The fine mesh calculation shows better scalability than the coarse mesh because the number of coarse mesh does not need to decompose the computational domain excessively. The fine mesh can be present good scalability when dividing geometry with considering the ratio between computation and communication time. For a given mesh, single-phase flow
International Nuclear Information System (INIS)
Strickert, R.; Friedman, A.M.; Fried, S.
1979-04-01
A computer program (ARDISC, the Argonne Dispersion Code) is described which simulates the migration of nuclides in porous media and includes first order kinetic effects on the retention constants. The code allows for different absorption and desorption rates and solves the coupled migration equations by arithmetic reiterations. Input data needed are the absorption and desorption rates, equilibrium surface absorption coefficients, flow rates and volumes, and media porosities
Design of convolutional tornado code
Zhou, Hui; Yang, Yao; Gao, Hongmin; Tan, Lu
2017-09-01
As a linear block code, the traditional tornado (tTN) code is inefficient in burst-erasure environment and its multi-level structure may lead to high encoding/decoding complexity. This paper presents a convolutional tornado (cTN) code which is able to improve the burst-erasure protection capability by applying the convolution property to the tTN code, and reduce computational complexity by abrogating the multi-level structure. The simulation results show that cTN code can provide a better packet loss protection performance with lower computation complexity than tTN code.
Neoclassical simulation of tokamak plasmas using the continuum gyrokinetic code TEMPEST.
Xu, X Q
2008-07-01
We present gyrokinetic neoclassical simulations of tokamak plasmas with a self-consistent electric field using a fully nonlinear (full- f ) continuum code TEMPEST in a circular geometry. A set of gyrokinetic equations are discretized on a five-dimensional computational grid in phase space. The present implementation is a method of lines approach where the phase-space derivatives are discretized with finite differences, and implicit backward differencing formulas are used to advance the system in time. The fully nonlinear Boltzmann model is used for electrons. The neoclassical electric field is obtained by solving the gyrokinetic Poisson equation with self-consistent poloidal variation. With a four-dimensional (psi,theta,micro) version of the TEMPEST code, we compute the radial particle and heat fluxes, the geodesic-acoustic mode, and the development of the neoclassical electric field, which we compare with neoclassical theory using a Lorentz collision model. The present work provides a numerical scheme for self-consistently studying important dynamical aspects of neoclassical transport and electric field in toroidal magnetic fusion devices.
Software quality and process improvement in scientific simulation codes
Energy Technology Data Exchange (ETDEWEB)
Ambrosiano, J.; Webster, R. [Los Alamos National Lab., NM (United States)
1997-11-01
This report contains viewgraphs on the quest to develope better simulation code quality through process modeling and improvement. This study is based on the experience of the authors and interviews with ten subjects chosen from simulation code development teams at LANL. This study is descriptive rather than scientific.
FRESCO: fusion reactor simulation code for tokamaks
International Nuclear Information System (INIS)
Mantsinen, M.J.
1995-03-01
The study of the dynamics of tokamak fusion reactors, a zero-dimensional particle and power balance code FRESCO (Fusion Reactor Simulation Code) has been developed at the Department of Technical Physics of Helsinki University of Technology. The FRESCO code is based on zero-dimensional particle and power balance equations averaged over prescribed plasma profiles. In the report the data structure of the FRESCO code is described, including the description of the COMMON statements, program input, and program output. The general structure of the code is described, including the description of subprograms and functions. The physical model used and examples of the code performance are also included in the report. (121 tabs.) (author)
The COSIMA-experiments, a data base for validation of two-phase flow computer codes
International Nuclear Information System (INIS)
Class, G.; Meyder, R.; Stratmanns, E.
1985-12-01
The report presents an overview on the large data base generated with COSIMA. The data base is to be used to validate and develop computer codes for two-phase flow. In terms of fuel rod behavior it was found that during blowdown under realistic conditions only small strains are reached. For clad rupture extremely high rod internal pressure is necessary. Additionally important results were found in the behavior of a fuel rod simulator and on the effect of thermocouples attached on the cladding outer surface. Post-test calculations, performed with the codes RELAP and DRUFAN show a good agreement with the experiments. This however can be improved if the phase separation models in the codes would be updated. (orig./HP) [de
System for simulating fluctuation diagnostics for application to turbulence computations
International Nuclear Information System (INIS)
Bravenec, R.V.; Nevins, W.M.
2006-01-01
Present-day nonlinear microstability codes are able to compute the saturated fluctuations of a turbulent fluid versus space and time, whether the fluid be liquid, gas, or plasma. They are therefore able to determine turbulence-induced fluid (or particle) and energy fluxes. These codes, however, must be tested against experimental data not only with respect to transport but also characteristics of the fluctuations. The latter is challenging because of limitations in the diagnostics (e.g., finite spatial resolution) and the fact that the diagnostics typically do not measure exactly the quantities that the codes compute. In this work, we present a system based on IDL registered analysis and visualization software in which user-supplied 'diagnostic filters' are applied to the code outputs to generate simulated diagnostic signals. The same analysis techniques as applied to the measurements, e.g., digital time-series analysis, may then be applied to the synthesized signals. Their statistical properties, such as rms fluctuation level, mean wave numbers, phase and group velocities, correlation lengths and times, and in some cases full S(k,ω) spectra, can then be compared directly to those of the measurements
A compendium of computer codes in fault tree analysis
International Nuclear Information System (INIS)
Lydell, B.
1981-03-01
In the past ten years principles and methods for a unified system reliability and safety analysis have been developed. Fault tree techniques serve as a central feature of unified system analysis, and there exists a specific discipline within system reliability concerned with the theoretical aspects of fault tree evaluation. Ever since the fault tree concept was established, computer codes have been developed for qualitative and quantitative analyses. In particular the presentation of the kinetic tree theory and the PREP-KITT code package has influenced the present use of fault trees and the development of new computer codes. This report is a compilation of some of the better known fault tree codes in use in system reliability. Numerous codes are available and new codes are continuously being developed. The report is designed to address the specific characteristics of each code listed. A review of the theoretical aspects of fault tree evaluation is presented in an introductory chapter, the purpose of which is to give a framework for the validity of the different codes. (Auth.)
Computer code ANISN multiplying media and shielding calculation II. Code description (input/output)
International Nuclear Information System (INIS)
Maiorino, J.R.
1990-01-01
The user manual of the ANISN computer code describing input and output subroutines is presented. ANISN code was developed to solve one-dimensional transport equation for neutron or gamma rays in slab, sphere or cylinder geometry with general anisotropic scattering. The solution technique is the discrete ordinate method. (M.C.K.)
V.S.O.P.-computer code system for reactor physics and fuel cycle simulation
International Nuclear Information System (INIS)
Teuchert, E.; Hansen, U.; Haas, K.A.
1980-03-01
V.S.O.P. (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based on neutron flux synthesis with depletion and shutdown features, incore and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to Pebble Bed HTRs). Various techniques have been employed to accelerate the iterative processes and to optimize the internal data transfer. A limitation of the storage requirement to 360 K-bites is achieved by an overlay structure. The code system has been used extensively for comparison studies of reactors, their fuel cycles, and related detailed features. Beside its use in research and development work for the high temperature reactor the system has been applied successfully to LWR and Heavy Water Reactors. (orig.) [de
Quantifying Uncertainty from Computational Factors in Simulations of a Model Ballistic System
2017-08-01
related to the numerical structuring of a problem, such as cell size, domain extent, and system orientation. Depth of penetration of a threat into a... system in the simulation codes is tied to the domain structure , with coordinate axes aligned with cell edges. However, the position of the coordinate...physical systems are generally described by sets of equations involving continuous variables, such as time and position. Computational simulations
Development, verification and validation of the fuel channel behaviour computer code FACTAR
Energy Technology Data Exchange (ETDEWEB)
Westbye, C J; Brito, A C; MacKinnon, J C; Sills, H E; Langman, V J [Ontario Hydro, Toronto, ON (Canada)
1996-12-31
FACTAR (Fuel And Channel Temperature And Response) is a computer code developed to simulate the transient thermal and mechanical behaviour of 37-element or 28-element fuel bundles within a single CANDU fuel channel for moderate loss of coolant accident conditions including transition and large break LOCA`s (loss of coolant accidents) with emergency coolant injection assumed available. FACTAR`s predictions of fuel temperature and sheath failure times are used to subsequent assessment of fission product releases and fuel string expansion. This paper discusses the origin and development history of FACTAR, presents the mathematical models and solution technique, the detailed quality assurance procedures that are followed during development, and reports the future development of the code. (author). 27 refs., 3 figs.
