PERIGEE computer codes for reactor simulation in 3 dimensions, using 1 or 2 neutron velocity groups
International Nuclear Information System (INIS)
Olson, A.P.
1964-02-01
PERIGEE is a code written in SNAP for the G-20 computer. It solves the one- or two-group neutron diffusion equations by finite-difference methods on a three-dimensional, uniform mesh having a common spacing in the two directions normal to the fuel channels. The positions of mesh points along a fuel channel, relative to points in adjacent channels, may correspond to either NPD or CANDU fuel bundle positions. The extrapolated flux boundary may be specified in sufficient detail to represent a tapered or stepped circumferential reflector, a variable axial length and, for a reactor with axis horizontal, a variable moderator level and a variable plane bottom surface equivalent to the CANDU dump structure. The neutron flux may be normalized to give a specified power output from the hottest fuel bundle or hottest channel, or to give a total thermal power limited by the turbine and generator. Reactor operation may be simulated in finite time steps, taking into account any fuel shifts, any changes in moderator level and the change in nuclear properties of the fuel with increasing irradiation. The appropriate properties are obtained by interpolation from tables supplied for as many as 8 types of fuel bundle. The mean fuel exit burnup can be calculated at equilibrium for a reactor in which the exit burnups for two zones may be adjusted to give radial power flattening and the fuelling schedules may be designed to give axial power flattening in one or both zones. (author)
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
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
LFSC - Linac Feedback Simulation Code
Energy Technology Data Exchange (ETDEWEB)
Ivanov, Valentin; /Fermilab
2008-05-01
The computer program LFSC (
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.
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
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
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)
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
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.
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...
Group representations, error bases and quantum codes
Energy Technology Data Exchange (ETDEWEB)
Knill, E
1996-01-01
This report continues the discussion of unitary error bases and quantum codes. Nice error bases are characterized in terms of the existence of certain characters in a group. A general construction for error bases which are non-abelian over the center is given. The method for obtaining codes due to Calderbank et al. is generalized and expressed purely in representation theoretic terms. The significance of the inertia subgroup both for constructing codes and obtaining the set of transversally implementable operations is demonstrated.
The Los Alamos accelerator code group
Energy Technology Data Exchange (ETDEWEB)
Krawczyk, F.L.; Billen, J.H.; Ryne, R.D.; Takeda, Harunori; Young, L.M.
1995-05-01
The Los Alamos Accelerator Code Group (LAACG) is a national resource for members of the accelerator community who use and/or develop software for the design and analysis of particle accelerators, beam transport systems, light sources, storage rings, and components of these systems. Below the authors describe the LAACG`s activities in high performance computing, maintenance and enhancement of POISSON/SUPERFISH and related codes and the dissemination of information on the INTERNET.
The Los Alamos accelerator code group
International Nuclear Information System (INIS)
Krawczyk, F.L.; Billen, J.H.; Ryne, R.D.; Takeda, Harunori; Young, L.M.
1995-01-01
The Los Alamos Accelerator Code Group (LAACG) is a national resource for members of the accelerator community who use and/or develop software for the design and analysis of particle accelerators, beam transport systems, light sources, storage rings, and components of these systems. Below the authors describe the LAACG's activities in high performance computing, maintenance and enhancement of POISSON/SUPERFISH and related codes and the dissemination of information on the INTERNET
Improvement of group collapsing in TRANSX code
International Nuclear Information System (INIS)
Jeong, Hyun Tae; Kim, Young Cheol; Kim, Young In; Kim, Young Kyun
1996-07-01
A cross section generating and processing computer code TRANSX version 2.15 in the K-CORE system, being developed by the KAERI LMR core design technology development team produces various cross section input files appropriated for flux calculation options from the cross section library MATXS. In this report, a group collapsing function of TRANSX has been improved to utilize the zone averaged flux file RZFLUX written in double precision as flux weighting functions. As a result, an iterative calculation system using double precision RZFLUX consisting of the cross section data library file MATXS, the effective cross section producing and processing code TRANSX, and the transport theory calculation code TWODANT has been set up and verified through a sample model calculation. 4 refs. (Author)
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)
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)
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)
International Nuclear Information System (INIS)
Pellacani, F.; Macian, R.; Chiva, S.; Pena, C.
2011-01-01
In this paper upward, isothermal and turbulent bubbly flow in tubes is numerically modeled by using ANSYS CFX 12.1 with the aim of creating a basis for the reliable simulation of the flow along a vertical channel in a nuclear reactor as long term goal. Two approaches based on the mono-dispersed model and on the one-group Interfacial Area Transport Equation (IATE) model are used in order to maintain the computational effort as low as possible. This work represents the necessary step to implement a two-group interfacial area transport equation that will be able to dynamically represent the changes in interfacial structure in the transition region from bubbly to slug flow. The drag coefficient is calculated using the Grace model and the interfacial non-drag forces are also included. The Antal model is used for the calculation of the wall lubrication force coefficient. The lift force coefficient is obtained from the Tomiyama model. The turbulent dispersion force is taken into account and is modeled using the FAD (Favre averaged drag) approach, while the turbulence transfer is simulated with the Sato's model. The liquid velocity is in the range between 0.5 and 2 m/s and the average void fraction varies between 5 and 15%.The source and sink terms for break-up and coalescence needed for the calculation of the implemented Interfacial Area Density are those proposed by Yao and Morel. The model has been checked using experimental results by Mendez. Radial profile distributions of void fraction, interfacial area density and bubble mean diameter are shown at the axial position equivalent to z/D=56. The results obtained by the simulations have a good agreement with the experimental data but show also the need of a better study of the coalescence and breakup phenomena to develop more accurate interaction models. (author)
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.
LACEwING: A New Moving Group Analysis Code
Energy Technology Data Exchange (ETDEWEB)
Riedel, Adric R. [Department of Astronomy, California Institute of Technology, Pasadena, CA 91125 (United States); Blunt, Sarah C.; Faherty, Jacqueline K. [Department of Astrophysics, American Museum of Natural History, New York, NY 10024 (United States); Lambrides, Erini L. [Department of Physics and Astronomy, Johns Hopkins University, Baltimore, MD 21218 (United States); Rice, Emily L. [Department of Engineering Science and Physics, The College of Staten Island, Staten Island, NY 10314 (United States); Cruz, Kelle L., E-mail: arr@astro.caltech.edu [Department of Physics and Astronomy, Hunter College, New York, NY 10065 (United States)
2017-03-01
We present a new nearby young moving group (NYMG) kinematic membership analysis code, LocAting Constituent mEmbers In Nearby Groups (LACEwING), a new Catalog of Suspected Nearby Young Stars, a new list of bona fide members of moving groups, and a kinematic traceback code. LACEwING is a convergence-style algorithm with carefully vetted membership statistics based on a large numerical simulation of the Solar Neighborhood. Given spatial and kinematic information on stars, LACEwING calculates membership probabilities in 13 NYMGs and three open clusters within 100 pc. In addition to describing the inputs, methods, and products of the code, we provide comparisons of LACEwING to other popular kinematic moving group membership identification codes. As a proof of concept, we use LACEwING to reconsider the membership of 930 stellar systems in the Solar Neighborhood (within 100 pc) that have reported measurable lithium equivalent widths. We quantify the evidence in support of a population of young stars not attached to any NYMGs, which is a possible sign of new as-yet-undiscovered groups or of a field population of young stars.
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
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)
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).
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)
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)
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.
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
Energy Technology Data Exchange (ETDEWEB)
Xiao, Hailin [Wenzhou University, College of Physics and Electronic Information Engineering, Wenzhou (China); Southeast University, National Mobile Communications Research Laboratory, Nanjing (China); Guilin University of Electronic Technology, Ministry of Education, Key Laboratory of Cognitive Radio and Information Processing, Guilin (China); Zhang, Zhongshan [University of Science and Technology Beijing, Beijing Engineering and Technology Research Center for Convergence Networks and Ubiquitous Services, Beijing (China); Chronopoulos, Anthony Theodore [University of Texas at San Antonio, Department of Computer Science, San Antonio, TX (United States)
2017-10-15
In quantum computing, nice error bases as generalization of the Pauli basis were introduced by Knill. These bases are known to be projective representations of finite groups. In this paper, we propose a group representation approach to the study of quantum stabilizer codes. We utilize this approach to define decoherence-free subspaces (DFSs). Unlike previous studies of DFSs, this type of DFSs does not involve any spatial symmetry assumptions on the system-environment interaction. Thus, it can be used to construct quantum error-avoiding codes (QEACs) that are fault tolerant automatically. We also propose a new simple construction of QEACs and subsequently develop several classes of QEACs. Finally, we present numerical simulation results encoding the logical error rate over physical error rate on the fidelity performance of these QEACs. Our study demonstrates that DFSs-based QEACs are capable of providing a generalized and unified framework for error-avoiding methods. (orig.)
MUSIC: a mesh-unrestricted simulation code
International Nuclear Information System (INIS)
Bonalumi, R.A.; Rouben, B.; Dastur, A.R.; Dondale, C.S.; Li, H.Y.H.
1978-01-01
A general formalism to solve the G-group neutron diffusion equation is described. The G-group flux is represented by complementing an ''asymptotic'' mode with (G-1) ''transient'' modes. A particular reduction-to-one-group technique gives a high computational efficiency. MUSIC, a 2-group code using the above formalism, is presented. MUSIC is demonstrated on a fine-mesh calculation and on 2 coarse-mesh core calculations: a heavy-water reactor (HWR) problem and the 2-D lightwater reactor (LWR) IAEA benchmark. Comparison is made to finite-difference results
Codes maintained by the LAACG [Los Alamos Accelerator Code Group] at the NMFECC
International Nuclear Information System (INIS)
Wallace, R.; Barts, T.
1990-01-01
The Los Alamos Accelerator Code Group (LAACG) maintains two groups of design codes at the National Magnetic Fusion Energy Computing Center (NMFECC). These codes, principally electromagnetic field solvers, are used for the analysis and design of electromagnetic components for accelerators, e.g., magnets, rf structures, pickups, etc. In this paper, the status and future of the installed codes will be discussed with emphasis on an experimental version of one set of codes, POISSON/SUPERFISH
Infinity-Norm Permutation Covering Codes from Cyclic Groups
Karni, Ronen; Schwartz, Moshe
2017-01-01
We study covering codes of permutations with the $\\ell_\\infty$-metric. We provide a general code construction, which uses smaller building-block codes. We study cyclic transitive groups as building blocks, determining their exact covering radius, and showing linear-time algorithms for finding a covering codeword. We also bound the covering radius of relabeled cyclic transitive groups under conjugation.
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
Quantum Codes From Negacyclic Codes over Group Ring ( Fq + υFq) G
International Nuclear Information System (INIS)
Koroglu, Mehmet E.; Siap, Irfan
2016-01-01
In this paper, we determine self dual and self orthogonal codes arising from negacyclic codes over the group ring ( F q + υF q ) G . By taking a suitable Gray image of these codes we obtain many good parameter quantum error-correcting codes over F q . (paper)
Directory of Open Access Journals (Sweden)
Vipin Balyan
2014-08-01
Full Text Available Orthogonal variable spreading factor codes are used in the downlink to maintain the orthogonality between different channels and are used to handle new calls arriving in the system. A period of operation leads to fragmentation of vacant codes. This leads to code blocking problem. The assignment scheme proposed in this paper is not affected by fragmentation, as the fragmentation is generated by the scheme itself. In this scheme, the code tree is divided into groups whose capacity is fixed and numbers of members (codes are variable. A group with maximum number of busy members is used for assignment, this leads to fragmentation of busy groups around code tree and compactness within group. The proposed scheme is well evaluated and compared with other schemes using parameters like code blocking probability and call establishment delay. Through simulations it has been demonstrated that the proposed scheme not only adequately reduces code blocking probability, but also requires significantly less time before assignment to locate a vacant code for assignment, which makes it suitable for the real-time calls.
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.
On the use of SERPENT Monte Carlo code to generate few group diffusion constants
Energy Technology Data Exchange (ETDEWEB)
Piovezan, Pamela, E-mail: pamela.piovezan@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Carluccio, Thiago; Domingos, Douglas Borges; Rossi, Pedro Russo; Mura, Luiz Felipe, E-mail: fermium@cietec.org.b, E-mail: thiagoc@ipen.b [Fermium Tecnologia Nuclear, Sao Paulo, SP (Brazil); Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)
2011-07-01
The accuracy of diffusion reactor codes strongly depends on the quality of the groups constants processing. For many years, the generation of such constants was based on 1-D infinity cell transport calculations. Some developments using collision probability or the method of characteristics allow, nowadays, 2-D assembly group constants calculations. However, these 1-D and 2-D codes how some limitations as , for example, on complex geometries and in the neighborhood of heavy absorbers. On the other hand, since Monte Carlos (MC) codes provide accurate neutro flux distributions, the possibility of using these solutions to provide group constants to full-core reactor diffusion simulators has been recently investigated, especially for the cases in which the geometry and reactor types are beyond the capability of the conventional deterministic lattice codes. The two greatest difficulties on the use of MC codes to group constant generation are the computational costs and the methodological incompatibility between analog MC particle transport simulation and deterministic transport methods based in several approximations. The SERPENT code is a 3-D continuous energy MC transport code with built-in burnup capability that was specially optimized to generate these group constants. In this work, we present the preliminary results of using the SERPENT MC code to generate 3-D two-group diffusion constants for a PWR like assembly. These constants were used in the CITATION diffusion code to investigate the effects of the MC group constants determination on the neutron multiplication factor diffusion estimate. (author)
Development code for group constant processing
International Nuclear Information System (INIS)
Su'ud, Z.
1997-01-01
In this paper methods, formalism and algorithm related to group constant processing problem from basic library such as ENDF/B VI will be described. Basically the problems can be grouped as follows; the treatment of resolved resonance using NR approximation, the treatment of unresolved resonance using statistical method, the treatment of low lying resonance using intermediate resonance approximation, the treatment of thermal energy regions, and the treatment group transfer matrices cross sections. it is necessary to treat interference between resonance properly especially in the unresolved region. in this paper the resonance problems are treated based on Breit-wigner method, and doppler function is treated using Pade approximation for calculation efficiency. finally, some samples of calculational result for some nuclei, mainly for comparison between many methods are discussed in this paper
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
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
Paracantor: A two group, two region reactor code
Energy Technology Data Exchange (ETDEWEB)
Stone, Stuart
1956-07-01
Paracantor I a two energy group, two region, time independent reactor code, which obtains a closed solution for a critical reactor assembly. The code deals with cylindrical reactors of finite length and with a radial reflector of finite thickness. It is programmed for the 1.B.M: Magnetic Drum Data-Processing Machine, Type 650. The limited memory space available does not permit a flux solution to be included in the basic Paracantor code. A supplementary code, Paracantor 11, has been programmed which computes fluxes, .including adjoint fluxes, from the .output of Paracamtor I.