Evaluation of the FRAPCON-3 Computer Code
Energy Technology Data Exchange (ETDEWEB)
Jernkvist, Lars Olof; Massih, Ali [Quantum Technologies AB, Uppsala (Sweden)
2002-03-01
The FRAPCON-3 computer code has been evaluated with respect to its applicability, modeling capability, user friendliness, source code structure and supporting experimental database. The code is intended for thermo-mechanical analyses of light water reactor nuclear fuel rods under steady-state operational conditions and moderate power excursions. It is applicable to both boiling- and pressurized water reactor fuel rods with UO{sub 2} fuel, ranging up to about 65 MWd/kg U in rod average burnup. The models and numerical methods in FRAPCON-3 are relatively simple, which makes the code transparent and also fairly easy to modify and extend for the user. The fundamental equations for heat transfer, structural analysis and fuel fission gas release are solved in one-dimensional (radial) and stationary (time-independent) form, and interaction between axial segments of the rod is confined to calculations of coolant axial flow and rod internal gas pressure. The code is fairly easy to use; fuel rod design data and time histories of fuel rod power and coolant inlet conditions are input via a single text file, and the corresponding calculated variation with time of important fuel rod parameters are printed to a single output file in textual form. The results can also be presented in graphical form through an interface to the general graphics program xmgr. FRAPCON-3 also provides the possibility to export calculated results to the transient fuel rod analysis code FRAPTRAN, where the data can be used as burnup-dependent initial conditions to a postulated transient. Most of the source code to FRAPCON-3 is written in Fortran-IV, which is an archaic, non-standard dialect of the Fortran programming language. Since Fortran-IV is not accepted by all compilers for the latest standard of the language, Fortran-95, there is a risk that the source code must be partly rewritten in the future. Documentation of the code comprises (i) a general code description, which briefly presents models
Evaluation of the FRAPCON-3 Computer Code
International Nuclear Information System (INIS)
Jernkvist, Lars Olof; Massih, Ali
2002-03-01
The FRAPCON-3 computer code has been evaluated with respect to its applicability, modeling capability, user friendliness, source code structure and supporting experimental database. The code is intended for thermo-mechanical analyses of light water reactor nuclear fuel rods under steady-state operational conditions and moderate power excursions. It is applicable to both boiling- and pressurized water reactor fuel rods with UO 2 fuel, ranging up to about 65 MWd/kg U in rod average burnup. The models and numerical methods in FRAPCON-3 are relatively simple, which makes the code transparent and also fairly easy to modify and extend for the user. The fundamental equations for heat transfer, structural analysis and fuel fission gas release are solved in one-dimensional (radial) and stationary (time-independent) form, and interaction between axial segments of the rod is confined to calculations of coolant axial flow and rod internal gas pressure. The code is fairly easy to use; fuel rod design data and time histories of fuel rod power and coolant inlet conditions are input via a single text file, and the corresponding calculated variation with time of important fuel rod parameters are printed to a single output file in textual form. The results can also be presented in graphical form through an interface to the general graphics program xmgr. FRAPCON-3 also provides the possibility to export calculated results to the transient fuel rod analysis code FRAPTRAN, where the data can be used as burnup-dependent initial conditions to a postulated transient. Most of the source code to FRAPCON-3 is written in Fortran-IV, which is an archaic, non-standard dialect of the Fortran programming language. Since Fortran-IV is not accepted by all compilers for the latest standard of the language, Fortran-95, there is a risk that the source code must be partly rewritten in the future. Documentation of the code comprises (i) a general code description, which briefly presents models
Code for the core simulation in pressurized water reactors
International Nuclear Information System (INIS)
Serrano, M.A.B.
1978-08-01
A computer code was developed for the simulation of the steady-state and transient behaviour of the average channel of a Pressurizer Water Reactor core. Point kinetics equations were used with the reactivity calculated for average temperatures in the channel with the fuel and moderator temperature feedbacks. The radial heat conduction equation in the fuel was solved numericaly. For calculating the thermodynamic properties of the coolant, the fundamental equations of conservation (mass, energy and momentum) were solved. The gap and clad were treated as a resistence added to the film coeficient. The fuel system equations were decoupled from the coolant equations. The program permitted the changes in the heat transfer correlations and the flow patterns along the coolant channel. Various test were performed to determine the steady-state and transient response employing the PWR core simulator developed, obtaining results with adequate precision. (Author) [pt
Independent peer review of nuclear safety computer codes
International Nuclear Information System (INIS)
Boyack, B.E.; Jenks, R.P.
1993-01-01
A structured, independent computer code peer-review process has been developed to assist the US Nuclear Regulatory Commission (NRC) and the US Department of Energy in their nuclear safety missions. This paper describes a structured process of independent code peer review, benefits associated with a code-independent peer review, as well as the authors' recent peer-review experience. The NRC adheres to the principle that safety of plant design, construction, and operation are the responsibility of the licensee. Nevertheless, NRC staff must have the ability to independently assess plant designs and safety analyses submitted by license applicants. According to Ref. 1, open-quotes this requires that a sound understanding be obtained of the important physical phenomena that may occur during transients in operating power plants.close quotes The NRC concluded that computer codes are the principal products to open-quotes understand and predict plant response to deviations from normal operating conditionsclose quotes and has developed several codes for that purpose. However, codes cannot be used blindly; they must be assessed and found adequate for the purposes they are intended. A key part of the qualification process can be accomplished through code peer reviews; this approach has been adopted by the NRC
International Nuclear Information System (INIS)
Rahatgaonkar, P. S.; Datta, D.; Malhotra, P. K.; Ghadge, S. G.
2012-01-01
Prediction of groundwater movement and contaminant transport in soil is an important problem in many branches of science and engineering. This includes groundwater hydrology, environmental engineering, soil science, agricultural engineering and also nuclear engineering. Specifically, in nuclear engineering it is applicable in the design of spent fuel storage pools and waste management sites in the nuclear power plants. Ground water modeling involves the simulation of flow and contaminant transport by groundwater flow. In the context of contaminated soil and groundwater system, numerical simulations are typically used to demonstrate compliance with regulatory standard. A one-dimensional Computational Fluid Dynamics code GFLOW had been developed based on the Finite Difference Method for simulating groundwater flow and contaminant transport through saturated and unsaturated soil. The code is validated with the analytical model and the benchmarking cases available in the literature. (authors)
International Nuclear Information System (INIS)
Maekawa, Akira; Serizawa, Hisashi; Nakacho, Keiji; Murakawa, Hidekazu
2011-01-01
There are many weld zones in the apparatus and piping installed in nuclear power plants and residual stress generated in the zone by weld process is the most important influence factor for maintaining structural integrity. Though the weld residual stress is frequently evaluated using numerical simulation, fast simulation techniques have been demanded because of the enormous calculation times used. Recently, the fast weld residual stress evaluation based on three-dimensional accurate analysis became available through development of the Iterative Substructure Method (ISM). In this study, the computational performance of the welding simulation code using the ISM was improved to get faster computations and more accurate welding simulation. By adding functions such as parallel processing, the computation speed was much faster than that of the conventional finite element method code. Furthermore, the accuracy of the improved code was validated by measurements. The influence of two different weld heat source models on the simulation results was also investigated and it was found that the moving heat source was effective to achieve accurate weld simulation for multi-pass welds. (author)
Energy Technology Data Exchange (ETDEWEB)
Hoffmann, Alexander; Merk, Bruno [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany); Hirsch, Tobias; Pitz-Paal, Robert [DLR Deutsches Zentrum fuer Luft- und Raumfahrt e.V., Stuttgart (Germany). Inst. fuer Solarforschung
2014-06-15
In the present feasibility study the system code ATHLET, which originates from nuclear engineering, is applied to a parabolic trough test facility. A model of the DISS (DIrect Solar Steam) test facility at Plataforma Solar de Almeria in Spain is assembled and the results of the simulations are compared to measured data and the simulation results of the Modelica library 'DissDyn'. A profound comparison between ATHLET Mod 3.0 Cycle A and the 'DissDyn' library reveals the capabilities of these codes. The calculated mass and energy balance in the ATHLET simulations are in good agreement with the results of the measurements and confirm the applicability for thermodynamic simulations of DSG processes in principle. Supplementary, the capabilities of the 6-equation model with transient momentum balances in ATHLET are used to study the slip between liquid and gas phases and to investigate pressure wave oscillations after a sudden valve closure. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Kurosu, K [Department of Radiation Oncology, Osaka University Graduate School of Medicine, Osaka (Japan); Department of Medical Physics ' Engineering, Osaka University Graduate School of Medicine, Osaka (Japan); Takashina, M; Koizumi, M [Department of Medical Physics ' Engineering, Osaka University Graduate School of Medicine, Osaka (Japan); Das, I; Moskvin, V [Department of Radiation Oncology, Indiana University School of Medicine, Indianapolis, IN (United States)
2014-06-01
Purpose: Monte Carlo codes are becoming important tools for proton beam dosimetry. However, the relationships between the customizing parameters and percentage depth dose (PDD) of GATE and PHITS codes have not been reported which are studied for PDD and proton range compared to the FLUKA code and the experimental data. Methods: The beam delivery system of the Indiana University Health Proton Therapy Center was modeled for the uniform scanning beam in FLUKA and transferred identically into GATE and PHITS. This computational model was built from the blue print and validated with the commissioning data. Three parameters evaluated are the maximum step size, cut off energy and physical and transport model. The dependence of the PDDs on the customizing parameters was compared with the published results of previous studies. Results: The optimal parameters for the simulation of the whole beam delivery system were defined by referring to the calculation results obtained with each parameter. Although the PDDs from FLUKA and the experimental data show a good agreement, those of GATE and PHITS obtained with our optimal parameters show a minor discrepancy. The measured proton range R90 was 269.37 mm, compared to the calculated range of 269.63 mm, 268.96 mm, and 270.85 mm with FLUKA, GATE and PHITS, respectively. Conclusion: We evaluated the dependence of the results for PDDs obtained with GATE and PHITS Monte Carlo generalpurpose codes on the customizing parameters by using the whole computational model of the treatment nozzle. The optimal parameters for the simulation were then defined by referring to the calculation results. The physical model, particle transport mechanics and the different geometrybased descriptions need accurate customization in three simulation codes to agree with experimental data for artifact-free Monte Carlo simulation. This study was supported by Grants-in Aid for Cancer Research (H22-3rd Term Cancer Control-General-043) from the Ministry of Health
Radiological impact assessment in Malaysia using RESRAD computer code
International Nuclear Information System (INIS)
Syed Hakimi Sakuma Syed Ahmad; Khairuddin Mohamad Kontol; Razali Hamzah
1999-01-01
Radiological Impact Assessment (RIA) can be conducted in Malaysia by using the RESRAD computer code developed by Argonne National Laboratory, U.S.A. The code can do analysis to derive site specific guidelines for allowable residual concentrations of radionuclides in soil. Concepts of the RIA in the context of waste management concern in Malaysia, some regulatory information and assess status of data collection are shown. Appropriate use scenarios and site specific parameters are used as much as possible so as to be realistic so that will reasonably ensure that individual dose limits and or constraints will be achieved. Case study have been conducted to fulfil Atomic Energy Licensing Board (AELB) requirements where for disposal purpose the operator must be required to carry out. a radiological impact assessment to all proposed disposals. This is to demonstrate that no member of public will be exposed to more than 1 mSv/year from all activities. Results obtained from analyses show the RESRAD computer code is able to calculate doses, risks, and guideline values. Sensitivity analysis by the computer code shows that the parameters used as input are justified so as to improve confidence to the public and the AELB the results of the analysis. The computer code can also be used as an initial assessment to conduct screening assessment in order to determine a proper disposal site. (Author)
An object oriented code for simulating supersymmetric Yang-Mills theories
Catterall, Simon; Joseph, Anosh
2012-06-01
We present SUSY_LATTICE - a C++ program that can be used to simulate certain classes of supersymmetric Yang-Mills (SYM) theories, including the well known N=4 SYM in four dimensions, on a flat Euclidean space-time lattice. Discretization of SYM theories is an old problem in lattice field theory. It has resisted solution until recently when new ideas drawn from orbifold constructions and topological field theories have been brought to bear on the question. The result has been the creation of a new class of lattice gauge theories in which the lattice action is invariant under one or more supersymmetries. The resultant theories are local, free of doublers and also possess exact gauge-invariance. In principle they form the basis for a truly non-perturbative definition of the continuum SYM theories. In the continuum limit they reproduce versions of the SYM theories formulated in terms of twisted fields, which on a flat space-time is just a change of the field variables. In this paper, we briefly review these ideas and then go on to provide the details of the C++ code. We sketch the design of the code, with particular emphasis being placed on SYM theories with N=(2,2) in two dimensions and N=4 in three and four dimensions, making one-to-one comparisons between the essential components of the SYM theories and their corresponding counterparts appearing in the simulation code. The code may be used to compute several quantities associated with the SYM theories such as the Polyakov loop, mean energy, and the width of the scalar eigenvalue distributions. Program summaryProgram title: SUSY_LATTICE Catalogue identifier: AELS_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AELS_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC license, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 9315 No. of bytes in distributed program
A DOE Computer Code Toolbox: Issues and Opportunities
International Nuclear Information System (INIS)
Vincent, A.M. III
2001-01-01
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
Parallel computing by Monte Carlo codes MVP/GMVP
International Nuclear Information System (INIS)
Nagaya, Yasunobu; Nakagawa, Masayuki; Mori, Takamasa
2001-01-01
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)
Simulation of Water Chemistry using and Geochemistry Code, PHREEQE
Energy Technology Data Exchange (ETDEWEB)
Chi, J.H. [Korea Electric Power Research Institute, Taejeon (Korea)
2001-07-01
This report introduces principles and procedures of simulation for water chemistry using a geochemistry code, PHREEQE. As and example of the application of this code, we described the simulation procedure for titration of an aquatic sample with strong acid to investigate the state of Carbonates in aquatic solution. Major contents of this report are as follows; Concepts and principles of PHREEQE, Kinds of chemical reactions which may be properly simulated by PHREEQE, The definition and meaning of each input data, An example of simulation using PHREEQE. (author). 2 figs., 1 tab.
Utilization of KENO-IV computer code with HANSEN-ROACH library
International Nuclear Information System (INIS)
Lima Barros, M. de; Vellozo, S.O.
1982-01-01
Several analysis with KENO-IV computer code, which is based in the Monte Carlo method, and the cross section library HANSEN-ROACH, were done, aiming to present the more convenient form to execute criticality calculations with this computer code and this cross sections. (E.G.) [pt
Heat Transfer treatment in computer codes for safety analysis
International Nuclear Information System (INIS)
Jerele, A.; Gregoric, M.
1984-01-01
Increased number of operating nuclear power plants has stressed importance of nuclear safety evaluation. For this reason, accordingly to regulatory commission request, safety analyses with computer codes are preformed. In this paper part of this thermohydraulic models dealing with wall-to-fluid heat transfer correlations in computer codes TRAC=PF1, RELAP4/MOD5, RELAP5/MOD1 and COBRA-IV is discussed. (author)
Coupled CFD - system-code simulation of a gas cooled reactor
International Nuclear Information System (INIS)
Yan, Yizhou; Rizwan-uddin
2011-01-01
A generic coupled CFD - system-code thermal hydraulic simulation approach was developed based on FLUENT and RELAP-3D, and applied to LWRs. The flexibility of the coupling methodology enables its application to advanced nuclear energy systems. Gas Turbine - Modular Helium Reactor (GT-MHR) is a Gen IV reactor design which can benefit from this innovative coupled simulation approach. Mixing in the lower plenum of the GT-MHR is investigated here using the CFD - system-code coupled simulation tool. Results of coupled simulations are presented and discussed. The potential of the coupled CFD - system-code approach for next generation of nuclear power plants is demonstrated. (author)
Computing the Feng-Rao distances for codes from order domains
DEFF Research Database (Denmark)
Ruano Benito, Diego
2007-01-01
We compute the Feng–Rao distance of a code coming from an order domain with a simplicial value semigroup. The main tool is the Apéry set of a semigroup that can be computed using a Gröbner basis.......We compute the Feng–Rao distance of a code coming from an order domain with a simplicial value semigroup. The main tool is the Apéry set of a semigroup that can be computed using a Gröbner basis....
Development Of A Navier-Stokes Computer Code
Yoon, Seokkwan; Kwak, Dochan
1993-01-01
Report discusses aspects of development of CENS3D computer code, solving three-dimensional Navier-Stokes equations of compressible, viscous, unsteady flow. Implements implicit finite-difference or finite-volume numerical-integration scheme, called "lower-upper symmetric-Gauss-Seidel" (LU-SGS), offering potential for very low computer time per iteration and for fast convergence.