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.
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)
Multi-group diffusion perturbation calculation code. PERKY (2002)
Energy Technology Data Exchange (ETDEWEB)
Iijima, Susumu; Okajima, Shigeaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2002-12-01
Perturbation calculation code based on the diffusion theory ''PERKY'' is designed for nuclear characteristic analyses of fast reactor. The code calculates reactivity worth on the multi-group diffusion perturbation theory in two or three dimensional core model and kinetics parameters such as effective delayed neutron fraction, prompt neutron lifetime and absolute reactivity scale factor ({rho}{sub 0} {delta}k/k) for FCA experiments. (author)
APR1400 Containment Simulation with CONTAIN code
Energy Technology Data Exchange (ETDEWEB)
Hwang, Moon Kyu; Chung, Bub Dong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2010-05-15
The more realistic containment pressure variation predicted by the CONTAIN code through the coupled analysis during a large break loss of coolant accident in the nuclear power plant is expected to provide more accurate prediction for the plant behavior than a standalone MARS-KS calculation. The input deck has been generated based on the already available ARP- 1400 input for CONTEMPT code. Similarly to the CONTEMPT input deck, a simple two-cell model was adopted to model the containment behavior, one cell for the containment inner volume and another cell for the environment condition. The developed input for the CONTAIN code is to be eventually applied for the coupled code calculation of MARS-KS/CONTAIN
APR1400 Containment Simulation with CONTAIN code
International Nuclear Information System (INIS)
Hwang, Moon Kyu; Chung, Bub Dong
2010-01-01
The more realistic containment pressure variation predicted by the CONTAIN code through the coupled analysis during a large break loss of coolant accident in the nuclear power plant is expected to provide more accurate prediction for the plant behavior than a standalone MARS-KS calculation. The input deck has been generated based on the already available ARP- 1400 input for CONTEMPT code. Similarly to the CONTEMPT input deck, a simple two-cell model was adopted to model the containment behavior, one cell for the containment inner volume and another cell for the environment condition. The developed input for the CONTAIN code is to be eventually applied for the coupled code calculation of MARS-KS/CONTAIN
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
Topological color codes on Union Jack lattices: a stable implementation of the whole Clifford group
International Nuclear Information System (INIS)
Katzgraber, Helmut G.; Bombin, H.; Andrist, Ruben S.; Martin-Delgado, M. A.
2010-01-01
We study the error threshold of topological color codes on Union Jack lattices that allow for the full implementation of the whole Clifford group of quantum gates. After mapping the error-correction process onto a statistical mechanical random three-body Ising model on a Union Jack lattice, we compute its phase diagram in the temperature-disorder plane using Monte Carlo simulations. Surprisingly, topological color codes on Union Jack lattices have a similar error stability to color codes on triangular lattices, as well as to the Kitaev toric code. The enhanced computational capabilities of the topological color codes on Union Jack lattices with respect to triangular lattices and the toric code combined with the inherent robustness of this implementation show good prospects for future stable quantum computer implementations.
Improved mesh generator for the POISSON Group Codes
International Nuclear Information System (INIS)
Gupta, R.C.
1987-01-01
This paper describes the improved mesh generator of the POISSON Group Codes. These improvements enable one to have full control over the way the mesh is generated and in particular the way the mesh density is distributed throughout this model. A higher mesh density in certain regions coupled with a successively lower mesh density in others keeps the accuracy of the field computation high and the requirements on the computer time and computer memory low. The mesh is generated with the help of codes AUTOMESH and LATTICE; both have gone through a major upgrade. Modifications have also been made in the POISSON part of these codes. We shall present an example of a superconducting dipole magnet to explain how to use this code. The results of field computations are found to be reliable within a few parts in a hundred thousand even in such complex geometries
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)
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
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/....
Physical model of the nuclear fuel cycle simulation code SITON
International Nuclear Information System (INIS)
Brolly, Á.; Halász, M.; Szieberth, M.; Nagy, L.; Fehér, S.
2017-01-01
Finding answers to main challenges of nuclear energy, like resource utilisation or waste minimisation, calls for transient fuel cycle modelling. This motivation led to the development of SITON v2.0 a dynamic, discrete facilities/discrete materials and also discrete events fuel cycle simulation code. The physical model of the code includes the most important fuel cycle facilities. Facilities can be connected flexibly; their number is not limited. Material transfer between facilities is tracked by taking into account 52 nuclides. Composition of discharged fuel is determined using burnup tables except for the 2400 MW thermal power design of the Gas-Cooled Fast Reactor (GFR2400). For the GFR2400 the FITXS method is used, which fits one-group microscopic cross-sections as polynomial functions of the fuel composition. This method is accurate and fast enough to be used in fuel cycle simulations. Operation of the fuel cycle, i.e. material requests and transfers, is described by discrete events. In advance of the simulation reactors and plants formulate their requests as events; triggered requests are tracked. After that, the events are simulated, i.e. the requests are fulfilled and composition of the material flow between facilities is calculated. To demonstrate capabilities of SITON v2.0, a hypothetical transient fuel cycle is presented in which a 4-unit VVER-440 reactor park was replaced by one GFR2400 that recycled its own spent fuel. It is found that the GFR2400 can be started if the cooling time of its spent fuel is 2 years. However, if the cooling time is 5 years it needs an additional plutonium feed, which can be covered from the spent fuel of a Generation III light water reactor.
A novel QC-LDPC code based on the finite field multiplicative group for optical communications
Yuan, Jian-guo; Xu, Liang; Tong, Qing-zhen
2013-09-01
A novel construction method of quasi-cyclic low-density parity-check (QC-LDPC) code is proposed based on the finite field multiplicative group, which has easier construction, more flexible code-length code-rate adjustment and lower encoding/decoding complexity. Moreover, a regular QC-LDPC(5334,4962) code is constructed. The simulation results show that the constructed QC-LDPC(5334,4962) code can gain better error correction performance under the condition of the additive white Gaussian noise (AWGN) channel with iterative decoding sum-product algorithm (SPA). At the bit error rate (BER) of 10-6, the net coding gain (NCG) of the constructed QC-LDPC(5334,4962) code is 1.8 dB, 0.9 dB and 0.2 dB more than that of the classic RS(255,239) code in ITU-T G.975, the LDPC(32640,30592) code in ITU-T G.975.1 and the SCG-LDPC(3969,3720) code constructed by the random method, respectively. So it is more suitable for optical communication systems.
Classical diffusion: theory and simulation codes
International Nuclear Information System (INIS)
Grad, H.; Hu, P.N.
1978-03-01
A survey is given of the development of classical diffusion theory which arose from the observation of Grad and Hogan that the Pfirsch-Schluter and Neoclassical theories are very special and frequently inapplicable because they require that plasma mass flow be treated as transport rather than as a state variable of the plasma. The subsequent theory, efficient numerical algorithms, and results of various operating codes are described
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.)
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.
The status of simulation codes for extraction process using mixer-settler
Energy Technology Data Exchange (ETDEWEB)
Byeon, Kee Hoh; Lee, Eil Hee; Kwon, Seong Gil; Kim, Kwang Wook; Yang, Han Beom; Chung, Dong Yong; Lim, Jae Kwan; Shin, Hyun Kyoo; Kim, Soo Ho
1999-10-01
We have studied and analyzed the mixer-settler simulation codes such as three kinds of SEPHIS series, PUBG, and EXTRA.M, which is the most recently developed code. All of these are sufficiently satisfactory codes in the fields of process/device modeling, but it is necessary to formulate the accurate distribution data and chemical reaction mechanism for the aspect of accuracy and reliability. In the aspect of application to be the group separation process, the mixer-settler model of these codes have no problems, but the accumulation and formulation of partitioning and reaction equilibrium data of chemical elements used in group separation process is very important. (author)
Scalable Simulation of Electromagnetic Hybrid Codes
International Nuclear Information System (INIS)
Perumalla, Kalyan S.; Fujimoto, Richard; Karimabadi, Dr. Homa
2006-01-01
New discrete-event formulations of physics simulation models are emerging that can outperform models based on traditional time-stepped techniques. Detailed simulation of the Earth's magnetosphere, for example, requires execution of sub-models that are at widely differing timescales. In contrast to time-stepped simulation which requires tightly coupled updates to entire system state at regular time intervals, the new discrete event simulation (DES) approaches help evolve the states of sub-models on relatively independent timescales. However, parallel execution of DES-based models raises challenges with respect to their scalability and performance. One of the key challenges is to improve the computation granularity to offset synchronization and communication overheads within and across processors. Our previous work was limited in scalability and runtime performance due to the parallelization challenges. Here we report on optimizations we performed on DES-based plasma simulation models to improve parallel performance. The net result is the capability to simulate hybrid particle-in-cell (PIC) models with over 2 billion ion particles using 512 processors on supercomputing platforms
Modifications in the AUTOMESH and other POISSON Group Codes
International Nuclear Information System (INIS)
Gupta, R.C.
1986-01-01
Improvements in the POISSON Group Codes are discussed. These improvements allow one to compute magnetic field to an accuracy of a few parts in 100,000 in quite complicated geometries with a reduced requirement on computational time and computer memory. This can be accomplished mainly by making the mesh dense at some places and sparse at other places. AUTOMESH has been modified so that one can use variable mesh size conveniently and efficiently at a number of places. We will present an example to illustrate these techniques. Several other improvements in the codes AUTOMESH, LATTICE and POISSON will also be discussed
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
Towards a realistic plasma simulation code
International Nuclear Information System (INIS)
Anderson, D.V.
1991-06-01
Several new developments in the technology of simulating plasmas, both in particle and fluid models, now allow a stage of synthesis in which many of these advances can be combined into one simulation model. Accuracy and efficiency are the criteria to be satisfied in this quest. We want to build on the following research: 1. the development of the δf method of Barnes. 2. The moving node Galerkin model of Glasser, Miller and Carlson. 3. Particle moving schemes on unstructured grids by Ambrosiano and Bradon. 4. Particle simulations using sorted particles Anderson and Shumaker. Rather than being competing developments,these presumably can be combined into one computational model. We begin by summarizing the physics model for the plasma. The Vlasov equation can be solved as an initial value problem by integrating the plasma distribution function forward in time. 5 refs
Code development for nuclear reactor simulation
International Nuclear Information System (INIS)
Chauliac, C.; Verwaerde, D.; Pavageau, O.
2006-01-01
Full text of publication follows: Since several years, CEA, EDF and FANP have developed several numerical codes which are currently used for nuclear industry applications and will be remain in use for the coming years. Complementary to this set of codes and in order to better meet the present and future needs, a new system is being developed through a joint venture between CEA, EDF and FANP, with a ten year prospect and strong intermediate milestones. The focus is put on a multi-scale and multi-physics approach enabling to take into account phenomena from microscopic to macroscopic scale, and to describe interactions between various physical fields such as neutronics (DESCARTES), thermal-hydraulics (NEPTUNE) and fuel behaviour (PLEIADES). This approach is based on a more rational design of the softwares and uses a common integration platform providing pre-processing, supervision of computation and post-processing. This paper will describe the overall system under development and present the first results obtained. (authors)
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.
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)
Simulations with COREDIV Code of DEMO Discharges
Energy Technology Data Exchange (ETDEWEB)
Zagorski, R.; Stankiewicz, R.; Ivanova-Stanik, I., E-mail: roman.zagorski@ipplm.pl [Institute of Plasma Physics and Laser Microfusion, Warsaw (Poland)
2012-09-15
Full text: The reduction of divertor target power load due to radiation of sputtered and externally seeded impurities in fusion reactor is investigated in this paper. The approach is based on integrated numerical modelling of DEMO discharges using the COREDIV code, which self-consistently solves 1D radial transport equations of plasma and impurities in the core region and 2D multifluid transport in the SOL. The model is fully self-consistent with respect to both the effects of impurities on the alpha-power level and the interaction between seeded and intrinsic impurities. The code has been already successfully benchmarked with the data from present day experiments (JET, ADEX). Calculations have been performed for inductive DEMO scenario and DEMO Steady-State configuration with tungsten walls and Ar seeding. In case of DEMO Steady-State scenario strong increase of Z{sub eff} and significant reduction of the alpha power are observed with the increase of Ar influx which is caused by the decrease of fuel ions density due to the dilution effect. It leads to the reduction of the target plate heat loads but surprisingly the radiation level remains almost constant with the increased seeding which is the result of the interplay between the energy losses and tungsten source due to sputtering processes. It has been found that the W radiation is the dominant energy loss mechanism and it accounts for 90% of all radiation losses. In case of pulsed DEMO scenario, it appears that the helium accumulation might be a serious problem. Even without seeding the resulting Z{sub eff} is very large (> 2.6) and consequently only relatively weak seeding can be applied for pulsed scenario. It is found that helium accumulation depends strongly on the transport model used for helium, if the helium diffusion is increased than the accumulation effect is mitigated. (author)
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)
Monte Carlo simulation on nuclear energy study. Annual report of Nuclear Code Evaluation Committee
International Nuclear Information System (INIS)
Sakurai, Kiyoshi; Yamamoto, Toshihiro
1999-03-01
In this report, research results discussed in 1998 fiscal year at Nuclear Code Evaluation Special Committee of Nuclear Code Committee were summarised. Present status of Monte Carlo calculation in high energy region investigated / discussed at Monte Carlo simulation working-group and automatic compilation system for MCNP cross sections developed at MCNP high temperature library compilation working-group were described. The 6 papers are indexed individually. (J.P.N.)
International Nuclear Information System (INIS)
Bartlett, D.V.
1983-06-01
The codes which have been developed for the analysis of electron cyclotron emission measurements in JET are described. Their principal function is to interpret the spectra measured by the diagnostic so as to give the spatial distribution of the electron temperature in the poloidal cross-section. Various systematic effects in the data are corrected using look-up tables generated by an elaborate simulation code. The part of this code responsible for the accurate calculation of single-pass emission and refraction has been written at CNR-Milan and is described in a separate report. The present report is divided into two parts. This first part describes the methods used for the simulation and interpretation of spectra, the physical/mathematical basis of the codes written at CEA-Fontenay and presents some illustrative results
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.
Reference manual for the POISSON/SUPERFISH Group of Codes
Energy Technology Data Exchange (ETDEWEB)
1987-01-01
The POISSON/SUPERFISH Group codes were set up to solve two separate problems: the design of magnets and the design of rf cavities in a two-dimensional geometry. The first stage of either problem is to describe the layout of the magnet or cavity in a way that can be used as input to solve the generalized Poisson equation for magnets or the Helmholtz equations for cavities. The computer codes require that the problems be discretized by replacing the differentials (dx,dy) by finite differences ({delta}X,{delta}Y). Instead of defining the function everywhere in a plane, the function is defined only at a finite number of points on a mesh in the plane.
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.
DART: a simulation code for charged particle beams
International Nuclear Information System (INIS)
White, R.C.; Barr, W.L.; Moir, R.W.