Directory of Open Access Journals (Sweden)
Almeida Jonas S
2006-03-01
Full Text Available Abstract Background Matlab, a powerful and productive language that allows for rapid prototyping, modeling and simulation, is widely used in computational biology. Modeling and simulation of large biological systems often require more computational resources then are available on a single computer. Existing distributed computing environments like the Distributed Computing Toolbox, MatlabMPI, Matlab*G and others allow for the remote (and possibly parallel execution of Matlab commands with varying support for features like an easy-to-use application programming interface, load-balanced utilization of resources, extensibility over the wide area network, and minimal system administration skill requirements. However, all of these environments require some level of access to participating machines to manually distribute the user-defined libraries that the remote call may invoke. Results mGrid augments the usual process distribution seen in other similar distributed systems by adding facilities for user code distribution. mGrid's client-side interface is an easy-to-use native Matlab toolbox that transparently executes user-defined code on remote machines (i.e. the user is unaware that the code is executing somewhere else. Run-time variables are automatically packed and distributed with the user-defined code and automated load-balancing of remote resources enables smooth concurrent execution. mGrid is an open source environment. Apart from the programming language itself, all other components are also open source, freely available tools: light-weight PHP scripts and the Apache web server. Conclusion Transparent, load-balanced distribution of user-defined Matlab toolboxes and rapid prototyping of many simple parallel applications can now be done with a single easy-to-use Matlab command. Because mGrid utilizes only Matlab, light-weight PHP scripts and the Apache web server, installation and configuration are very simple. Moreover, the web
Karpievitch, Yuliya V; Almeida, Jonas S
2006-03-15
Matlab, a powerful and productive language that allows for rapid prototyping, modeling and simulation, is widely used in computational biology. Modeling and simulation of large biological systems often require more computational resources then are available on a single computer. Existing distributed computing environments like the Distributed Computing Toolbox, MatlabMPI, Matlab*G and others allow for the remote (and possibly parallel) execution of Matlab commands with varying support for features like an easy-to-use application programming interface, load-balanced utilization of resources, extensibility over the wide area network, and minimal system administration skill requirements. However, all of these environments require some level of access to participating machines to manually distribute the user-defined libraries that the remote call may invoke. mGrid augments the usual process distribution seen in other similar distributed systems by adding facilities for user code distribution. mGrid's client-side interface is an easy-to-use native Matlab toolbox that transparently executes user-defined code on remote machines (i.e. the user is unaware that the code is executing somewhere else). Run-time variables are automatically packed and distributed with the user-defined code and automated load-balancing of remote resources enables smooth concurrent execution. mGrid is an open source environment. Apart from the programming language itself, all other components are also open source, freely available tools: light-weight PHP scripts and the Apache web server. Transparent, load-balanced distribution of user-defined Matlab toolboxes and rapid prototyping of many simple parallel applications can now be done with a single easy-to-use Matlab command. Because mGrid utilizes only Matlab, light-weight PHP scripts and the Apache web server, installation and configuration are very simple. Moreover, the web-based infrastructure of mGrid allows for it to be easily extensible over
International Nuclear Information System (INIS)
Kirillov, Igor R.; Obukhov, Denis M.; Ogorodnikov, Anatoly P.; Araseki, Hideo
2004-01-01
The paper describes and compares three computer codes that are able to estimate the double-supply-frequency (DSF) pulsations in annular linear induction pumps (ALIPs). The DSF pulsations are the result of interaction of the magnetic field and induced in liquid metal currents both changing with supply-frequency. They may be of some concern for electromagnetic pumps (EMP) exploitation and need to be evaluated at their design. The results of computer simulation are compared with experimental ones for annular linear induction pump ALIP-1
International Nuclear Information System (INIS)
Kljenak, I.
2001-01-01
Experiments on aerosol behavior in a vapor-saturated atmosphere, which were performed in the KAEVER experimental facility and proposed for the OECD International Standard Problem No. 44, were simulated with the CONTAIN thermal-hydraulic computer code. The purpose of the work was to assess the capability of the CONTAIN code to model aerosol condensation and deposition in a containment of a light-water-reactor nuclear power plant at severe accident conditions. Results of dry and wet aerosol concentrations are presented and analyzed.(author)
Control rod computer code IAMCOS: general theory and numerical methods
International Nuclear Information System (INIS)
West, G.
1982-11-01
IAMCOS is a computer code for the description of mechanical and thermal behavior of cylindrical control rods for fast breeders. This code version was applied, tested and modified from 1979 to 1981. In this report are described the basic model (02 version), theoretical definitions and computation methods [fr
KUPOL-M code for simulation of the VVER's accident localization system under LOCA conditions
International Nuclear Information System (INIS)
Efanov, A.D.; Lukyanov, A.A.; Shangin, N.N.; Zajtsev, A.A.; Solov'ev, S.L.
2004-01-01
Computer code KUPOL-M is developed for analysis of thermodynamic parameters of medium within full pressure containment for NPPs with VVER under LOCA conditions. The analysis takes into account the effects of non-stationary heat-mass transfer of gas-drop mixture in the containment compartments with natural convection, volume and surface steam condensation in the presence of noncondensables, heat-mass exchange of the compartment atmosphere with water in the sumps. The operation of the main safety systems like a spray system, hydrogen catalytic recombiners, emergency core cooling pumps, valves and a fan system is simulated in KUPOL-M code. The main results of the code verification including the ones of the participation in ISP-47 International Standard Problem on containment thermal-hydraulics are presented. (author)
Developments of HTGR thermofluid dynamic analysis codes and HTGR plant dynamic simulation code
International Nuclear Information System (INIS)
Tanaka, Mitsuhiro; Izaki, Makoto; Koike, Hiroyuki; Tokumitsu, Masashi
1983-01-01
In nuclear power plants as well as high temperature gas-cooled reactor plants, the design is mostly performed on the basis of the results after their characteristics have been grasped by carrying out the numerical simulation using the analysis code. Also in Kawasaki Heavy Industries Ltd., on the basis of the system engineering accumulated with gas-cooled reactors since several years ago, the preparation and systematization of analysis codes have been advanced, aiming at lining up the analysis codes for heat transferring flow and control characteristics, taking up HTGR plants as the main object. In this report, a part of the results is described. The example of the analysis applying the two-dimensional compressible flow analysis codes SOLA-VOF and SALE-2D, which were developed by Los Alamos National Laboratory in USA and modified for use in Kawasaki, to HTGR system is reported. Besides, Kawasaki has developed the control characteristics analyzing code DYSCO by which the change of system composition is easy and high versatility is available. The outline, fundamental equations, fundamental algorithms and examples of application of the SOLA-VOF and SALE-2D, the present status of system characteristic simulation codes and the outline of the DYSCO are described. (Kako, I.)
Thermohydraulic analysis of nuclear power plant accidents by computer codes
International Nuclear Information System (INIS)
Petelin, S.; Stritar, A.; Istenic, R.; Gregoric, M.; Jerele, A.; Mavko, B.
1982-01-01
RELAP4/MOD6, BRUCH-D-06, CONTEMPT-LT-28, RELAP5/MOD1 and COBRA-4-1 codes were successful y implemented at the CYBER 172 computer in Ljubljana. Input models of NPP Krsko for the first three codes were prepared. Because of the high computer cost only one analysis of double ended guillotine break of the cold leg of NPP Krsko by RELAP4 code has been done. BRUCH code is easier and cheaper for use. Several analysis have been done. Sensitivity study was performed with CONTEMPT-LT-28 for double ended pump suction break. These codes are intended to be used as a basis for independent safety analyses. (author)
Coupling methods for parallel running RELAPSim codes in nuclear power plant simulation
Energy Technology Data Exchange (ETDEWEB)
Li, Yankai; Lin, Meng, E-mail: linmeng@sjtu.edu.cn; Yang, Yanhua
2016-02-15
When the plant is modeled detailedly for high precision, it is hard to achieve real-time calculation for one single RELAP5 in a large-scale simulation. To improve the speed and ensure the precision of simulation at the same time, coupling methods for parallel running RELAPSim codes were proposed in this study. Explicit coupling method via coupling boundaries was realized based on a data-exchange and procedure-control environment. Compromise of synchronization frequency was well considered to improve the precision of simulation and guarantee the real-time simulation at the same time. The coupling methods were assessed using both single-phase flow models and two-phase flow models and good agreements were obtained between the splitting–coupling models and the integrated model. The mitigation of SGTR was performed as an integral application of the coupling models. A large-scope NPP simulator was developed adopting six splitting–coupling models of RELAPSim and other simulation codes. The coupling models could improve the speed of simulation significantly and make it possible for real-time calculation. In this paper, the coupling of the models in the engineering simulator is taken as an example to expound the coupling methods, i.e., coupling between parallel running RELAPSim codes, and coupling between RELAPSim code and other types of simulation codes. However, the coupling methods are also referable in other simulator, for example, a simulator employing ATHLETE instead of RELAP5, other logic code instead of SIMULINK. It is believed the coupling method is commonly used for NPP simulator regardless of the specific codes chosen in this paper.
The SEDA computer code and its utilization for Angra 1
International Nuclear Information System (INIS)
Fernandes Filho, T.L.
1988-11-01
The implementation of SEDA 2.0 computer code, developed at Ezeiza Atomic Center, Argentine for Angra 1 reactor is described. The SEDA code gives an estimate for radiological consequences of nuclear accidents with release of radiactive materials for the environment. This code is now available for an IBM PC-XT. The computer environment, the files used, data, the programining structure and the models used are presented. The input data and results for two sample case are described. (author) [pt
PC-Reactor-core transient simulation code
International Nuclear Information System (INIS)
Nakata, H.
1989-10-01
PC-REATOR, a reactor core transient simulation code has been developed for the real-time operator training on a IBM-PC microcomputer. The program presents capabilities for on-line exchange of the operating parameters during the transient simulation, by friendly keyboard instructions. The model is based on the point-kinetics approximation, with 2 delayed neutron percursors and up to 11 decay power generating groups. (author) [pt
Interface between computational fluid dynamics (CFD) and plant analysis computer codes
International Nuclear Information System (INIS)
Coffield, R.D.; Dunckhorst, F.F.; Tomlinson, E.T.; Welch, J.W.
1993-01-01
Computational fluid dynamics (CFD) can provide valuable input to the development of advanced plant analysis computer codes. The types of interfacing discussed in this paper will directly contribute to modeling and accuracy improvements throughout the plant system and should result in significant reduction of design conservatisms that have been applied to such analyses in the past
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.
Structure analysis of Pd(1 1 0) surface by computer simulation of NAACO's
International Nuclear Information System (INIS)
Takeuchi, Wataru; Yamamura, Yasunori
2001-01-01
The relaxation at a Pd(1 1 0) surface has been estimated using the computer simulation of 165 deg. neutral impact-collision ion scattering spectroscopy (NICISS). The computer simulations employing ACOCT program code treated three-dimensionally the atomic collisions and based on the binary collision approximation (BCA) were performed for the case of 2.08 keV Ne + ions incident along the [1 1-bar 2] azimuth of the Pd(1 1 0) surface. The experimental results of Speller et al. (Surf. Sci. 383 (1997) 131) are well reproduced by the ACOCT simulations including the inward relaxation of 14% of the first interlayer spacing and including the vertical component of surface Debye temperature of 125 K
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.
2015-01-01
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
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.
Comparative static simulations of a CANDU6 cell using different transport codes
Energy Technology Data Exchange (ETDEWEB)
Mahjoub, M.; Koclas, J., E-mail: mehdi.mahjoub@polymtl.ca [Ecole Polytechnique de Montreal, QC (Canada)
2015-07-01
The solution of the time dependent Boltzmann equation remains quite a challenge. We are in the process of developing such a method using the stochastic Monte Carlo approach for two reasons: First, at the cell level, we will be able to obtain time dependent homogenized cross sections for use in full core diffusion calculations. Second, the Monte Carlo methods are scalable to perform full core if and when appropriate computer resources become available. The Time dependent approach will be concretized a new module that will be added to an existing Monte Carlo code. As a first step towards this goal, we need to choose the initial Monte Carlo code to be used as start point. For this reason, we have compared results concerning the void reactivity of a fresh fuel CANDU6 cell, using two Monte Carlo codes, namely VTT developed SERPENT and MIT developed OpenMC together with the deterministic DRAGON code. Several libraries are used in this comparison. We conclude that OpenMC is a good candidate for implementation of a time dependent simulation. (author)
Computer simulation of atomic collision processes in solids
International Nuclear Information System (INIS)
Robinson, M.T.
1992-11-01
Computer simulation is a major tool for studying the interactions of swift ions with solids which underlie processes such as particle backscattering, ion implantation, radiation damage, and sputtering. Numerical models are classed as molecular dynamics or binary collision models, along with some intermediate types. Binary collision models are divided into those for crystalline targets and those for structureless ones. The foundations of such models are reviewed, including interatomic potentials, electron excitations, and relationships among the various types of codes. Some topics of current interest are summarized
Efficient Proximity Computation Techniques Using ZIP Code Data for Smart Cities †.
Murdani, Muhammad Harist; Kwon, Joonho; Choi, Yoon-Ho; Hong, Bonghee
2018-03-24
In this paper, we are interested in computing ZIP code proximity from two perspectives, proximity between two ZIP codes ( Ad-Hoc ) and neighborhood proximity ( Top-K ). Such a computation can be used for ZIP code-based target marketing as one of the smart city applications. A naïve approach to this computation is the usage of the distance between ZIP codes. We redefine a distance metric combining the centroid distance with the intersecting road network between ZIP codes by using a weighted sum method. Furthermore, we prove that the results of our combined approach conform to the characteristics of distance measurement. We have proposed a general and heuristic approach for computing Ad-Hoc proximity, while for computing Top-K proximity, we have proposed a general approach only. Our experimental results indicate that our approaches are verifiable and effective in reducing the execution time and search space.
Efficient Proximity Computation Techniques Using ZIP Code Data for Smart Cities †
Directory of Open Access Journals (Sweden)
Muhammad Harist Murdani
2018-03-01
Full Text Available In this paper, we are interested in computing ZIP code proximity from two perspectives, proximity between two ZIP codes (Ad-Hoc and neighborhood proximity (Top-K. Such a computation can be used for ZIP code-based target marketing as one of the smart city applications. A naïve approach to this computation is the usage of the distance between ZIP codes. We redefine a distance metric combining the centroid distance with the intersecting road network between ZIP codes by using a weighted sum method. Furthermore, we prove that the results of our combined approach conform to the characteristics of distance measurement. We have proposed a general and heuristic approach for computing Ad-Hoc proximity, while for computing Top-K proximity, we have proposed a general approach only. Our experimental results indicate that our approaches are verifiable and effective in reducing the execution time and search space.
Computer Simulation of the UMER Gridded Gun
Haber, Irving; Friedman, Alex; Grote, D P; Kishek, Rami A; Reiser, Martin; Vay, Jean-Luc; Zou, Yun
2005-01-01
The electron source in the University of Maryland Electron Ring (UMER) injector employs a grid 0.15 mm from the cathode to control the current waveform. Under nominal operating conditions, the grid voltage during the current pulse is sufficiently positive relative to the cathode potential to form a virtual cathode downstream of the grid. Three-dimensional computer simulations have been performed that use the mesh refinement capability of the WARP particle-in-cell code to examine a small region near the beam center in order to illustrate some of the complexity that can result from such a gridded structure. These simulations have been found to reproduce the hollowed velocity space that is observed experimentally. The simulations also predict a complicated time-dependent response to the waveform applied to the grid during the current turn-on. This complex temporal behavior appears to result directly from the dynamics of the virtual cathode formation and may therefore be representative of the expected behavior in...
International Nuclear Information System (INIS)
Bae, S. W.; Chung, B. D.
2010-07-01
This report records the results of the code validation for the one-dimensional module of the MARS-KS thermal hydraulics analysis code by means of result-comparison with the RELAP5/MOD3.3 computer code. For the validation calculations, simulations of the RELAP5 Code Developmental Assessment Problem, which consists of 22 simulation problems in 3 categories, have been selected. The results of the 3 categories of simulations demonstrate that the one-dimensional module of the MARS code and the RELAP5/MOD3.3 code are essentially the same code. This is expected as the two codes have basically the same set of field equations, constitutive equations and main thermal hydraulic models. The result suggests that the high level of code validity of the RELAP5/MOD3.3 can be directly applied to the MARS one-dimensional module
Grech, Mickael; Derouillat, J.; Beck, A.; Chiaramello, M.; Grassi, A.; Niel, F.; Perez, F.; Vinci, T.; Fle, M.; Aunai, N.; Dargent, J.; Plotnikov, I.; Bouchard, G.; Savoini, P.; Riconda, C.
2016-10-01
Over the last decades, Particle-In-Cell (PIC) codes have been central tools for plasma simulations. Today, new trends in High-Performance Computing (HPC) are emerging, dramatically changing HPC-relevant software design and putting some - if not most - legacy codes far beyond the level of performance expected on the new and future massively-parallel super computers. SMILEI is a new open-source PIC code co-developed by both plasma physicists and HPC specialists, and applied to a wide range of physics-related studies: from laser-plasma interaction to astrophysical plasmas. It benefits from an innovative parallelization strategy that relies on a super-domain-decomposition allowing for enhanced cache-use and efficient dynamic load balancing. Beyond these HPC-related developments, SMILEI also benefits from additional physics modules allowing to deal with binary collisions, field and collisional ionization and radiation back-reaction. This poster presents the SMILEI project, its HPC capabilities and illustrates some of the physics problems tackled with SMILEI.
Parallel Computing Characteristics of Two-Phase Thermal-Hydraulics code, CUPID
International Nuclear Information System (INIS)
Lee, Jae Ryong; Yoon, Han Young
2013-01-01
Parallelized CUPID code has proved to be able to reproduce multi-dimensional thermal hydraulic analysis by validating with various conceptual problems and experimental data. In this paper, the characteristics of the parallelized CUPID code were investigated. Both single- and two phase simulation are taken into account. Since the scalability of a parallel simulation is known to be better for fine mesh system, two types of mesh system are considered. In addition, the dependency of the preconditioner for matrix solver was also compared. The scalability for the single-phase flow is better than that for two-phase flow due to the less numbers of iterations for solving pressure matrix. The CUPID code was investigated the parallel performance in terms of scalability. The CUPID code was parallelized with domain decomposition method. The MPI library was adopted to communicate the information at the interface cells. As increasing the number of mesh, the scalability is improved. For a given mesh, single-phase flow simulation with diagonal preconditioner shows the best speedup. However, for the two-phase flow simulation, the ILU preconditioner is recommended since it reduces the overall simulation time
International Nuclear Information System (INIS)
Kljenak, I.; Skerlavaj, A.; Parzer, I.