1988-01-01
This paper presents a recently modified verion of the 2-D DART code designed to simulate the behavior of a beam of charged particles whose paths are affected 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 technique includes space charge, secondary electron effects, and neutral gas ionization. Calculations of electrode placement and energy conversion efficiency are described. Basic operation procedures are given including sample input files and output. 7 refs., 18 figs
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
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)
Simulation of the turbine discharge transient with the code Trace
International Nuclear Information System (INIS)
Mejia S, D. M.; Filio L, C.
2014-10-01
In this paper the results of the simulation of the turbine discharge transient are shown, occurred in Unit 1 of nuclear power plant of Laguna Verde (NPP-L V), carried out with the model of this unit for the best estimate code Trace. The results obtained by the code Trace are compared with those obtained from the Process Information Integral System (PIIS) of the NPP-L V. The reactor pressure, level behavior in the down-comer, steam flow and flow rate through the recirculation circuits are compared. The results of the simulation for the operation power of 2027 MWt, show concordance with the system PIIS. (Author)
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)
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
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.)
Simulations of linear and Hamming codes using SageMath
Timur, Tahta D.; Adzkiya, Dieky; Soleha
2018-03-01
Digital data transmission over a noisy channel could distort the message being transmitted. The goal of coding theory is to ensure data integrity, that is, to find out if and where this noise has distorted the message and what the original message was. Data transmission consists of three stages: encoding, transmission, and decoding. Linear and Hamming codes are codes that we discussed in this work, where encoding algorithms are parity check and generator matrix, and decoding algorithms are nearest neighbor and syndrome. We aim to show that we can simulate these processes using SageMath software, which has built-in class of coding theory in general and linear codes in particular. First we consider the message as a binary vector of size k. This message then will be encoded to a vector with size n using given algorithms. And then a noisy channel with particular value of error probability will be created where the transmission will took place. The last task would be decoding, which will correct and revert the received message back to the original message whenever possible, that is, if the number of error occurred is smaller or equal to the correcting radius of the code. In this paper we will use two types of data for simulations, namely vector and text data.
Numerical simulations of inertial confinement fusion hohlraum with LARED-integration code
International Nuclear Information System (INIS)
Li Jinghong; Li Shuanggui; Zhai Chuanlei
2011-01-01
In the target design of the Inertial Confinement Fusion (ICF) program, it is common practice to apply radiation hydrodynamics code to study the key physical processes happened in ICF process, such as hohlraum physics, radiation drive symmetry, capsule implosion physics in the radiation-drive approach of ICF. Recently, many efforts have been done to develop our 2D integrated simulation capability of laser fusion with a variety of optional physical models and numerical methods. In order to effectively integrate the existing codes and to facilitate the development of new codes, we are developing an object-oriented structured-mesh parallel code-supporting infrastructure, called JASMIN. Based on two-dimensional three-temperature hohlraum physics code LARED-H and two-dimensional multi-group radiative transfer code LARED-R, we develop a new generation two-dimensional laser fusion code under the JASMIN infrastructure, which enable us to simulate the whole process of laser fusion from the laser beams' entrance into the hohlraum to the end of implosion. In this paper, we will give a brief description of our new-generation two-dimensional laser fusion code, named LARED-Integration, especially in its physical models, and present some simulation results of holhraum. (author)
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.
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
Contribution to study and design of PWR plant simulation code
International Nuclear Information System (INIS)
Delourme, Didier.
1980-11-01
This paper presents an improvement of PICOLO, a package for PWR plants simulation. Its describes principally the integration to the code of a primary loop and pressurizer model and the corresponding control loops. Fast transients are tested on the packages and results are compared with real transients obtained on plants [fr
MCB. A continuous energy Monte Carlo burnup simulation code
International Nuclear Information System (INIS)
Cetnar, J.; Wallenius, J.; Gudowski, W.
1999-01-01
A code for integrated simulation of neutrinos and burnup based upon continuous energy Monte Carlo techniques and transmutation trajectory analysis has been developed. Being especially well suited for studies of nuclear waste transmutation systems, the code is an extension of the well validated MCNP transport program of Los Alamos National Laboratory. Among the advantages of the code (named MCB) is a fully integrated data treatment combined with a time-stepping routine that automatically corrects for burnup dependent changes in reaction rates, neutron multiplication, material composition and self-shielding. Fission product yields are treated as continuous functions of incident neutron energy, using a non-equilibrium thermodynamical model of the fission process. In the present paper a brief description of the code and applied methods are given. (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.
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.
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
UNIPIC code for simulations of high power microwave devices
International Nuclear Information System (INIS)
Wang Jianguo; Zhang Dianhui; Wang Yue; Qiao Hailiang; Li Xiaoze; Liu Chunliang; Li Yongdong; Wang Hongguang
2009-01-01
In this paper, UNIPIC code, a new member in the family of fully electromagnetic particle-in-cell (PIC) codes for simulations of high power microwave (HPM) generation, is introduced. In the UNIPIC code, the electromagnetic fields are updated using the second-order, finite-difference time-domain (FDTD) method, and the particles are moved using the relativistic Newton-Lorentz force equation. The convolutional perfectly matched layer method is used to truncate the open boundaries of HPM devices. To model curved surfaces and avoid the time step reduction in the conformal-path FDTD method, CP weakly conditional-stable FDTD (WCS FDTD) method which combines the WCS FDTD and CP-FDTD methods, is implemented. UNIPIC is two-and-a-half dimensional, is written in the object-oriented C++ language, and can be run on a variety of platforms including WINDOWS, LINUX, and UNIX. Users can use the graphical user's interface to create the geometric structures of the simulated HPM devices, or input the old structures created before. Numerical experiments on some typical HPM devices by using the UNIPIC code are given. The results are compared to those obtained from some well-known PIC codes, which agree well with each other.
UNIPIC code for simulations of high power microwave devices
Wang, Jianguo; Zhang, Dianhui; Liu, Chunliang; Li, Yongdong; Wang, Yue; Wang, Hongguang; Qiao, Hailiang; Li, Xiaoze
2009-03-01
In this paper, UNIPIC code, a new member in the family of fully electromagnetic particle-in-cell (PIC) codes for simulations of high power microwave (HPM) generation, is introduced. In the UNIPIC code, the electromagnetic fields are updated using the second-order, finite-difference time-domain (FDTD) method, and the particles are moved using the relativistic Newton-Lorentz force equation. The convolutional perfectly matched layer method is used to truncate the open boundaries of HPM devices. To model curved surfaces and avoid the time step reduction in the conformal-path FDTD method, CP weakly conditional-stable FDTD (WCS FDTD) method which combines the WCS FDTD and CP-FDTD methods, is implemented. UNIPIC is two-and-a-half dimensional, is written in the object-oriented C++ language, and can be run on a variety of platforms including WINDOWS, LINUX, and UNIX. Users can use the graphical user's interface to create the geometric structures of the simulated HPM devices, or input the old structures created before. Numerical experiments on some typical HPM devices by using the UNIPIC code are given. The results are compared to those obtained from some well-known PIC codes, which agree well with each other.
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
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.
Simulation of water hammer phenomena using the system code ATHLET
Energy Technology Data Exchange (ETDEWEB)
Bratfisch, Christoph; Koch, Marco K. [Bochum Univ. (Germany). Reactor Simulation and Safety Group
2017-07-15
Water Hammer Phenomena can endanger the integrity of structures leading to a possible failure of pipes in nuclear power plants as well as in many industrial applications. These phenomena can arise in nuclear power plants in the course of transients and accidents induced by the start-up of auxiliary feed water systems or emergency core cooling systems in combination with rapid acting valves and pumps. To contribute to further development and validation of the code ATHLET (Analysis of Thermalhydraulics of Leaks and Transients), an experiment performed in the test facility Pilot Plant Pipework (PPP) at Fraunhofer UMSICHT is simulated using the code version ATHLET 3.0A.
Simulation of water hammer phenomena using the system code ATHLET
International Nuclear Information System (INIS)
Bratfisch, Christoph; Koch, Marco K.
2017-01-01
Water Hammer Phenomena can endanger the integrity of structures leading to a possible failure of pipes in nuclear power plants as well as in many industrial applications. These phenomena can arise in nuclear power plants in the course of transients and accidents induced by the start-up of auxiliary feed water systems or emergency core cooling systems in combination with rapid acting valves and pumps. To contribute to further development and validation of the code ATHLET (Analysis of Thermalhydraulics of Leaks and Transients), an experiment performed in the test facility Pilot Plant Pipework (PPP) at Fraunhofer UMSICHT is simulated using the code version ATHLET 3.0A.
[Code of ethics for nurses and territory hospital group].
Danan, Jane-Laure; Giraud-Rochon, François
2017-09-01
The publication of the decree relating to the code of ethics for nurses means that the State is producing a text for all nursing professionals, whatever their sector or their mode of practice. However, faced with the standardisation of nursing procedures, the production of a new standard by a government is not a neutral issue. On the one hand, it could constitute a reinforcement of the professional credibility of this corporation; on the other this text becomes enforceable on all nurses and employers. Within a territory hospital group, this reflection must form part of nursing and managerial practices and the relationships with the hospital administration. Copyright © 2017 Elsevier Masson SAS. All rights reserved.
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.
Development of dynamic simulation code for fuel cycle fusion reactor
Energy Technology Data Exchange (ETDEWEB)
Aoki, Isao; Seki, Yasushi [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki (Japan); Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan
1999-02-01
A dynamic simulation code for fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during 2 days pulse operation cycles. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the fuel burn and the function of exhaust, purification, and supply. The processing constants of subsystem for steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using this code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)
Steam explosion simulation code JASMINE v.3 user's guide
International Nuclear Information System (INIS)
Moriyama, Kiyofumi; Maruyama, Yu; Nakamura, Hideo
2008-07-01
A steam explosion occurs when hot liquid contacts with cold volatile liquid. In this phenomenon, fine fragmentation of the hot liquid causes extremely rapid heat transfer from the hot liquid to the cold volatile liquid, and explosive vaporization, bringing shock waves and destructive forces. The steam explosion due to the contact of the molten core material and coolant water during severe accidents of light water reactors has been regarded as a potential threat to the integrity of the containment vessel. We developed a mechanistic steam explosion simulation code, JASMINE, that is applicable to plant scale assessment of the steam explosion loads. This document, as a manual for users of JASMINE code, describes the models, numerical solution methods, and also some verification and example calculations, as well as practical instructions for input preparation and usage of the code. (author)
openQ*D simulation code for QCD+QED
Campos, Isabel; Fritzsch, Patrick; Hansen, Martin; Krstić Marinković, Marina; Patella, Agostino; Ramos, Alberto; Tantalo, Nazario
2018-03-01
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 action, the multilevel integrator, the 4th order OMF integrator, the SSE/AVX intrinsics, etc. The photon field is treated as fully dynamical and C* boundary conditions can be chosen in the spatial directions. We discuss the main features of openQ*D, and we show basic test results and performance analysis. An alpha version of this code is publicly available and can be downloaded from http://rcstar.web.cern.ch/.
Simulation of vanadium-48 production using MCNPX code
Directory of Open Access Journals (Sweden)
Sadeghi Mahdi
2012-01-01
Full Text Available Vanadium-48 was produced through the irradiation of the natural titanium target via the natTi(p, xn48V reaction. The titanium target was irradiated at 1 mA current and by a 21 MeV proton beam for 4 hours. In this paper, the activity of 48V, 43Sc, and 46Sc radionuclides and the efficacy of the 47Ti(p, g, 48Ti(p, n, and 49Ti(p, 2n channel reactions to form 48V radionuclide were determined using MCNPX code. Furthermore, the experimental activity of 48V was compared with the estimated value for the thick target yield produced in the irradiation time according to MCNPX code. Good agreement between production yield of the 48V and the simulation yield was observed. In conclusion, MCNPX code can be used for the estimation of the production yield.
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.
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
Simulation of linac operation using the tracking code L
International Nuclear Information System (INIS)
Drevlak, M.; Timm, M.; Weiland, T.
1996-01-01
In linear accelerators, misalignments of the machine elements can cause considerable emittance growth due to wake fields, dispersion and other effects. Hence, tight limits are imposed on machine tolerances, design parameters and methods of machine operation. In order to simulate the beam dynamics in linacs, the tracking code L has been developed. Including both single- and multi-bunch effects, the behaviour of the beam in the machine can be simulated and adjustments on parameters of the machine elements up to complete correction techniques and operation procedures can be applied. Utilization of the program is facilitated by a graphical user interface. In this paper we will give an overview over the capabilities of this code and demonstrate its efficiency at attacking the problems associated with large linear accelerators. (author)
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.
Enhanced verification test suite for physics simulation codes
Energy Technology Data Exchange (ETDEWEB)
Kamm, James R.; Brock, Jerry S.; Brandon, Scott T.; Cotrell, David L.; Johnson, Bryan; Knupp, Patrick; Rider, William J.; Trucano, Timothy G.; Weirs, V. Gregory
2008-09-01
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.
Nuclear densimeter of soil simulated in MCNP-4C code
International Nuclear Information System (INIS)
Braga, Mario R.M.S.S.; Penna, Rodrigo; Vasconcelos, Danilo C.; Pereira, Claubia; Guerra, Bruno T.; Silva, Clemente J.G.C.
2009-01-01
The Monte Carlo code (MCNPX) was used to simulate a nuclear densimeter for measuring soil density. An Americium source (E = 60 keV) and a NaI (Tl) detector were placed on soil surface. Results from MCNP shown that scattered photon fluxes may be used to determining soil density. Linear regressions between scattered photons fluxes and soil density were calculated and shown correlation coefficients near unity. (author)
Use of advanced simulations in fuel performance codes
International Nuclear Information System (INIS)
Van Uffelen, P.
2015-01-01
The simulation of the cylindrical fuel rod behaviour in a reactor or a storage pool for spent fuel requires a fuel performance code. Such tool solves the equations for the heat transfer, the stresses and strains in fuel and cladding, the evolution of several isotopes and the behaviour of various fission products in the fuel rod. The main equations along with their limitations are briefly described. The current approaches adopted for overcoming these limitations and the perspectives are also outlined. (author)
Low-temperature plasma simulations with the LSP PIC code
Carlsson, Johan; Khrabrov, Alex; Kaganovich, Igor; Keating, David; Selezneva, Svetlana; Sommerer, Timothy
2014-10-01
The LSP (Large-Scale Plasma) PIC-MCC code has been used to simulate several low-temperature plasma configurations, including a gas switch for high-power AC/DC conversion, a glow discharge and a Hall thruster. Simulation results will be presented with an emphasis on code comparison and validation against experiment. High-voltage, direct-current (HVDC) power transmission is becoming more common as it can reduce construction costs and power losses. Solid-state power-electronics devices are presently used, but it has been proposed that gas switches could become a compact, less costly, alternative. A gas-switch conversion device would be based on a glow discharge, with a magnetically insulated cold cathode. Its operation is similar to that of a sputtering magnetron, but with much higher pressure (0.1 to 0.3 Torr) in order to achieve high current density. We have performed 1D (axial) and 2D (axial/radial) simulations of such a gas switch using LSP. The 1D results were compared with results from the EDIPIC code. To test and compare the collision models used by the LSP and EDIPIC codes in more detail, a validation exercise was performed for the cathode fall of a glow discharge. We will also present some 2D (radial/azimuthal) LSP simulations of a Hall thruster. The information, data, or work presented herein was funded in part by the Advanced Research Projects Agency-Energy (ARPA-E), U.S. Department of Energy, under Award Number DE-AR0000298.