1999-01-01
Hydrogen distribution during a severe accident in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN computer code. The accidents is initiated by a large-break loss-of-coolant accident which is nit successfully mitigated by the action of the emergency core cooling system. Cases with and without successful actuation of spray systems and fan coolers were considered. The simulations predicted hydrogen stratification within the containment main compartment with intensive hydrogen mixing in the containment dome region. Pressure and temperature responses were analyzed as well.(author)
Case studies in Gaussian process modelling of computer codes
International Nuclear Information System (INIS)
Kennedy, Marc C.; Anderson, Clive W.; Conti, Stefano; O'Hagan, Anthony
2006-01-01
In this paper we present a number of recent applications in which an emulator of a computer code is created using a Gaussian process model. Tools are then applied to the emulator to perform sensitivity analysis and uncertainty analysis. Sensitivity analysis is used both as an aid to model improvement and as a guide to how much the output uncertainty might be reduced by learning about specific inputs. Uncertainty analysis allows us to reflect output uncertainty due to unknown input parameters, when the finished code is used for prediction. The computer codes themselves are currently being developed within the UK Centre for Terrestrial Carbon Dynamics
Computer codes for beam dynamics analysis of cyclotronlike accelerators
Smirnov, V.
2017-12-01
Computer codes suitable for the study of beam dynamics in cyclotronlike (classical and isochronous cyclotrons, synchrocyclotrons, and fixed field alternating gradient) accelerators are reviewed. Computer modeling of cyclotron segments, such as the central zone, acceleration region, and extraction system is considered. The author does not claim to give a full and detailed description of the methods and algorithms used in the codes. Special attention is paid to the codes already proven and confirmed at the existing accelerating facilities. The description of the programs prepared in the worldwide known accelerator centers is provided. The basic features of the programs available to users and limitations of their applicability are described.
Latest improvements on TRACPWR six-equations thermohydraulic code
International Nuclear Information System (INIS)
Rivero, N.; Batuecas, T.; Martinez, R.; Munoz, J.; Lenhardt, G.; Serrano, P.
1999-01-01
The paper presents the latest improvements on TRACPWR aimed at adapting the code to present trends on computer platforms, architectures and training requirements as well as extending the scope of the code itself and its applicability to other technologies different from Westinghouse PWR one. Firstly major features of TRACPWR as best estimate and real time simulation code are summed, then the areas where TRACPWR is being improved are presented. These areas comprising: (1) Architecture: integrating TRACPWR and RELAP5 codes, (2) Code scope enhancement: modelling the Mid-Loop operation, (3) Code speed-up: applying parallelization techniques, (4) Code platform downswing: porting to Windows N1 platform, (5) On-line performance: allowing simulation initialisation from a Plant Process Computer, and (6) Code scope extension: using the code for modelling VVER and PHWR technology. (author)
Qiao, Shan; Jackson, Edward; Coussios, Constantin C; Cleveland, Robin O
2016-09-01
Nonlinear acoustics plays an important role in both diagnostic and therapeutic applications of biomedical ultrasound and a number of research and commercial software packages are available. In this manuscript, predictions of two solvers available in a commercial software package, pzflex, one using the finite-element-method (FEM) and the other a pseudo-spectral method, spectralflex, are compared with measurements and the Khokhlov-Zabolotskaya-Kuznetsov (KZK) Texas code (a finite-difference time-domain algorithm). The pzflex methods solve the continuity equation, momentum equation and equation of state where they account for nonlinearity to second order whereas the KZK code solves a nonlinear wave equation with a paraxial approximation for diffraction. Measurements of the field from a single element 3.3 MHz focused transducer were compared with the simulations and there was good agreement for the fundamental frequency and the harmonics; however the FEM pzflex solver incurred a high computational cost to achieve equivalent accuracy. In addition, pzflex results exhibited non-physical oscillations in the spatial distribution of harmonics when the amplitudes were relatively low. It was found that spectralflex was able to accurately capture the nonlinear fields at reasonable computational cost. These results emphasize the need to benchmark nonlinear simulations before using codes as predictive tools.
LATTICE: an interactive lattice computer code
International Nuclear Information System (INIS)
Staples, J.
1976-10-01
LATTICE is a computer code which enables an interactive user to calculate the functions of a synchrotron lattice. This program satisfies the requirements at LBL for a simple interactive lattice program by borrowing ideas from both TRANSPORT and SYNCH. A fitting routine is included
International Nuclear Information System (INIS)
Bestion, D.
2010-01-01
A multi-scale analysis of water-cooled reactor thermal hydraulics can be used to take advantage of increased computer power and improved simulation tools, including Direct Numerical Simulation (DNS), Computational Fluid Dynamics (CFD) (in both open and porous mediums), and system thermalhydraulic codes. This paper presents a general strategy for this procedure for various thermalhydraulic scales. A short state of the art is given for each scale, and the role of the scale in the overall multi-scale analysis process is defined. System thermalhydraulic codes will remain a privileged tool for many investigations related to safety. CFD in porous medium is already being frequently used for core thermal hydraulics, either in 3D modules of system codes or in component codes. CFD in open medium allows zooming on some reactor components in specific situations, and may be coupled to the system and component scales. Various modeling approaches exist in the domain from DNS to CFD which may be used to improve the understanding of flow processes, and as a basis for developing more physically based models for macroscopic tools. A few examples are given to illustrate the multi-scale approach. Perspectives for the future are drawn from the present state of the art and directions for future research and development are given
Simulation of the MHD stabilities of the experiment on HL-2A tokamak by GATO code
International Nuclear Information System (INIS)
Pan Wei; Chen Liaoyuan; Dong Jiaqi; Shen Yong; Zhang Jinhua
2009-01-01
The ideal two-dimensional MHD stabilities code, GATO, has been successfully immigrated to the high-performance computing system of HL-2A and used to the simulation study of the ideal MHD stabilities of the plasmas produced by one of the pellets injection experiments on HL-2A tokamak. The EFIT code was used to reconstruct the equilibrium configures firstly and the GATO was used to compute their MHD stabilities secondly whose source data were obtained by the NO.4050 discharge of the experiments on HL-2A, and finally by analyzing these results the preliminary conclusion was devised that the confinement performance of the plasma was improved because of the stabilization effect of the anti-sheared configures created by the pellets injection. (authors)
User's manual for computer code RIBD-II, a fission product inventory code
International Nuclear Information System (INIS)
Marr, D.R.
1975-01-01
The computer code RIBD-II is used to calculate inventories, activities, decay powers, and energy releases for the fission products generated in a fuel irradiation. Changes from the earlier RIBD code are: the expansion to include up to 850 fission product isotopes, input in the user-oriented NAMELIST format, and run-time choice of fuels from an extensively enlarged library of nuclear data. The library that is included in the code package contains yield data for 818 fission product isotopes for each of fourteen different fissionable isotopes, together with fission product transmutation cross sections for fast and thermal systems. Calculational algorithms are little changed from those in RIBD. (U.S.)
Migliorati, M
2015-01-01
The simulation of beam dynamics in presence of collective effects requires a strong computational effort to take into account, in a self consistent way, the wakefield acting on a given charge and produced by all the others. Generally this is done by means of a convolution integral or sum. Moreover, if the electromagnetic fields consist of resonant modes with high quality factors, responsible, for example, of coupled bunch instabilities, a charge is also affected by itself in previous turns, and a very long record of wakefield must be properly taken into account. In this paper we present a new simulation code for the longitudinal beam dynamics in a circular accelerator, which exploits an alternative approach to the currently used convolution sum, reducing the computing time and avoiding the issues related to the length of wakefield for coupled bunch instabilities. With this approach it is possible to simulate, without the need of a large computing power, simultaneously, the single and multi-bunch beam dynamics...
Development of code PRETOR for stellarator simulation
International Nuclear Information System (INIS)
Dies, J.; Fontanet, J.; Fontdecaba, J.M.; Castejon, F.; Alejandre, C.
1998-01-01
The Department de Fisica i Enginyeria Nuclear (DFEN) of the UPC has some experience in the development of the transport code PRETOR. This code has been validated with shots of DIII-D, JET and TFTR, it has also been used in the simulation of operational scenarios of ITER fast burnt termination. Recently, the association EURATOM-CIEMAT has started the operation of the TJ-II stellarator. Due to the need of validating the results given by others transport codes applied to stellarators and because all of them made some approximations, as a averaging magnitudes in each magnetic surface, it was thought suitable to adapt the PRETOR code to devices without axial symmetry, like stellarators, which is very suitable for the specific needs of the study of TJ-II. Several modifications are required in PRETOR; the main concerns to the models of: magnetic equilibrium, geometry and transport of energy and particles. In order to solve the complex magnetic equilibrium geometry the powerful numerical code VMEC has been used. This code gives the magnetic surface shape as a Fourier series in terms of the harmonics (m,n). Most of the geometric magnitudes are also obtained from the VMEC results file. The energy and particle transport models will be replaced by other phenomenological models that are better adapted to stellarator simulation. Using the proposed models, it is pretended to reproduce experimental data available from present stellarators, given especial attention to the TJ-II of the association EURATOM-CIEMAT. (Author)
Parallel and vector implementation of APROS simulator code
International Nuclear Information System (INIS)
Niemi, J.; Tommiska, J.
1990-01-01
In this paper the vector and parallel processing implementation of a general purpose simulator code is discussed. In this code the utilization of vector processing is straightforward. In addition to the loop level parallel processing, the functional decomposition and the domain decomposition have been considered. Results represented for a PWR-plant simulation illustrate the potential speed-up factors of the alternatives. It turns out that the loop level parallelism and the domain decomposition are the most promising alternative to employ the parallel processing. (author)
SWIMS: a small-angle multiple scattering computer code
International Nuclear Information System (INIS)
Sayer, R.O.
1976-07-01
SWIMS (Sigmund and WInterbon Multiple Scattering) is a computer code for calculation of the angular dispersion of ion beams that undergo small-angle, incoherent multiple scattering by gaseous or solid media. The code uses the tabulated angular distributions of Sigmund and Winterbon for a Thomas-Fermi screened Coulomb potential. The fraction of the incident beam scattered into a cone defined by the polar angle α is computed as a function of α for reduced thicknesses over the range 0.01 less than or equal to tau less than or equal to 10.0. 1 figure, 2 tables
International Nuclear Information System (INIS)
Eslinger, Paul W.; Engel, David W.; Gerhardstein, Lawrence H.; Lopresti, Charles A.; Nichols, William E.; Strenge, Dennis L.
2001-12-01
One activity of the Department of Energy's Groundwater/Vadose Zone Integration Project is an assessment of cumulative impacts from Hanford Site wastes on the subsurface environment and the Columbia River. Through the application of a system assessment capability (SAC), decisions for each cleanup and disposal action will be able to take into account the composite effect of other cleanup and disposal actions. The SAC has developed a suite of computer programs to simulate the migration of contaminants (analytes) present on the Hanford Site and to assess the potential impacts of the analytes, including dose to humans, socio-cultural impacts, economic impacts, and ecological impacts. The general approach to handling uncertainty in the SAC computer codes is a Monte Carlo approach. Conceptually, one generates a value for every stochastic parameter in the code (the entire sequence of modules from inventory through transport and impacts) and then executes the simulation, obtaining an output value, or result. This document provides user instructions for the SAC codes that handle inventory tracking, release of contaminants to the environment, and transport of contaminants through the unsaturated zone, saturated zone, and the Columbia River
War of ontology worlds: mathematics, computer code, or Esperanto?
Rzhetsky, Andrey; Evans, James A
2011-09-01
The use of structured knowledge representations-ontologies and terminologies-has become standard in biomedicine. Definitions of ontologies vary widely, as do the values and philosophies that underlie them. In seeking to make these views explicit, we conducted and summarized interviews with a dozen leading ontologists. Their views clustered into three broad perspectives that we summarize as mathematics, computer code, and Esperanto. Ontology as mathematics puts the ultimate premium on rigor and logic, symmetry and consistency of representation across scientific subfields, and the inclusion of only established, non-contradictory knowledge. Ontology as computer code focuses on utility and cultivates diversity, fitting ontologies to their purpose. Like computer languages C++, Prolog, and HTML, the code perspective holds that diverse applications warrant custom designed ontologies. Ontology as Esperanto focuses on facilitating cross-disciplinary communication, knowledge cross-referencing, and computation across datasets from diverse communities. We show how these views align with classical divides in science and suggest how a synthesis of their concerns could strengthen the next generation of biomedical ontologies.
DART: A simulation code for charged particle beams
International Nuclear Information System (INIS)
White, R.C.; Barr, W.L.; Moir, R.W.
1989-01-01
This paper presents a recently modified version of the 2-D code, DART, which can simulate the behavior of a beam of charged particles whose trajectories are determined by electric and magnetic fields. This code was originally used to design laboratory-scale and full-scale beam direct converters. Since then, its utility has been expanded to allow more general applications. The simulation includes space charge, secondary electrons, and the ionization of neutral gas. A beam can contain up to nine superimposed beamlets of different energy and species. The calculation of energy conversion efficiency and the method of specifying the electrode geometry are described. Basic procedures for using the code are given, and sample input and output fields are shown. 7 refs., 18 figs
Tokamak Simulation Code modeling of NSTX
International Nuclear Information System (INIS)
Jardin, S.C.; Kaye, S.; Menard, J.; Kessel, C.; Glasser, A.H.
2000-01-01
The Tokamak Simulation Code [TSC] is widely used for the design of new axisymmetric toroidal experiments. In particular, TSC was used extensively in the design of the National Spherical Torus eXperiment [NSTX]. The authors have now benchmarked TSC with initial NSTX results and find excellent agreement for plasma and vessel currents and magnetic flux loops when the experimental coil currents are used in the simulations. TSC has also been coupled with a ballooning stability code and with DCON to provide stability predictions for NSTX operation. TSC has also been used to model initial CHI experiments where a large poloidal voltage is applied to the NSTX vacuum vessel, causing a force-free current to appear in the plasma. This is a phenomenon that is similar to the plasma halo current that sometimes develops during a plasma disruption
Compendium of computer codes for the safety analysis of LMFBR's
International Nuclear Information System (INIS)
1975-06-01
A high level of mathematical sophistication is required in the safety analysis of LMFBR's to adequately meet the demands for realism and confidence in all areas of accident consequence evaluation. The numerical solution procedures associated with these analyses are generally so complex and time consuming as to necessitate their programming into computer codes. These computer codes have become extremely powerful tools for safety analysis, combining unique advantages in accuracy, speed and cost. The number, diversity and complexity of LMFBR safety codes in the U. S. has grown rapidly in recent years. It is estimated that over 100 such codes exist in various stages of development throughout the country. It is inevitable that such a large assortment of codes will require rigorous cataloguing and abstracting to aid individuals in identifying what is available. It is the purpose of this compendium to provide such a service through the compilation of code summaries which describe and clarify the status of domestic LMFBR safety codes. (U.S.)
User's manual for the NEFTRAN II computer code
International Nuclear Information System (INIS)
Olague, N.E.; Campbell, J.E.; Leigh, C.D.; Longsine, D.E.
1991-02-01
This document describes the NEFTRAN II (NEtwork Flow and TRANsport in Time-Dependent Velocity Fields) computer code and is intended to provide the reader with sufficient information to use the code. NEFTRAN II was developed as part of a performance assessment methodology for storage of high-level nuclear waste in unsaturated, welded tuff. NEFTRAN II is a successor to the NEFTRAN and NWFT/DVM computer codes and contains several new capabilities. These capabilities include: (1) the ability to input pore velocities directly to the transport model and bypass the network fluid flow model, (2) the ability to transport radionuclides in time-dependent velocity fields, (3) the ability to account for the effect of time-dependent saturation changes on the retardation factor, and (4) the ability to account for time-dependent flow rates through the source regime. In addition to these changes, the input to NEFTRAN II has been modified to be more convenient for the user. This document is divided into four main sections consisting of (1) a description of all the models contained in the code, (2) a description of the program and subprograms in the code, (3) a data input guide and (4) verification and sample problems. Although NEFTRAN II is the fourth generation code, this document is a complete description of the code and reference to past user's manuals should not be necessary. 19 refs., 33 figs., 25 tabs
Atomic-level computer simulation
International Nuclear Information System (INIS)
Adams, J.B.; Rockett, Angus; Kieffer, John; Xu Wei; Nomura, Miki; Kilian, K.A.; Richards, D.F.; Ramprasad, R.
1994-01-01
This paper provides a broad overview of the methods of atomic-level computer simulation. It discusses methods of modelling atomic bonding, and computer simulation methods such as energy minimization, molecular dynamics, Monte Carlo, and lattice Monte Carlo. ((orig.))