Electron cloud effects: codes and simulations at KEK
International Nuclear Information System (INIS)
Ohmi, K
2013-01-01
Electron cloud effects had been studied at KEK-Photon Factory since 1995. e-p instability had been studied in proton rings since 1965 in BINP, ISR and PSR. Study of electron cloud effects with the present style, which was based on numerical simulations, started at 1995 in positron storage rings. The instability observed in KEKPF gave a strong impact to B factories, KEKB and PEPII, which were final stage of their design in those days. History of cure for electron cloud instability overlapped the progress of luminosity performance in KEKB. The studies on electron cloud codes and simulations in KEK are presented. (author)
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)
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.
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
LOLA SYSTEM: A code block for nodal PWR simulation. Part. I - Simula-3 Code
Energy Technology Data Exchange (ETDEWEB)
Aragones, J M; Ahnert, C; Gomez Santamaria, J; Rodriguez Olabarria, I
1985-07-01
Description of the theory and users manual of the SIMULA-3 code, which is part of the core calculation system by nodal theory in one group, called LOLA SYSTEM. SIMULA-3 is the main module of the system, it uses a modified nodal theory, with interface leakages equivalent to the diffusion theory. (Author) 4 refs.
LOLA SYSTEM: A code block for nodal PWR simulation. Part. I - Simula-3 Code
International Nuclear Information System (INIS)
Aragones, J. M.; Ahnert, C.; Gomez Santamaria, J.; Rodriguez Olabarria, I.
1985-01-01
Description of the theory and users manual of the SIMULA-3 code, which is part of the core calculation system by nodal theory in one group, called LOLA SYSTEM. SIMULA-3 is the main module of the system, it uses a modified nodal theory, with interface leakages equivalent to the diffusion theory. (Author) 4 refs
Control of complex physically simulated robot groups
Brogan, David C.
2001-10-01
Actuated systems such as robots take many forms and sizes but each requires solving the difficult task of utilizing available control inputs to accomplish desired system performance. Coordinated groups of robots provide the opportunity to accomplish more complex tasks, to adapt to changing environmental conditions, and to survive individual failures. Similarly, groups of simulated robots, represented as graphical characters, can test the design of experimental scenarios and provide autonomous interactive counterparts for video games. The complexity of writing control algorithms for these groups currently hinders their use. A combination of biologically inspired heuristics, search strategies, and optimization techniques serve to reduce the complexity of controlling these real and simulated characters and to provide computationally feasible solutions.
On the use of the Serpent Monte Carlo code for few-group cross section generation
International Nuclear Information System (INIS)
Fridman, E.; Leppaenen, J.
2011-01-01
Research highlights: → B1 methodology was used for generation of leakage-corrected few-group cross sections in the Serpent Monte-Carlo code. → Few-group constants generated by Serpent were compared with those calculated by Helios deterministic lattice transport code. → 3D analysis of a PWR core was performed by a nodal diffusion code DYN3D employing two-group cross section sets generated by Serpent and Helios. → An excellent agreement in the results of 3D core calculations obtained with Helios and Serpent generated cross-section libraries was observed. - Abstract: Serpent is a recently developed 3D continuous-energy Monte Carlo (MC) reactor physics burnup calculation code. Serpent is specifically designed for lattice physics applications including generation of homogenized few-group constants for full-core core simulators. Currently in Serpent, the few-group constants are obtained from the infinite-lattice calculations with zero neutron current at the outer boundary. In this study, in order to account for the non-physical infinite-lattice approximation, B1 methodology, routinely used by deterministic lattice transport codes, was considered for generation of leakage-corrected few-group cross sections in the Serpent code. A preliminary assessment of the applicability of the B1 methodology for generation of few-group constants in the Serpent code was carried out according to the following steps. Initially, the two-group constants generated by Serpent were compared with those calculated by Helios deterministic lattice transport code. Then, a 3D analysis of a Pressurized Water Reactor (PWR) core was performed by the nodal diffusion code DYN3D employing two-group cross section sets generated by Serpent and Helios. At this stage thermal-hydraulic (T-H) feedback was neglected. The DYN3D results were compared with those obtained from the 3D full core Serpent MC calculations. Finally, the full core DYN3D calculations were repeated taking into account T-H feedback and
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
Development of multi-group spectral code TVS-M
International Nuclear Information System (INIS)
Lazarenko, A. P.; Pryanichnikov, A. V.; Kalugin, M. A.; Gurevich, M. I.
2011-01-01
This paper is dedicated to the latest version of TVS-M code - TVS-M 2007, which allows the neutron flux distribution inside fuel assemblies to be calculated without using the diffusion approximation. The new spatial calculation module PERST introduced in TBS-M code is based on the first collisions probability method and allows the scattering anisotropy to be accounted for. This paper presents some preliminary results calculated with the use of the new version of TVS-M code. (Authors)
Image Quality Assessment via Quality-aware Group Sparse Coding
Directory of Open Access Journals (Sweden)
Minglei Tong
2014-12-01
Full Text Available Image quality assessment has been attracting growing attention at an accelerated pace over the past decade, in the fields of image processing, vision and machine learning. In particular, general purpose blind image quality assessment is technically challenging and lots of state-of-the-art approaches have been developed to solve this problem, most under the supervised learning framework where the human scored samples are needed for training a regression model. In this paper, we propose an unsupervised learning approach that work without the human label. In the off-line stage, our method trains a dictionary covering different levels of image quality patch atoms across the training samples without knowing the human score, where each atom is associated with a quality score induced from the reference image; at the on-line stage, given each image patch, our method performs group sparse coding to encode the sample, such that the sample quality can be estimated from the few labeled atoms whose encoding coefficients are nonzero. Experimental results on the public dataset show the promising performance of our approach and future research direction is also discussed.
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
Large-eddy simulation of stratified atmospheric flows with the CFD code Code-Saturne
International Nuclear Information System (INIS)
Dall'Ozzo, Cedric
2013-01-01
Large-eddy simulation (LES) of the physical processes in the atmospheric boundary layer (ABL) remains a complex subject. LES models have difficulties to capture the evolution of the turbulence in different conditions of stratification. Consequently, LES of the whole diurnal cycle of the ABL including convective situations in daytime and stable situations in the nighttime is seldom documented. The simulation of the stable atmospheric boundary layer which is characterized by small eddies and by weak and sporadic turbulence is especially difficult. Therefore The LES ability to well reproduce real meteorological conditions, particularly in stable situations, is studied with the CFD code developed by EDF R and D, Code-Saturne. The first study consist in validate LES on a quasi-steady state convective case with homogeneous terrain. The influence of the sub-grid-scale models (Smagorinsky model, Germano-Lilly model, Wong-Lilly model and Wall-Adapting Local Eddy-viscosity model) and the sensitivity to the parametrization method on the mean fields, flux and variances are discussed. In a second study, the diurnal cycle of the ABL during Wangara experiment is simulated. The deviation from the measurement is weak during the day, so this work is focused on the difficulties met during the night to simulate the stable atmospheric boundary layer. The impact of the different sub-grid-scale models and the sensitivity to the Smagorinsky constant are been analysed. By coupling radiative forcing with LES, the consequences of infra-red and solar radiation on the nocturnal low level jet and on thermal gradient, close to the surface, are exposed. More, enhancement of the domain resolution to the turbulence intensity and the strong atmospheric stability during the Wangara experiment are analysed. Finally, a study of the numerical oscillations inherent to Code-Saturne is realized in order to decrease their effects. (author) [fr
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.)
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)
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)
Comparing DINA code simulations with TCV experimental plasma equilibrium responses
International Nuclear Information System (INIS)
Khayrutdinov, R.R.; Lister, J.B.; Lukash, V.E.; Wainwright, J.P.
2000-08-01
The DINA non-linear time dependent simulation code has been validated against an extensive set of plasma equilibrium response experiments carried out on the TCV tokamak. Limited and diverted plasmas are found to be well modelled during the plasma current flat top. In some simulations the application of the PF coil voltage stimulation pulse sufficiently changed the plasma equilibrium that the vertical position feedback control loop became unstable. This behaviour was also found in the experimental work, and cannot be reproduced using linear time-independent models. A single null diverted plasma discharge was also simulated from start-up to shut-down and the results were found to accurately reproduce their experimental equivalents. The most significant difference noted was the penetration time of the poloidal flux, leading to a delayed onset of sawtoothing in the DINA simulation. The complete set of frequency stimulation experiments used to measure the open loop tokamak plasma equilibrium response was also simulated using DINA and the results were analysed in an identical fashion to the experimental data. The frequency response of the DINA simulations agrees with the experimental results. Comparisons with linear models are also discussed to identify areas of good and only occasionally less good agreement. (author)
Focus Group Research on the Implications of Adopting the Unified English Braille Code
Wetzel, Robin; Knowlton, Marie
2006-01-01
Five focus groups explored concerns about adopting the Unified English Braille Code. The consensus was that while the proposed changes to the literary braille code would be minor, those to the mathematics braille code would be much more extensive. The participants emphasized that "any code that reduces the number of individuals who can access…
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.
Calibration of the TIME2 environmental simulation code
International Nuclear Information System (INIS)
Wilmot, R.D.; Hiscock, K.; Lloyd, J.
1991-04-01
The TARGET finite-difference groundwater modelling code has been used to reconstruct the hydrogeological environment of the area around Killingholme, Humberside, UK. Reconstructions have been made for the present day and for three periods during the past 120,000 years. Permeability development in the Chalk and the stratified nature of the current groundwater system act as boundary conditions for these reconstructions. The results from these reconstructions have been compared with values used by the environmental simulation code TIME2. With optimisation of partition coefficients within the water budget sub-model, values for recharge from TIME2 accord closely with those from this study for temperate and boreal conditions. TIME2 over-estimates recharge during tundra climate states because it does not account for permafrost. (author)
Comparison of ANL containment codes with SNR-300 simulation experiments
International Nuclear Information System (INIS)
Marchertas, A.H.; Wang, C.Y.; Fistedis, S.H.
1976-01-01
A comparison of REXCO and ICECO code predictions is made with data obtained from experiments of LMFBR excursion models. The comparisons are based on published results of tests conducted for the safety analysis of the SNR-300 fast breeder. The test configurations consist of a centrally located spherical source immersed in a pool of water which is encased in a cylindrical container. The cylinical walls of the container are prestressed by holddown bolts which span the two rigid ends. The space above the surface of the water within the container is occupied by air. Although certain aspects of the tests could not be simulated by the analytical models exactly, the comparison of results shows quite close agreement. The fact that the REXCO and ICECO codes involve different analytical formulations, their own close correspondence of results lends added credence to the value of analytical predictions
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
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
GOTHIC code simulation of thermal stratification in POOLEX facility
International Nuclear Information System (INIS)
Li, H.; Kudinov, P.
2009-07-01
Pressure suppression pool is an important element of BWR containment. It serves as a heat sink and steam condenser to prevent containment pressure buildup during loss of coolant accident or safety relief valve opening during normal operations of a BWR. Insufficient mixing in the pool, in case of low mass flow rate of steam, can cause development of thermal stratification and reduction of pressure suppression pool capacity. For reliable prediction of mixing and stratification phenomena validation of simulation tools has to be performed. Data produced in POOLEX/PPOOLEX facility at Lappeenranta University of Technology about development of thermal stratification in a large scale model of a pressure suppression pool is used for GOTHIC lumped and distributed parameter validation. Sensitivity of GOTHIC solution to different boundary conditions and grid convergence study for 2D simulations of POOLEX STB-20 experiment are performed in the present study. CFD simulation was carried out with FLUENT code in order to get additional insights into physics of stratification phenomena. In order to support development of experimental procedures for new tests in the PPOOLEX facility lumped parameter pre-test GOTHIC simulations were performed. Simulations show that drywell and wetwell pressures can be kept within safety margins during a long transient necessary for development of thermal stratification. (au)
GOTHIC code simulation of thermal stratification in POOLEX facility
Energy Technology Data Exchange (ETDEWEB)
Li, H.; Kudinov, P. (Royal Institute of Technology (KTH) (Sweden))
2009-07-15
Pressure suppression pool is an important element of BWR containment. It serves as a heat sink and steam condenser to prevent containment pressure buildup during loss of coolant accident or safety relief valve opening during normal operations of a BWR. Insufficient mixing in the pool, in case of low mass flow rate of steam, can cause development of thermal stratification and reduction of pressure suppression pool capacity. For reliable prediction of mixing and stratification phenomena validation of simulation tools has to be performed. Data produced in POOLEX/PPOOLEX facility at Lappeenranta University of Technology about development of thermal stratification in a large scale model of a pressure suppression pool is used for GOTHIC lumped and distributed parameter validation. Sensitivity of GOTHIC solution to different boundary conditions and grid convergence study for 2D simulations of POOLEX STB-20 experiment are performed in the present study. CFD simulation was carried out with FLUENT code in order to get additional insights into physics of stratification phenomena. In order to support development of experimental procedures for new tests in the PPOOLEX facility lumped parameter pre-test GOTHIC simulations were performed. Simulations show that drywell and wetwell pressures can be kept within safety margins during a long transient necessary for development of thermal stratification. (au)
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 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)
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
ATES/heat pump simulations performed with ATESSS code
Vail, L. W.
1989-01-01
Modifications to the Aquifer Thermal Energy Storage System Simulator (ATESSS) allow simulation of aquifer thermal energy storage (ATES)/heat pump systems. The heat pump algorithm requires a coefficient of performance (COP) relationship of the form: COP = COP sub base + alpha (T sub ref minus T sub base). Initial applications of the modified ATES code to synthetic building load data for two sizes of buildings in two U.S. cities showed insignificant performance advantage of a series ATES heat pump system over a conventional groundwater heat pump system. The addition of algorithms for a cooling tower and solar array improved performance slightly. Small values of alpha in the COP relationship are the principal reason for the limited improvement in system performance. Future studies at Pacific Northwest Laboratory (PNL) are planned to investigate methods to increase system performance using alternative system configurations and operations scenarios.