Light & Skin Interactions Simulations for Computer Graphics Applications
Baranoski, Gladimir V G
2010-01-01
Light and Skin Interactions immerses you in one of the most fascinating application areas of computer graphics: appearance simulation. The book first illuminates the fundamental biophysical processes that affect skin appearance, and reviews seminal related works aimed at applications in life and health sciences. It then examines four exemplary modeling approaches as well as definitive algorithms that can be used to generate realistic images depicting skin appearance. An accompanying companion site also includes complete code and data sources for the BioSpec model, which is considered to be the
Gender codes why women are leaving computing
Misa, Thomas J
2010-01-01
The computing profession is facing a serious gender crisis. Women are abandoning the computing field at an alarming rate. Fewer are entering the profession than anytime in the past twenty-five years, while too many are leaving the field in mid-career. With a maximum of insight and a minimum of jargon, Gender Codes explains the complex social and cultural processes at work in gender and computing today. Edited by Thomas Misa and featuring a Foreword by Linda Shafer, Chair of the IEEE Computer Society Press, this insightful collection of essays explores the persisting gender imbalance in computing and presents a clear course of action for turning things around.
The computer simulation of 3d gas dynamics in a gas centrifuge
Borman, V. D.; Bogovalov, S. V.; Borisevich, V. D.; Tronin, I. V.; Tronin, V. N.
2016-09-01
We argue on the basis of the results of 2D analysis of the gas flow in gas centrifuges that a reliable calculation of the circulation of the gas and gas content in the gas centrifuge is possible only in frameworks of 3D numerical simulation of gas dynamics in the gas centrifuge (hereafter GC). The group from National research nuclear university, MEPhI, has created a computer code for 3D simulation of the gas flow in GC. The results of the computer simulations of the gas flows in GC are presented. A model Iguassu centrifuge is explored for the simulations. A nonaxisymmetric gas flow is produced due to interaction of the hypersonic rotating flow with the scoops for extraction of the product and waste flows from the GC. The scoops produce shock waves penetrating into a working camera of the GC and form spiral waves there.
The computer simulation of 3d gas dynamics in a gas centrifuge
International Nuclear Information System (INIS)
Borman, V D; Bogovalov, S V; Borisevich, V D; Tronin, I V; Tronin, V N
2016-01-01
We argue on the basis of the results of 2D analysis of the gas flow in gas centrifuges that a reliable calculation of the circulation of the gas and gas content in the gas centrifuge is possible only in frameworks of 3D numerical simulation of gas dynamics in the gas centrifuge (hereafter GC). The group from National research nuclear university, MEPhI, has created a computer code for 3D simulation of the gas flow in GC. The results of the computer simulations of the gas flows in GC are presented. A model Iguassu centrifuge is explored for the simulations. A nonaxisymmetric gas flow is produced due to interaction of the hypersonic rotating flow with the scoops for extraction of the product and waste flows from the GC. The scoops produce shock waves penetrating into a working camera of the GC and form spiral waves there. (paper)
Marcolongo, Juan P.; Zeida, Ari; Semelak, Jonathan A.; Foglia, Nicolás O.; Morzan, Uriel N.; Estrin, Dario A.; González Lebrero, Mariano C.; Scherlis, Damián A.
2018-03-01
In this work we present the current advances in the development and the applications of LIO, a lab-made code designed for density functional theory calculations in graphical processing units (GPU), that can be coupled with different classical molecular dynamics engines. This code has been thoroughly optimized to perform efficient molecular dynamics simulations at the QM/MM DFT level, allowing for an exhaustive sampling of the configurational space. Selected examples are presented for the description of chemical reactivity in terms of free energy profiles, and also for the computation of optical properties, such as vibrational and electronic spectra in solvent and protein environments.
OpenQ∗D simulation code for QCD+QED
DEFF Research Database (Denmark)
Campos, Isabel; Fritzsch, Patrick; Hansen, Martin
2018-01-01
The openQ∗D code for the simulation of QCD+QED with C∗ boundary conditions is presented. This code is based on openQCD-1.6, from which it inherits the core features that ensure its efficiency: the locally-deflated SAP-preconditioned GCR solver, the twisted-mass frequency splitting of the fermion....... An alpha version of this code is publicly available and can be downloaded from http://rcstar.web.cern.ch/....
A methodology for the rigorous verification of plasma simulation codes
Riva, Fabio
2016-10-01
The methodology used to assess the reliability of numerical simulation codes constitutes the Verification and Validation (V&V) procedure. V&V is composed by two separate tasks: the verification, which is a mathematical issue targeted to assess that the physical model is correctly solved, and the validation, which determines the consistency of the code results, and therefore of the physical model, with experimental data. In the present talk we focus our attention on the verification, which in turn is composed by the code verification, targeted to assess that a physical model is correctly implemented in a simulation code, and the solution verification, that quantifies the numerical error affecting a simulation. Bridging the gap between plasma physics and other scientific domains, we introduced for the first time in our domain a rigorous methodology for the code verification, based on the method of manufactured solutions, as well as a solution verification based on the Richardson extrapolation. This methodology was applied to GBS, a three-dimensional fluid code based on a finite difference scheme, used to investigate the plasma turbulence in basic plasma physics experiments and in the tokamak scrape-off layer. Overcoming the difficulty of dealing with a numerical method intrinsically affected by statistical noise, we have now generalized the rigorous verification methodology to simulation codes based on the particle-in-cell algorithm, which are employed to solve Vlasov equation in the investigation of a number of plasma physics phenomena.
Commercial applications of large-scale Research and Development computer simulation technologies
International Nuclear Information System (INIS)
Kuok Mee Ling; Pascal Chen; Wen Ho Lee
1998-01-01
The potential commercial applications of two large-scale R and D computer simulation technologies are presented. One such technology is based on the numerical solution of the hydrodynamics equations, and is embodied in the two-dimensional Eulerian code EULE2D, which solves the hydrodynamic equations with various models for the equation of state (EOS), constitutive relations and fracture mechanics. EULE2D is an R and D code originally developed to design and analyze conventional munitions for anti-armor penetrations such as shaped charges, explosive formed projectiles, and kinetic energy rods. Simulated results agree very well with actual experiments. A commercial application presented here is the design and simulation of shaped charges for oil and gas well bore perforation. The other R and D simulation technology is based on the numerical solution of Maxwell's partial differential equations of electromagnetics in space and time, and is implemented in the three-dimensional code FDTD-SPICE, which solves Maxwell's equations in the time domain with finite-differences in the three spatial dimensions and calls SPICE for information when nonlinear active devices are involved. The FDTD method has been used in the radar cross-section modeling of military aircrafts and many other electromagnetic phenomena. The coupling of FDTD method with SPICE, a popular circuit and device simulation program, provides a powerful tool for the simulation and design of microwave and millimeter-wave circuits containing nonlinear active semiconductor devices. A commercial application of FDTD-SPICE presented here is the simulation of a two-element active antenna system. The simulation results and the experimental measurements are in excellent agreement. (Author)
Statistical screening of input variables in a complex computer code
International Nuclear Information System (INIS)
Krieger, T.J.
1982-01-01
A method is presented for ''statistical screening'' of input variables in a complex computer code. The object is to determine the ''effective'' or important input variables by estimating the relative magnitudes of their associated sensitivity coefficients. This is accomplished by performing a numerical experiment consisting of a relatively small number of computer runs with the code followed by a statistical analysis of the results. A formula for estimating the sensitivity coefficients is derived. Reference is made to an earlier work in which the method was applied to a complex reactor code with good results
Simulation models for computational plasma physics: Concluding report
International Nuclear Information System (INIS)
Hewett, D.W.
1994-01-01
In this project, the authors enhanced their ability to numerically simulate bounded plasmas that are dominated by low-frequency electric and magnetic fields. They moved towards this goal in several ways; they are now in a position to play significant roles in the modeling of low-frequency electromagnetic plasmas in several new industrial applications. They have significantly increased their facility with the computational methods invented to solve the low frequency limit of Maxwell's equations (DiPeso, Hewett, accepted, J. Comp. Phys., 1993). This low frequency model is called the Streamlined Darwin Field model (SDF, Hewett, Larson, and Doss, J. Comp. Phys., 1992) has now been implemented in a fully non-neutral SDF code BEAGLE (Larson, Ph.D. dissertation, 1993) and has further extended to the quasi-neutral limit (DiPeso, Hewett, Comp. Phys. Comm., 1993). In addition, they have resurrected the quasi-neutral, zero-electron-inertia model (ZMR) and began the task of incorporating internal boundary conditions into this model that have the flexibility of those in GYMNOS, a magnetostatic code now used in ion source work (Hewett, Chen, ICF Quarterly Report, July--September, 1993). Finally, near the end of this project, they invented a new type of banded matrix solver that can be implemented on a massively parallel computer -- thus opening the door for the use of all their ADI schemes on these new computer architecture's (Mattor, Williams, Hewett, submitted to Parallel Computing, 1993)
A restructuring of CF package for MIDAS computer code
International Nuclear Information System (INIS)
Park, S. H.; Kim, K. R.; Kim, D. H.; Cho, S. W.
2004-01-01
CF package, which evaluates user-specified 'control functions' and applies them to define or control various aspects