The proceedings of the KEK FEL simulation code workshop
International Nuclear Information System (INIS)
Kamitani, Takuya
1992-11-01
This is the record of the lectures in free electron laser simulation code workshop held in National Laboratory for High Energy Physics on March 15, 1991. As the device that can generate especially powerful and coherent light in the wide wavelength region from long wavelength like microwave to short wavelength like X-ray and gamma ray, the interest in free electron laser has heightened in Japan and foreign countries, and also the experiments have been carried out actively. Also the necessity of the quantitative theoretical calculation using the simulation has become high, and the researches have been carried out in various places. This workshop was held for the intention of offering the place for the interchange of researches, the exchange of information and discussion. 39 persons took part in the workshop, and 11 lectures were given, and it was very useful. (K.I.)
Monte Carlo simulation of medical linear accelerator using primo code
International Nuclear Information System (INIS)
Omer, Mohamed Osman Mohamed Elhasan
2014-12-01
The use of monte Carlo simulation has become very important in the medical field and especially in calculation in radiotherapy. Various Monte Carlo codes were developed simulating interactions of particles and photons with matter. One of these codes is PRIMO that performs simulation of radiation transport from the primary electron source of a linac to estimate the absorbed dose in a water phantom or computerized tomography (CT). PRIMO is based on Penelope Monte Carlo code. Measurements of 6 MV photon beam PDD and profile were done for Elekta precise linear accelerator at Radiation and Isotopes Center Khartoum using computerized Blue water phantom and CC13 Ionization Chamber. accept Software was used to control the phantom to measure and verify dose distribution. Elektalinac from the list of available linacs in PRIMO was tuned to model Elekta precise linear accelerator. Beam parameter of 6.0 MeV initial electron energy, 0.20 MeV FWHM, and 0.20 cm focal spot FWHM were used, and an error of 4% between calculated and measured curves was found. The buildup region Z max was 1.40 cm and homogenous profile in cross line and in line were acquired. A number of studies were done to verily the model usability one of them is the effect of the number of histories on accuracy of the simulation and the resulted profile for the same beam parameters. The effect was noticeable and inaccuracies in the profile were reduced by increasing the number of histories. Another study was the effect of Side-step errors on the calculated dose which was compared with the measured dose for the same setting.It was in range of 2% for 5 cm shift, but it was higher in the calculated dose because of the small difference between the tuned model and measured dose curves. Future developments include simulating asymmetrical fields, calculating the dose distribution in computerized tomographic (CT) volume, studying the effect of beam modifiers on beam profile for both electron and photon beams.(Author)
Modelling guidelines for core exit temperature simulations with system codes
Energy Technology Data Exchange (ETDEWEB)
Freixa, J., E-mail: jordi.freixa-terradas@upc.edu [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain); Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland); Martínez-Quiroga, V., E-mail: victor.martinez@nortuen.com [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain); Zerkak, O., E-mail: omar.zerkak@psi.ch [Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland); Reventós, F., E-mail: francesc.reventos@upc.edu [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain)
2015-05-15
Highlights: • Core exit temperature is used in PWRs as an indication of core heat up. • Modelling guidelines of CET response with system codes. • Modelling of heat transfer processes in the core and UP regions. - Abstract: Core exit temperature (CET) measurements play an important role in the sequence of actions under accidental conditions in pressurized water reactors (PWR). Given the difficulties in placing measurements in the core region, CET readings are used as criterion for the initiation of accident management (AM) procedures because they can indicate a core heat up scenario. However, the CET responses have some limitation in detecting inadequate core cooling and core uncovery simply because the measurement is not placed inside the core. Therefore, it is of main importance in the field of nuclear safety for PWR power plants to assess the capabilities of system codes for simulating the relation between the CET and the peak cladding temperature (PCT). The work presented in this paper intends to address this open question by making use of experimental work at integral test facilities (ITF) where experiments related to the evolution of the CET and the PCT during transient conditions have been carried out. In particular, simulations of two experiments performed at the ROSA/LSTF and PKL facilities are presented. The two experiments are part of a counterpart exercise between the OECD/NEA ROSA-2 and OECD/NEA PKL-2 projects. The simulations are used to derive guidelines in how to correctly reproduce the CET response during a core heat up scenario. Three aspects have been identified to be of main importance: (1) the need for a 3-dimensional representation of the core and Upper Plenum (UP) regions in order to model the heterogeneity of the power zones and axial areas, (2) the detailed representation of the active and passive heat structures, and (3) the use of simulated thermocouples instead of steam temperatures to represent the CET readings.
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.
Huang, Sheng; Ao, Xiang; Li, Yuan-yuan; Zhang, Rui
2016-09-01
In order to meet the needs of high-speed development of optical communication system, a construction method of quasi-cyclic low-density parity-check (QC-LDPC) codes based on multiplicative group of finite field is proposed. The Tanner graph of parity check matrix of the code constructed by this method has no cycle of length 4, and it can make sure that the obtained code can get a good distance property. Simulation results show that when the bit error rate ( BER) is 10-6, in the same simulation environment, the net coding gain ( NCG) of the proposed QC-LDPC(3 780, 3 540) code with the code rate of 93.7% in this paper is improved by 2.18 dB and 1.6 dB respectively compared with those of the RS(255, 239) code in ITU-T G.975 and the LDPC(3 2640, 3 0592) code in ITU-T G.975.1. In addition, the NCG of the proposed QC-LDPC(3 780, 3 540) code is respectively 0.2 dB and 0.4 dB higher compared with those of the SG-QC-LDPC(3 780, 3 540) code based on the two different subgroups in finite field and the AS-QC-LDPC(3 780, 3 540) code based on the two arbitrary sets of a finite field. Thus, the proposed QC-LDPC(3 780, 3 540) code in this paper can be well applied in optical communication systems.
Comparing TCV experimental VDE responses with DINA code simulations
Favez, J.-Y.; Khayrutdinov, R. R.; Lister, J. B.; Lukash, V. E.
2002-02-01
The DINA free-boundary equilibrium simulation code has been implemented for TCV, including the full TCV feedback and diagnostic systems. First results showed good agreement with control coil perturbations and correctly reproduced certain non-linear features in the experimental measurements. The latest DINA code simulations, presented in this paper, exploit discharges with different cross-sectional shapes and different vertical instability growth rates which were subjected to controlled vertical displacement events (VDEs), extending previous work with the DINA code on the DIII-D tokamak. The height of the TCV vessel allows observation of the non-linear evolution of the VDE growth rate as regions of different vertical field decay index are crossed. The vertical movement of the plasma is found to be well modelled. For most experiments, DINA reproduces the S-shape of the vertical displacement in TCV with excellent precision. This behaviour cannot be modelled using linear time-independent models because of the predominant exponential shape due to the unstable pole of any linear time-independent model. The other most common equilibrium parameters like the plasma current Ip, the elongation κ, the triangularity δ, the safety factor q, the ratio between the averaged plasma kinetic pressure and the pressure of the poloidal magnetic field at the edge of the plasma βp, and the internal self inductance li also show acceptable agreement. The evolution of the growth rate γ is estimated and compared with the evolution of the closed-loop growth rate calculated with the RZIP linear model, confirming the origin of the observed behaviour.
Comparing TCV experimental VDE responses with DINA code simulations
International Nuclear Information System (INIS)
Favez, J.Y.; Khayrutdinov, J.B.; Lister, J.B.; Lukash, V.E.
2001-10-01
The DINA free-boundary equilibrium simulation code has been implemented for TCV, including the full TCV feedback and diagnostic systems. First results showed good agreement with control coil perturbations and correctly reproduced certain non-linear features in the experimental measurements. The latest DINA code simulations, presented in this paper, exploit discharges with different cross- sectional shapes and different vertical instability growth rates which were subjected to controlled Vertical Displacement Events, extending previous work with the DINA code on the DIII-D tokamak. The height of the TCV vessel allows observation of the non- linear evolution of the VDE growth rate as regions of different vertical field decay index are crossed. The vertical movement of the plasma is found to be well modelled. For most experiments, DINA reproduces the S-shape of the vertical displacement in TCV with excellent precision. This behaviour cannot be modelled using linear time-independent models because of the predominant exponential shape due to the unstable pole of any linear time-independent model. The other most common equilibrium parameters like the plasma current Ip, the elongation K, the triangularity d, the safety factor q, the ratio between the averaged plasma kinetic pressure and the pressure of the poloidal magnetic field at the edge of the plasma bp and the internal self inductance l also show acceptable agreement. The evolution of the growth rate g is estimated and compared with the evolution of the closed loop growth rate calculated with the RZIP linear model, confirming the origin of the observed behaviour. (author)
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].
Simulation of spreading with solidification: assessment synthesis of Thema code
Energy Technology Data Exchange (ETDEWEB)
Spindler, B.; Veteau, J.M. [CEA Grenoble, Direction de l' Energie Nucleaire, Dept. de Technologie Nucleaire, Service d' Etudes Thermohydrauliques et Technologiques, 38 (France)
2004-07-01
After a presentation of the models included in THEMA code, which simulates the spreading of a fluid with solidification, the whole assessment calculations are presented. The first series concerns the comparison with analytical or numerical solutions: dam break, conduction for the heat transfer in the substrate, crust growth. The second series concerns the comparison with the CORINE isothermal tests (simulating fluid at low temperature). The third series concerns the CORINE tests with heat transfer. The fourth series concerns the tests with simulating materials at medium or high temperature (RIT, KATS). The fifth series concerns the tests with prototypical materials (COMAS, FARO, VULCANO). Finally the blind simulations of the ECOKATS tests are presented. All the calculations are performed with the same physical models (THEMA version 2.5), without any variable tuning parameter according to the test under consideration. Sensitivity studies concern the influence of the viscosity model in the solidification interval, and for the tests with prototypical materials the inlet temperature and the solid fraction. The relative difference between the calculated and measured spreading areas is generally less than 20 % except for the test with prototypical materials, for which the assessment is not easy due to the large experimental uncertainties. The level of validation of THEMA is considered as satisfactory, taking into account the required accuracy. (authors)
Large interface simulation in an averaged two-fluid code
International Nuclear Information System (INIS)
Henriques, A.
2006-01-01
Different ranges of size of interfaces and eddies are involved in multiphase flow phenomena. Classical formalisms focus on a specific range of size. This study presents a Large Interface Simulation (LIS) two-fluid compressible formalism taking into account different sizes of interfaces. As in the single-phase Large Eddy Simulation, a filtering process is used to point out Large Interface (LI) simulation and Small interface (SI) modelization. The LI surface tension force is modelled adapting the well-known CSF method. The modelling of SI transfer terms is done calling for classical closure laws of the averaged approach. To simulate accurately LI transfer terms, we develop a LI recognition algorithm based on a dimensionless criterion. The LIS model is applied in a classical averaged two-fluid code. The LI transfer terms modelling and the LI recognition are validated on analytical and experimental tests. A square base basin excited by a horizontal periodic movement is studied with the LIS model. The capability of the model is also shown on the case of the break-up of a bubble in a turbulent liquid flow. The break-up of a large bubble at a grid impact performed regime transition between two different scales of interface from LI to SI and from PI to LI. (author) [fr
Simulation of spreading with solidification: assessment synthesis of Thema code
International Nuclear Information System (INIS)
Spindler, B.; Veteau, J.M.
2004-01-01
After a presentation of the models included in THEMA code, which simulates the spreading of a fluid with solidification, the whole assessment calculations are presented. The first series concerns the comparison with analytical or numerical solutions: dam break, conduction for the heat transfer in the substrate, crust growth. The second series concerns the comparison with the CORINE isothermal tests (simulating fluid at low temperature). The third series concerns the CORINE tests with heat transfer. The fourth series concerns the tests with simulating materials at medium or high temperature (RIT, KATS). The fifth series concerns the tests with prototypical materials (COMAS, FARO, VULCANO). Finally the blind simulations of the ECOKATS tests are presented. All the calculations are performed with the same physical models (THEMA version 2.5), without any variable tuning parameter according to the test under consideration. Sensitivity studies concern the influence of the viscosity model in the solidification interval, and for the tests with prototypical materials the inlet temperature and the solid fraction. The relative difference between the calculated and measured spreading areas is generally less than 20 % except for the test with prototypical materials, for which the assessment is not easy due to the large experimental uncertainties. The level of validation of THEMA is considered as satisfactory, taking into account the required accuracy. (authors)
Development and assessment of ASTEC code for severe accident simulation
International Nuclear Information System (INIS)
Van Dorsselaere, J.P.; Pignet, S.; Seropian, C.; Montanelli, T.; Giordano, P.; Jacq, F.; Schwinges, B.
2005-01-01
Full text of publication follows: The ASTEC integral code, jointly developed by IRSN and GRS since several years for evaluation of source term during a severe accident (SA) in a Light Water Reactor, will play a central role in the SARNET network of excellence of the 6. Framework Programme (FwP) of the European Commission which started in spring 2004. It should become the reference European SA integral code in the next years. The version V1.1, released in June 2004, allows to model most of the main physical phenomena (except steam explosion) near or at the state of the art. In order to allow to study a great number of scenarios, a compromise must be found between precision of results and calculation time: one day of accident time usually takes less than one day of real time to be simulated on a PC computer. Important efforts are being made on validation by covering more than 30 reference experiments, often International Standard Problems from OECD (CORA, LOFT, PACTEL, BETA, VANAM, ACE-RTF, Phebus.FPT1...). The code is also used for the detailed interpretation of all the integral Phebus.FP experiments. Eighteen European partners performed a first independent evaluation of the code capabilities in 2000-03 within the frame of the EVITA 5. FwP project on one hand by comparison to experiments and on another hand by benchmarking with MAAP4 and MELCOR integral codes on plant applications on PWR and VVER. Their main conclusions were the needs of improvement of code robustness (especially the 2 new modules CESAR and DIVA simulating respectively circuit thermal hydraulics and core degradation) and of post-processing tools. Some improvements have already been achieved in the latest version V 1.1 on these two aspects. A new module MEDICIS devoted to Molten Core Concrete Interaction (MCCI) is implemented in this version, with a tight coupling to the containment thermal hydraulics module CPA. The paper presents a detailed analysis of a TMLB sequence on a French 900 MWe PWR, from
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
Simulation of TROI steam explosion behaviour using the COMETA code
International Nuclear Information System (INIS)
Arun Kumar Nayak; Hyun Sun Park; Bal Raj Sehgal; Alessandro Annunziato
2005-01-01
Full text of publication follows: During a severe accident in a nuclear reactor, the core can melt and the molten corium while interacting with water may cause an energetic fuel coolant interaction which is known as steam explosion. Such phenomena can occur inside the reactor vessel during flooding of a degraded core or when molten corium falls into the lower head filled with water. Similar phenomena may occur outside the reactor vessel when molten corium is ejected into a flooded reactor cavity or into the flooded containment after the vessel failure. The interaction of molten corium with water is one of the most complex thermal hydraulic and chemical phenomena. Recently in the TROI test series carried out at KAERI (Korean Atomic Energy Research Institute) in Korea, steam explosions were observed. In those tests, the UO 2 /ZrO 2 compositions were close to that of prototypic case. In this paper, we have numerically simulated the melt coolant interaction of TROI tests using the computer code, COMETA (Core MElt Thermalhydraulic Analysis) developed by JRC (Joint Research Center), at Ispra in Italy. The COMETA code was primarily developed to analyse, with sufficient detail, both the thermal-hydraulics and the fuel fragmentation phenomena during the melt quenching tests as conducted in the FARO facility. The code solves the conservation equations of mass, momentum and energy for the fluid using a conventional two-fluid model. Fuel fragmentation model considers the molten jet, its break up in drops and accumulation as fused-debris on the bottom. An explicit coupling between the thermal hydraulics and fuel fragmentation for the energy transfer is considered. The code has been extensively validated in the past for melt quenching in a series of experiments in the FARO facility. In this work, we first simulated the pre-mix and triggering phases of the TROI-13 tests for which the test data were available. The melt jet trajectory, void fraction and pressure profile were
International Nuclear Information System (INIS)
Murata, Isao; Mori, Takamasa; Nakagawa, Masayuki; Shirai, Hiroshi.
1996-03-01
High Temperature Gas-cooled Reactors (HTGRs) employ spherical fuels named coated fuel particles (CFPs) consisting of a microsphere of low enriched UO 2 with coating layers in order to prevent FP release. There exist many spherical fuels distributed randomly in the cores. Therefore, the nuclear design of HTGRs is generally performed on the basis of the multigroup approximation using a diffusion code, S N transport code or group-wise Monte Carlo code. This report summarizes a Monte Carlo hard sphere packing simulation code to simulate the packing of equal hard spheres and evaluate the necessary probability distribution of them, which is used for the application of the new Monte Carlo calculation method developed to treat randomly distributed spherical fuels with the continuous energy Monte Carlo method. By using this code, obtained are the various statistical values, namely Radial Distribution Function (RDF), Nearest Neighbor Distribution (NND), 2-dimensional RDF and so on, for random packing as well as ordered close packing of FCC and BCC. (author)
A multi-group neutron noise simulator for fast reactors
International Nuclear Information System (INIS)
Tran, Hoai Nam; Zylbersztejn, Florian; Demazière, Christophe; Jammes, Christian; Filliatre, Philippe
2013-01-01
Highlights: • The development of a neutron noise simulator for fast reactors. • The noise equation is solved fully in a frequency-domain. • A good agreement with ERANOS on the static calculations. • Noise calculations induced by a localized perturbation of absorption cross section. - Abstract: A neutron noise simulator has been developed for fast reactors based on diffusion theory with multi-energy groups and several groups of delayed neutron precursors. The tool is expected to be applicable for core monitoring of fast reactors and also for other reactor types with hexagonal fuel assemblies. The noise sources are modeled through small stationary fluctuations of macroscopic cross sections, and the induced first order noise is solved fully in the frequency domain. Numerical algorithms are implemented for solving both the static and noise equations using finite differences for spatial discretization, where a hexagonal assembly is radially divided into finer triangular meshes. A coarse mesh finite difference (CMFD) acceleration has been used for accelerating the convergence of both the static and noise calculations. Numerical calculations have been performed for the ESFR core with 33 energy groups and 8 groups of delayed neutron precursors using the cross section data generated by the ERANOS code. The results of the static state have been compared with those obtained using ERANOS. The results show an adequate agreement between the two calculations. Noise calculations for the ESFR core have also been performed and demonstrated with an assumption of the perturbation of the absorption cross section located at the central fuel ring
Pellet injection and plasma behavior simulation code PEPSI
International Nuclear Information System (INIS)
Takase, Haruhiko; Tobita, Kenji; Nishio, Satoshi
2003-08-01
Fueling is one of the major issues on design of nuclear fusion reactor and the injection of solid hydrogen pellet to the core plasma is a useful method. On the design of a nuclear fusion reactor, it is necessary to determine requirements on the pellet size, the number of pellets, the injection speed and the injection cycle. PEllet injection and Plasma behavior SImulation code PEPSI has been developed to assess these parameters. PEPSI has two special features: 1) Adopting two numerical pellet models, Parks model and Strauss model, 2) Calculating fusion power and other plasma parameters in combination with a time-dependent one-dimensional transport model. This report describes the numerical models, numerical scheme, sequence of calculation, list of subroutines, list of variables and an example of calculation. (author)
Parallelization of a numerical simulation code for isotropic turbulence
International Nuclear Information System (INIS)
Sato, Shigeru; Yokokawa, Mitsuo; Watanabe, Tadashi; Kaburaki, Hideo.
1996-03-01
A parallel pseudospectral code which solves the three-dimensional Navier-Stokes equation by direct numerical simulation is developed and execution time, parallelization efficiency, load balance and scalability are evaluated. A vector parallel supercomputer, Fujitsu VPP500 with up to 16 processors is used for this calculation for Fourier modes up to 256x256x256 using 16 processors. Good scalability for number of processors is achieved when number of Fourier mode is fixed. For small Fourier modes, calculation time of the program is proportional to NlogN which is ideal complexity of calculation for 3D-FFT on vector parallel processors. It is found that the calculation performance decreases as the increase of the Fourier modes. (author)
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
The MCUCN simulation code for ultracold neutron physics
Zsigmond, G.
2018-02-01
Ultracold neutrons (UCN) have very low kinetic energies 0-300 neV, thereby can be stored in specific material or magnetic confinements for many hundreds of seconds. This makes them a very useful tool in probing fundamental symmetries of nature (for instance charge-parity violation by neutron electric dipole moment experiments) and contributing important parameters for the Big Bang nucleosynthesis (neutron lifetime measurements). Improved precision experiments are in construction at new and planned UCN sources around the world. MC simulations play an important role in the optimization of such systems with a large number of parameters, but also in the estimation of systematic effects, in benchmarking of analysis codes, or as part of the analysis. The MCUCN code written at PSI has been extensively used for the optimization of the UCN source optics and in the optimization and analysis of (test) experiments within the nEDM project based at PSI. In this paper we present the main features of MCUCN and interesting benchmark and application examples.
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.
Energy Technology Data Exchange (ETDEWEB)
Aoki, Isao; Seki, Yasushi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan
1997-11-01
A dynamic simulation code for the fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during a single pulse operation. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the function of fuel burn, exhaust, purification, and supply. The processing constants of subsystem for the steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using the code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)
Monte Carlo simulation in UWB1 depletion code
International Nuclear Information System (INIS)
Lovecky, M.; Prehradny, J.; Jirickova, J.; Skoda, R.
2015-01-01
U W B 1 depletion code is being developed as a fast computational tool for the study of burnable absorbers in the University of West Bohemia in Pilsen, Czech Republic. In order to achieve higher precision, the newly developed code was extended by adding a Monte Carlo solver. Research of fuel depletion aims at development and introduction of advanced types of burnable absorbers in nuclear fuel. Burnable absorbers (BA) allow the compensation of the initial reactivity excess of nuclear fuel and result in an increase of fuel cycles lengths with higher enriched fuels. The paper describes the depletion calculations of VVER nuclear fuel doped with rare earth oxides as burnable absorber based on performed depletion calculations, rare earth oxides are divided into two equally numerous groups, suitable burnable absorbers and poisoning absorbers. According to residual poisoning and BA reactivity worth, rare earth oxides marked as suitable burnable absorbers are Nd, Sm, Eu, Gd, Dy, Ho and Er, while poisoning absorbers include Sc, La, Lu, Y, Ce, Pr and Tb. The presentation slides have been added to the article
Particle tracking code of simulating global RF feedback
International Nuclear Information System (INIS)
Mestha, L.K.
1991-09-01
It is well known in the ''control community'' that a good feedback controller design is deeply rooted in the physics of the system. For example, when accelerating the beam we must keep several parameters under control so that the beam travels within the confined space. Important parameters include the frequency and phase of the rf signal, the dipole field, and the cavity voltage. Because errors in these parameters will progressively mislead the beam from its projected path in the tube, feedback loops are used to correct the behavior. Since the feedback loop feeds energy to the system, it changes the overall behavior of the system and may drive it to instability. Various types of controllers are used to stabilize the feedback loop. Integrating the beam physics with the feedback controllers allows us to carefully analyze the beam behavior. This will not only guarantee optimal performance but will also significantly enhance the ability of the beam control engineer to deal effectively with the interaction of various feedback loops. Motivated by this theme, we developed a simple one-particle tracking code to simulate particle behavior with feedback controllers. In order to achieve our fundamental objective, we can ask some key questions: What are the input and output parameters? How can they be applied to the practical machine? How can one interface the rf system dynamics such as the transfer characteristics of the rf cavities and phasing between the cavities? Answers to these questions can be found by considering a simple case of a single cavity with one particle, tracking it turn-by-turn with appropriate initial conditions, then introducing constraints on crucial parameters. Critical parameters are rf frequency, phase, and amplitude once the dipole field has been given. These are arranged in the tracking code so that we can interface the feedback system controlling them
MOCCA Code for Star Cluster Simulation: Comparison with Optical Observations using COCOA
Askar, Abbas; Giersz, Mirek; Pych, Wojciech; Olech, Arkadiusz; Hypki, Arkadiusz
2014-01-01
We introduce and present preliminary results from COCOA (Cluster simulatiOn Comparison with ObservAtions) code for a star cluster after 12 Gyrs of evolution simulated using the MOCCA code. The COCOA code is being developed to quickly compare results of numerical simulations of star clusters with observational data. We use COCOA to obtain parameters of the projected cluster model. For comparison, a FITS file of the projected cluster was provided to observers so that they could use their observ...
Simulation of Hamming Coding and Decoding for Microcontroller Radiation Hardening
Rehab I. Abdul Rahman; Mazhar B. Tayel
2015-01-01
This paper presents a method of hardening the 8051 micro-controller, able to assure reliable operation in the presence of bit flips caused by radiation. Aiming at avoiding such faults in the 8051 micro-controller, Hamming code protection was used in its SRAM memory and registers. A VHDL code has been used for this hamming code protection.
A comparative study of MONTEBURNS and MCNPX 2.6.0 codes in ADS simulations
International Nuclear Information System (INIS)
Barros, Graiciany P.; Pereira, Claubia; Veloso, Maria A.F.; Velasquez, Carlos E.; Costa, Antonella L.
2013-01-01
The possible use of the MONTEBURNS and MCNPX 2.6.0 codes in Accelerator-driven systems (ADSs) simulations for fuel evolution description is discussed. ADSs are investigated for fuel breeding and long-lived fission product transmutation so simulations of fuel evolution have a great relevance. The burnup/depletion capability is present in both studied codes. MONTEBURNS code links Monte Carlo N-Particle Transport Code (MCNP) to the radioactive decay burnup code ORIGEN2, whereas MCNPX depletion/ burnup capability is a linked process involving steady-state flux calculations by MCNPX and nuclide depletion calculations by CINDER90. A lead-cooled accelerator-driven system fueled with thorium was simulated and the results obtained using MONTEBURNS code and the results from MCNPX 2.6.0 code were compared. The system criticality and the variation of the actinide inventory during the burnup were evaluated and the results indicate a similar behavior between the results of each code. (author)
Simulation of nonlinear propagation of biomedical ultrasound using PZFlex and the KZK Texas code
Qiao, Shan; Jackson, Edward; Coussios, Constantin-C.; Cleveland, Robin
2015-10-01
In biomedical ultrasound nonlinear acoustics can be important in both diagnostic and therapeutic applications and robust simulations tools are needed in the design process but also for day-to-day use such as treatment planning. For most biomedical application the ultrasound sources generate focused sound beams of finite amplitude. The KZK equation is a common model as it accounts for nonlinearity, absorption and paraxial diffraction and there are a number of solvers available, primarily developed by research groups. We compare the predictions of the KZK Texas code (a finite-difference time-domain algorithm) to an FEM-based commercial software, PZFlex. PZFlex solves the continuity equation and momentum conservation equation with a correction for nonlinearity in the equation of state incorporated using an incrementally linear, 2nd order accurate, explicit algorithm in time domain. Nonlinear ultrasound beams from two transducers driven at 1 MHz and 3.3 MHz respectively were simulated by both the KZK Texas code and PZFlex, and the pressure field was also measured by a fibre-optic hydrophone to validate the models. Further simulations were carried out a wide range of frequencies. The comparisons showed good agreement for the fundamental frequency for PZFlex, the KZK Texas code and the experiments. For the harmonic components, the KZK Texas code was in good agreement with measurements but PZFlex underestimated the amplitude: 32% for the 2nd harmonic and 66% for the 3rd harmonic. The underestimation of harmonics by PZFlex was more significant when the fundamental frequency increased. Furthermore non-physical oscillations in the axial profile of harmonics occurred in the PZFlex results when the amplitudes were relatively low. These results suggest that careful benchmarking of nonlinear simulations is important.
Simulation of nonlinear propagation of biomedical ultrasound using PZFlex and the KZK Texas code
Energy Technology Data Exchange (ETDEWEB)
Qiao, Shan, E-mail: shan.qiao@eng.ox.ac.uk; Jackson, Edward; Coussios, Constantin-C; Cleveland, Robin [Institute of Biomedical Engineering, Department of Engineering Science, University of Oxford, Oxford (United Kingdom)
2015-10-28
In biomedical ultrasound nonlinear acoustics can be important in both diagnostic and therapeutic applications and robust simulations tools are needed in the design process but also for day-to-day use such as treatment planning. For most biomedical application the ultrasound sources generate focused sound beams of finite amplitude. The KZK equation is a common model as it accounts for nonlinearity, absorption and paraxial diffraction and there are a number of solvers available, primarily developed by research groups. We compare the predictions of the KZK Texas code (a finite-difference time-domain algorithm) to an FEM-based commercial software, PZFlex. PZFlex solves the continuity equation and momentum conservation equation with a correction for nonlinearity in the equation of state incorporated using an incrementally linear, 2nd order accurate, explicit algorithm in time domain. Nonlinear ultrasound beams from two transducers driven at 1 MHz and 3.3 MHz respectively were simulated by both the KZK Texas code and PZFlex, and the pressure field was also measured by a fibre-optic hydrophone to validate the models. Further simulations were carried out a wide range of frequencies. The comparisons showed good agreement for the fundamental frequency for PZFlex, the KZK Texas code and the experiments. For the harmonic components, the KZK Texas code was in good agreement with measurements but PZFlex underestimated the amplitude: 32% for the 2nd harmonic and 66% for the 3rd harmonic. The underestimation of harmonics by PZFlex was more significant when the fundamental frequency increased. Furthermore non-physical oscillations in the axial profile of harmonics occurred in the PZFlex results when the amplitudes were relatively low. These results suggest that careful benchmarking of nonlinear simulations is important.
Simulation of nonlinear propagation of biomedical ultrasound using PZFlex and the KZK Texas code
International Nuclear Information System (INIS)
Qiao, Shan; Jackson, Edward; Coussios, Constantin-C; Cleveland, Robin
2015-01-01
In biomedical ultrasound nonlinear acoustics can be important in both diagnostic and therapeutic applications and robust simulations tools are needed in the design process but also for day-to-day use such as treatment planning. For most biomedical application the ultrasound sources generate focused sound beams of finite amplitude. The KZK equation is a common model as it accounts for nonlinearity, absorption and paraxial diffraction and there are a number of solvers available, primarily developed by research groups. We compare the predictions of the KZK Texas code (a finite-difference time-domain algorithm) to an FEM-based commercial software, PZFlex. PZFlex solves the continuity equation and momentum conservation equation with a correction for nonlinearity in the equation of state incorporated using an incrementally linear, 2nd order accurate, explicit algorithm in time domain. Nonlinear ultrasound beams from two transducers driven at 1 MHz and 3.3 MHz respectively were simulated by both the KZK Texas code and PZFlex, and the pressure field was also measured by a fibre-optic hydrophone to validate the models. Further simulations were carried out a wide range of frequencies. The comparisons showed good agreement for the fundamental frequency for PZFlex, the KZK Texas code and the experiments. For the harmonic components, the KZK Texas code was in good agreement with measurements but PZFlex underestimated the amplitude: 32% for the 2nd harmonic and 66% for the 3rd harmonic. The underestimation of harmonics by PZFlex was more significant when the fundamental frequency increased. Furthermore non-physical oscillations in the axial profile of harmonics occurred in the PZFlex results when the amplitudes were relatively low. These results suggest that careful benchmarking of nonlinear simulations is important
DEFROST: a new code for simulating preheating after inflation
International Nuclear Information System (INIS)
Frolov, Andrei V
2008-01-01
At the end of inflation, dynamical instability can rapidly deposit the energy of homogeneous cold inflaton into excitations of other fields. This process, known as preheating, is rather violent, inhomogeneous and non-linear, and has to be studied numerically. This paper presents a new code for simulating scalar field dynamics in an expanding universe written for that purpose. Compared to available alternatives, it significantly improves both the speed and the accuracy of calculations, and is fully instrumented for 3D visualization. We reproduce previously published results on preheating in simple chaotic inflation models, and further investigate non-linear dynamics of the inflaton decay. Surprisingly, we find that the fields do not 'want' to thermalize in quite the way that one would think. Instead of directly reaching equilibrium, the evolution appears to be stuck in a rather simple but quite inhomogeneous state. In particular, a one-point distribution function of total energy density appears to be universal among various two-field preheating models, and is exceedingly well described by a log-normal distribution. It is tempting to attribute this state to scalar field turbulence
FALCON Code Simulation for Verification of Fuel Preconditioning Guideline
Energy Technology Data Exchange (ETDEWEB)
Lee, Hee-Hun; Kwon, Oh-Hyun; Kim, Hong-Jin; Kim, Yong-Hwan [KEPCO Nuclear Fuel Co. Ltd., Daejeon (Korea, Republic of)
2015-10-15
The magnitude and rate of power increases are key factors in the PCI failure process. KEPCO NF (KNF) provides operational restrictions called fuel preconditioning guideline (FPG) to mitigate PCI failures. The FPG contains recommended power maneuvering restrictions that should be followed when the KNF supplied fuel is being operated in-reactor. This guideline typically includes controlled power ramp rates, threshold power levels to initiate controlled ramp rates, and restrictions on the operating conditions that impact the potential for PCI failure. The purpose of the FPG is to allow time for stress relaxation to reduce cladding stress buildup during power maneuvers. Two general approaches have been adopted in the development of FPG to mitigate PCI failure in operating commercial reactors. The first approach relies primarily on past operational experience and power ramp test. The second one uses an analytical methodology where a figure-of-merit representative of PCI vulnerability, generally cladding hoop stress, is calculated using a fuel performance code. FALCON simulation can be the identification of a PCI limit parameter, typically cladding hoop stress, which can be used to evaluate a power maneuvering restriction on FPG. The PCI analysis is to assess the cladding hoop stress under various power ramp conditions. Startup ramp rate doesn't affect PCI failure until 50% of rated thermal power.
LOLA SYSTEM: A code block for nodal PWR simulation. Part. II - MELON-3, CONCON and CONAXI Codes
Energy Technology Data Exchange (ETDEWEB)
Aragones, J M; Ahnert, C; Gomez Santamaria, J; Rodriguez Olabarria, I
1985-07-01
Description of the theory and users manual of the MELON-3, CONCON and CONAXI codes, which are part of the core calculation system by nodal theory in one group, called LOLA SYSTEM. These auxiliary codes, provide some of the input data for the main module SIMULA-3; these are, the reactivity correlations constants, the albe does and the transport factors. (Author) 7 refs.
LOLA SYSTEM: A code block for nodal PWR simulation. Part. II - MELON-3, CONCON and CONAXI Codes
International Nuclear Information System (INIS)
Aragones, J. M.; Ahnert, C.; Gomez Santamaria, J.; Rodriguez Olabarria, I.
1985-01-01
Description of the theory and users manual of the MELON-3, CONCON and CONAXI codes, which are part of the core calculation system by nodal theory in one group, called LOLA SYSTEM. These auxiliary codes, provide some of the input data for the main module SIMULA-3; these are, the reactivity correlations constants, the albe does and the transport factors. (Author) 7 refs
International Nuclear Information System (INIS)
Li, Junli; Qiu, Rui; Yan, Congchong; Xie, Wenzhang; Zeng, Zhi; Li, Chunyan; Wu, Zhen; Tung, Chuanjong
2015-01-01
The method of Monte Carlo simulation is a powerful tool to investigate the details of radiation biological damage at the molecular level. In this paper, a Monte Carlo code called NASIC (Nanodosimetry Monte Carlo Simulation Code) was developed. It includes physical module, pre-chemical module, chemical module, geometric module and DNA damage module. The physical module can simulate physical tracks of low-energy electrons in the liquid water event-by-event. More than one set of inelastic cross sections were calculated by applying the dielectric function method of Emfietzoglou's optical-data treatments, with different optical data sets and dispersion models. In the pre-chemical module, the ionised and excited water molecules undergo dissociation processes. In the chemical module, the produced radiolytic chemical species diffuse and react. In the geometric module, an atomic model of 46 chromatin fibres in a spherical nucleus of human lymphocyte was established. In the DNA damage module, the direct damages induced by the energy depositions of the electrons and the indirect damages induced by the radiolytic chemical species were calculated. The parameters should be adjusted to make the simulation results be agreed with the experimental results. In this paper, the influence study of the inelastic cross sections and vibrational excitation reaction on the parameters and the DNA strand break yields were studied. Further work of NASIC is underway (authors)
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
HELIOS/DRAGON/NESTLE codes' simulation of the Gentilly-2 loss of class 4 power event
International Nuclear Information System (INIS)
Sarsour, H.N.; Turinsky, P.J.; Rahnema, F.; Mosher, S.; Serghiuta, D.; Marleau, G.; Courau, T.
2002-01-01
A loss of electrical power occurred at Gentilly-2 in September of 1995 while the station was operating at full power. There was an unexpectedly rapid core power increase initiated by the drainage of the zone controllers and accelerated by coolant boiling. The core transient was terminated by Shutdown System No 1 (SDS1) tripping when the out-of-core ion chambers exceeded the 10%/sec high rate of power increase trip setpoint at 1.29 sec. This resulted in the station automatically shutting down within 2 sec of event initiation. In the first 2 sec, 26 of the 58 SDS1 and SDS2 in-core flux detectors reached there overpower trip (ROPT) setpoints. The peak reactor power reached approximately 110%FP. Reference 1 presented detailed results of the simulations performed with coupled thermalhydraulics and 3D neutron kinetics codes, SOPHT-G2 and the CERBERUS module of RFSP, and the various adjustments of these codes and plant representation that were needed to obtain the neutronic response observed in 1995. The purposes of this paper are to contrast a simulation prediction of the peak prompt core thermal power transient versus experimental estimate, and to note the impact of spatial discretization approach utilized on the prompt core thermal power transient and the channel power distribution as a function of time. In addition, adequacy of the time-step sizes employed and sensitivity to core's transient thermal-hydraulics conditions are studied. The work presented in this paper has been performed as part of a project sponsored by the Canadian Nuclear Safety Commission (CNSC). The purpose of the project was to gather information and assess the accuracy of best estimate methods using calculation methods and codes developed independently from the CANDU industry. The simulation of the accident was completed using the NESTLE core simulator, employing cross sections generated by the HELIOS lattice physics code, and incremental cross sections generated by the DRAGON lattice physics code
Simulation of water hammer experiments using RELAP5 code
International Nuclear Information System (INIS)
Kaliatka, A.; Vaisnoras, M.
2005-01-01
The rapid closing or opening of a valve causes pressure transients in pipelines. The fast deceleration of the liquid results in high pressure surges upstream the valve, thus the kinetic energy is transformed into the potential energy, which leads to the temporary pressure increases. This phenomenon is called water hammer. The intensity of water hammer effects will depend upon the rate of change in the velocity or momentum. Generally water hammer can occur in any thermal-hydraulic systems and it is extremely dangerous for the thermal-hydraulic system since, if the pressure induced exceeds the pressure range of a pipe given by the manufacturer, it can lead to the failure of the pipeline integrity. Due to its potential for damage of pipes, water hammer has been a subject of study since the middle of the nineteenth century. Many theoretical and experimental investigations were performed. The experimental investigation of the water hammer tests performed at Fraunhofer Institute for Environmental, Safety and Energy Technology (UMSICHT) [1] and Cold Water Hammer experiment performed by Forschungszentrum Rossendorf (CWHTF) [2] should be mentioned. The UMSICHT facility in Oberhausen was modified in order to simulate a piping system and associated supports that are typical for a nuclear power plant [3]. The Cold water hammer experiment is interesting and instructive because it covers a wide spectrum of particularities. One of them is sub-cooled water interaction with condensing steam at the closed end of the vertical pipe at room temperature and corresponding saturation pressure [4]. In the paper, the capabilities of RELAP5 code to correctly represent the water hammer phenomenon are presented. Paper presents the comparison of RELAP5 calculated and measured at UMSICHT and CWHTF test facilities pressure transient values after the fast closure (opening) of valves. The analyses of rarefaction wave travels inside the pipe and condensation of vapour bubbles in the liquid column
Annual coded wire tag program, Washington: Missing production groups. Annual report for 1998
International Nuclear Information System (INIS)
Byrne, J.; Fuss, H.
1999-01-01
The Bonneville Power Administration (BPA) funds the ''Annual Coded Wire Tag Program--Missing Production Groups for Columbia River Hatcheries'' project. The WDFW project has three main objectives: (1) coded-wire tag at least one production group of each species at each Columbia Basin hatchery to enable evaluation of survival and catch distribution over time, (2) recover coded-wire tags from the snouts of fish tagged under objective 1 and estimate survival, contribution, and stray rates for each group, and (3) report the findings under objective 2 for all broods of chinook, and coho released from WDFW Columbia Basin hatcheries
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)
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
SITA version 0. A simulation and code testing assistant for TOUGH2 and MARNIE
Energy Technology Data Exchange (ETDEWEB)
Seher, Holger; Navarro, Martin
2016-06-15
High quality standards have to be met by those numerical codes that are applied in long-term safety assessments for deep geological repositories for radioactive waste. The software environment SITA (''a simulation and code testing assistant for TOUGH2 and MARNIE'') has been developed by GRS in order to perform automated regression testing for the flow and transport simulators TOUGH2 and MARNIE. GRS uses the codes TOUGH2 and MARNIE in order to assess the performance of deep geological repositories for radioactive waste. With SITA, simulation results of TOUGH2 and MARNIE can be compared to analytical solutions and simulations results of other code versions. SITA uses data interfaces to operate with codes whose input and output depends on the code version. The present report is part of a wider GRS programme to assure and improve the quality of TOUGH2 and MARNIE. It addresses users as well as administrators of SITA.
Development of 2D particle-in-cell code to simulate high current, low ...
Indian Academy of Sciences (India)
Abstract. A code for 2D space-charge dominated beam dynamics study in beam trans- port lines is developed. The code is used for particle-in-cell (PIC) simulation of z-uniform beam in a channel containing solenoids and drift space. It can also simulate a transport line where quadrupoles are used for focusing the beam.
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)
A Steam Jet Plume Simulation in a Large Bulk Space with a System Code MARS
International Nuclear Information System (INIS)
Bae, Sung Won; Chung, Bub Dong
2006-01-01
From May 2002, the OECD-SETH group has launched the PANDA Project in order to provide an experimental data base for a multi-dimensional code assessment. OECD-SETH group expects the PANDA Project will meet the increasing needs for adequate experimental data for a 3D distribution of relevant variables like the temperature, velocity and steam-air concentrations that are measured with a sufficient resolution and accuracy. The scope of the PANDA Project is the mixture stratification and mixing phenomena in a large bulk space. Total of 24 test series are still being performed in PSI, Switzerland. The PANDA facility consists of 2 main large vessels and 1 connection pipe Within the large vessels, a steam injection nozzle and outlet vent are arranged for each test case. These tests are categorized into 3 modes, i.e. the high momentum, near wall plume, and free plume tests. KAERI has also participated in the SETH group since 1997 so that the multi-dimensional capability of the MARS code could be assessed and developed. Test 17, the high steam jet injection test, has already been simulated by MARS and shows promising results. Now, the test 9 and 9bis cases which use a low speed horizontal steam jet flow have been simulated and investigated
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)
Modelling group decision simulation through argumentation
Marreiros, Goreti; Novais, Paulo; Machado, José; Ramos, Carlos; Neves, José
2007-01-01
Group decision making plays an important role in today’s organisations. The impact of decision making is so high and complex, that rarely the decision making process is made individually. In Group Decision Argumentation, there is a set of participants, with different profiles and expertise levels, that exchange ideas or engage in a process of argumentation and counter-argumentation, negotiate, cooperate, collaborate or even discuss techniques and/or methodologies for problem solving. In this ...
SPIDERMAN: Fast code to simulate secondary transits and phase curves
Louden, Tom; Kreidberg, Laura
2017-11-01
SPIDERMAN calculates exoplanet phase curves and secondary eclipses with arbitrary surface brightness distributions in two dimensions. The code uses a geometrical algorithm to solve exactly the area of sections of the disc of the planet that are occulted by the star. Approximately 1000 models can be generated per second in typical use, which makes making Markov Chain Monte Carlo analyses practicable. The code is modular and allows comparison of the effect of multiple different brightness distributions for a dataset.
International Nuclear Information System (INIS)
Galicia A, J.; Francois L, J. L.; Bastida O, G. E.; Esquivel E, J.
2016-09-01
The development of the AZTLAN platform for the analysis and design of nuclear reactors is led by Instituto Nacional de Investigaciones Nucleares (ININ) and divided into four working groups, which have well-defined activities to achieve significant progress in this project individually and jointly. Within these working groups is the users group, whose main task is to use the codes that make up the AZTLAN platform to provide feedback to the developers, and in this way to make the final versions of the codes are efficient and at the same time reliable and easy to understand. In this paper we present the results provided by the AZNHEX v.1.0 code when simulating the core of a fast reactor cooled with sodium at steady state. The validation of these results is a fundamental part of the platform development and responsibility of the users group, so in this research the results obtained with AZNHEX are compared and analyzed with those provided by the Monte Carlo code MCNP-5, software worldwide used and recognized. A description of the methodology used with MCNP-5 is also presented for the calculation of the interest variables and the difference that is obtained with respect to the calculated with AZNHEX. (Author)
Effect of difference between group constants processed by codes TIMS and ETOX on integral quantities
International Nuclear Information System (INIS)
Takano, Hideki; Ishiguro, Yukio; Matsui, Yasushi.
1978-06-01
Group constants of 235 U, 238 U, 239 Pu, 240 Pu and 241 Pu have been produced with the processing code TIMS using the evaluated nuclear data of JENDL-1. The temperature and composition dependent self-shielding factors have been calculated for the two cases with and without considering mutual interference resonant nuclei. By using the group constants set produced by the TIMS code, the integral quantities, i.e. multiplication factor, Na-void reactivity effect and Doppler reactivity effect, are calculated and compared with those calculated with the use of the cross sections set produced by the ETOX code to evaluate accuracy of the approximate calculation method in ETOX. There is much difference in self-shielding factor in each energy group between the two codes. For the fast reactor assemblies under study, however, the integral quantities calculated with these two sets are in good agreement with each other, because of eventual cancelation of errors. (auth.)
Preface: Research advances in vadose zone hydrology through simulations with the TOUGH codes
International Nuclear Information System (INIS)
Finsterle, Stefan; Oldenburg, Curtis M.
2004-01-01
symposium, with special attention to issues related to the vadose zone and unsaturated flow systems. The first paper, written by the original developer of TOUGH, Karsten Pruess, provides an overview of the history of the TOUGH codes, the main physical processes considered, their mathematical and numerical implementation, and case studies. That paper is followed by a review article summarizing inverse modeling applications performed by iTOUGH2. A subsequent group of papers deals with diverse unsaturated zone systems, highlighting the versatility of the code to handle a variety of processes in different geologic settings. Simulation capabilities of the TOUGH codes are increasingly used for geologic carbon sequestration studies as testified by the next group of papers. The final series of papers demonstrates the use of the TOUGH codes in support of remediation and engineering applications. These studies discuss biological and reactive chemical transport simulations, the design of clean-up operations and landfill management, and the analysis of engineered soil stabilization. As guest editors, we thank the authors for their interesting contributions, and the many reviewers for their careful and constructive review comments. Finally, on behalf of all of the authors and ourselves, we express our sincerest appreciation to Rien van Genuchten for providing the opportunity to publish these papers together in a Special Section of ''Vadose Zone Journal''
Energy Technology Data Exchange (ETDEWEB)
Rosenthal, Andrew [New Mexico State Univ., Las Cruces, NM (United States)
2013-12-30
The DOE grant, “An Integrated Partnership to Create and Lead the Solar Codes and Standards Working Group,” to New Mexico State University created the Solar America Board for Codes and Standards (Solar ABCs). From 2007 – 2013 with funding from this grant, Solar ABCs identified current issues, established a dialogue among key stakeholders, and catalyzed appropriate activities to support the development of codes and standards that facilitated the installation of high quality, safe photovoltaic systems. Solar ABCs brought the following resources to the PV stakeholder community; Formal coordination in the planning or revision of interrelated codes and standards removing “stove pipes” that have only roofing experts working on roofing codes, PV experts on PV codes, fire enforcement experts working on fire codes, etc.; A conduit through which all interested stakeholders were able to see the steps being taken in the development or modification of codes and standards and participate directly in the processes; A central clearing house for new documents, standards, proposed standards, analytical studies, and recommendations of best practices available to the PV community; A forum of experts that invites and welcomes all interested parties into the process of performing studies, evaluating results, and building consensus on standards and code-related topics that affect all aspects of the market; and A biennial gap analysis to formally survey the PV community to identify needs that are unmet and inhibiting the market and necessary technical developments.
International Nuclear Information System (INIS)
Mochi, Ignacio
2005-01-01
The principal parameters of nuclear reactors are determined in the conceptual design stage.For that purpose, it is necessary to have flexible calculation tools that represent the principal dependencies of such parameters.This capability is of critical importance in the design of innovative nuclear reactors.In order to have a proper tool that could assist the conceptual design of innovative nuclear reactors, we developed and implemented a neutronic core calculus code: DIANA (Diffusion Integral Analytic Neutron Analysis).To calculate the required parameters, this code generates its own cross sections using an analytic two group, two zones diffusion scheme based only on a minimal set of data (i.e. 2200 m/s and fission averaged microscopic cross sections, Wescott factors and Effective Resonance Integrals).Both to calculate cross sections and core parameters, DIANA takes into account heterogeneity effects that are included when it evaluates each zone.Among them lays the disadvantage factor of each energy group.DIANA was totally implemented through Object Oriented Programming using C++ language. This eases source code understanding and would allow a quick expansion of its capabilities if needed.The final product is a versatile and easy-to-use code that allows core calculations with a minimal amount of data.It also contains the required tools needed to perform many variational calculations such as the parameterisation of effective multiplication factors for different radii of the core.The diffusion scheme s simplicity allows an easy following of the involved phenomena, making DIANA the most suitable tool to design reactors whose physics lays beyond the parameters of present reactors.All this reasons make DIANA a good candidate for future innovative reactor analysis
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)
Assessing Group Dynamics in a Mars Simulation
Bishop, S. L.
2007-10-01
International interest in psychosocial functioning generally and issues of group and inter-group function for space crews has increased as focus has shifted towards longer duration spaceflight and, particularly, the issues involved in sending a human crew to Mars (Kanas, et al., 2001; Dawson, 2002). Planning documents for a human mission to Mars such as the NASA Design Reference Mission (DRM 1.0) emphasize the need for adaptability of crewmembers and autonomy in the crew as a whole (Hoffman and Kaplan, 1997). Similarly a major study by the International Space University (ISU, 1991) emphasized the need for autonomy and initiative for a Mars crew given that many of the scenarios that will be encountered on Mars cannot be rehearsed on earth and given the lack of any realistic possibility for rescue of the crew. This research project was only one subset of data collected during the larger AustroMars Expedition at the Mars Desert Research Facility (MDRS) in 2006. The participating crew comprises part of a multi-year investigation on teams utilizing the MDRS facility. The program of research has included numerous researchers since 2002 with a progressive evolution of key foci addressing stress, personality, coping, adaptation, cognitive functioning, and group identity assessed across the duration period of the individual missions.
2005-01-01
Tech-X Corporation releases simulation code for solving complex problems in plasma physics : VORPAL code provides a robust environment for simulating plasma processes in high-energy physics, IC fabrications and material processing applications
78 FR 63459 - GPS Satellite Simulator Control Working Group Meeting
2013-10-24
... DEPARTMENT OF DEFENSE Air Force GPS Satellite Simulator Control Working Group Meeting AGENCY: Department of the Air Force. ACTION: Meeting Notice. SUMMARY: This meeting notice is to inform GPS simulator manufacturers, who supply products to the Department of Defense (DoD), and GPS simulator users, both government...
78 FR 67132 - GPS Satellite Simulator Control Working Group Meeting
2013-11-08
... DEPARTMENT OF DEFENSE Department of the Air Force GPS Satellite Simulator Control Working Group... simulator manufacturers, who supply products to the Department of Defense (DoD), and GPS simulator users..., and email address) to [email protected]us.af.mil and have your security personnel submit your VAR...
Forced circulation type steam generator simulation code: HT4
International Nuclear Information System (INIS)
Okamoto, Masaharu; Tadokoro, Yoshihiro
1982-08-01
The purpose of this code is a understanding of dynamic characteristics of the steam generator, which is a component of High-temperature Heat Transfer Components Test Unit. This unit is a number 4th test section of Helium Engineering Demonstration Loop (HENDEL). Features of this report are as follows, modeling of the steam generator, a basic relationship for the continuity equation, numerical analysis techniques of a non-linear simultaneous equation and computer graphics output techniques. Forced circulation type steam generator with strait tubes and horizontal cut baffles, applied in this code, have be designed at the Over All System Design of the VHTRex. The code is for use with JAERI's digital computer FACOM M200. About 1.5 sec required for each time step reiteration, then about 40 sec cpu time required for a standard problem. (author)
The FLUKA code: An accurate simulation tool for particle therapy
Battistoni, Giuseppe; Böhlen, Till T; Cerutti, Francesco; Chin, Mary Pik Wai; Dos Santos Augusto, Ricardo M; Ferrari, Alfredo; Garcia Ortega, Pablo; Kozlowska, Wioletta S; Magro, Giuseppe; Mairani, Andrea; Parodi, Katia; Sala, Paola R; Schoofs, Philippe; Tessonnier, Thomas; Vlachoudis, Vasilis
2016-01-01
Monte Carlo (MC) codes are increasingly spreading in the hadrontherapy community due to their detailed description of radiation transport and interaction with matter. The suitability of a MC code for application to hadrontherapy demands accurate and reliable physical models capable of handling all components of the expected radiation field. This becomes extremely important for correctly performing not only physical but also biologically-based dose calculations, especially in cases where ions heavier than protons are involved. In addition, accurate prediction of emerging secondary radiation is of utmost importance in innovative areas of research aiming at in-vivo treatment verification. This contribution will address the recent developments of the FLUKA MC code and its practical applications in this field. Refinements of the FLUKA nuclear models in the therapeutic energy interval lead to an improved description of the mixed radiation field as shown in the presented benchmarks against experimental data with bot...
FARO and KROTOS code simulation and analysis at JRC Ispra
Energy Technology Data Exchange (ETDEWEB)
Annunziato, A.; Yerkess, A.; Addabbo, C. [European Commission-Joint Research Centre, Inst. for Systems, Informatics and Safety, 21020 Ispra (Italy)
1998-01-01
The paper summarizes relevant results from the pre and post test calculations of fuel coolant interaction and quenching tests performed in the FARO and KROTOS test facilities. The main analytical tools adopted at JRC Ispra are the COMETA and the TEXAS codes. COMETA pre and post test calculations of FARO Test L-20 as well as an application of the code to KROTOS test facility are presented. The analysis provides the need to account for H{sub 2} generation models into the pre-mixing calculations. In addition salient results from the application of TEXAS to FARO and KROTOS tests are shown. (author)
Simulation codes and the impact of validation/uncertainty requirements
International Nuclear Information System (INIS)
Sills, H.E.
1995-01-01
Several of the OECD/CSNI members have adapted a proposed methodology for code validation and uncertainty assessment. Although the validation process adapted by members has a high degree of commonality, the uncertainty assessment processes selected are more variable, ranaing from subjective to formal. This paper describes the validation and uncertainty assessment process, the sources of uncertainty, methods of reducing uncertainty, and methods of assessing uncertainty.Examples are presented from the Ontario Hydro application of the validation methodology and uncertainty assessment to the system thermal hydraulics discipline and the TUF (1) system thermal hydraulics code. (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 Use of Code-Mixing among Pamonanese in Parata Ndaya Closed-Group Facebook
Directory of Open Access Journals (Sweden)
Joice Yulinda Luke
2015-05-01
Full Text Available Research intended to figure out why Pamonanes did code-mixing in Parata Ndaya, a Facebook closed-group site. The research applied qualitative method to get the types of code-mixing and reasons for doing code-mixing, while the analysis used Hoffman’s theory. Data were taken from comments of three active members of Parata Ndaya. Comments selected were mainly focused on political issues that happened during Regional House Representative Election in 2014. Data analysis reveals that code-mixing is mostly found in jokes and some comments about political leaders. Thus, the results can provide insights for Parata Ndaya members to build awareness on preserving their local language (i.e. Pamona language as well as to enhance solidarity among members of the group site.
NULIF: neutron spectrum generator, few-group constant calculator, and fuel depletion code
International Nuclear Information System (INIS)
Wittkopf, W.A.; Tilford, J.M.; Andrews, J.B. II; Kirschner, G.; Hassan, N.M.; Colpo, P.N.
1977-02-01
The NULIF code generates a microgroup neutron spectrum and calculates spectrum-weighted few-group parameters for use in a spatial diffusion code. A wide variety of fuel cells, non-fuel cells, and fuel lattices, typical of PWR (or BWR) lattices, are treated. A fuel depletion routine and change card capability allow a broad range of problems to be studied. Coefficient variation with fuel burnup, fuel temperature change, moderator temperature change, soluble boron concentration change, burnable poison variation, and control rod insertion are readily obtained. Heterogeneous effects, including resonance shielding and thermal flux depressions, are treated. Coefficients are obtained for one thermal group and up to three epithermal groups. A special output routine writes the few-group coefficient data in specified format on an output tape for automated fitting in the PDQ07-HARMONY system of spatial diffusion-depletion codes
Cognitive distance, absorptive capacity and group rationality : a simulation study
Curseu, P.L.; Krehel, O.; Evers, J.H.M.; Muntean, A.
2014-01-01
We report the results of a simulation study in which we explore the joint effect of group absorptive capacity (as the average individual rationality of the group members) and cognitive distance (as the distance between the most rational group member and the rest of the group) on the emergence of
Cognitive distance, absorptive capacity and group rationality : A simulation study
Curseu, P.L.; Krehel, O.; Evers, J.H.M.; Muntean, A.
2014-01-01
We report the results of a simulation study in which we explore the joint effect of group absorptive capacity (as the average individual rationality of the group members) and cognitive distance (as the distance between the most rational group member and the rest of the group) on the emergence of
GYSELA, a full-f global gyrokinetic Semi-Lagrangian code for ITG turbulence simulations
International Nuclear Information System (INIS)
Grandgirard, V.; Sarazin, Y.; Garbet, X.; Dif-Pradalier, G.; Ghendrih, Ph.; Crouseilles, N.; Latu, G.; Sonnendruecker, E.; Besse, N.; Bertrand, P.
2006-01-01
This work addresses non-linear global gyrokinetic simulations of ion temperature gradient (ITG) driven turbulence with the GYSELA code. The particularity of GYSELA code is to use a fixed grid with a Semi-Lagrangian (SL) scheme and this for the entire distribution function. The 4D non-linear drift-kinetic version of the code already showns the interest of such a SL method which exhibits good properties of energy conservation in non-linear regime as well as an accurate description of fine spatial scales. The code has been upgrated to run 5D simulations of toroidal ITG turbulence. Linear benchmarks and non-linear first results prove that semi-lagrangian codes can be a credible alternative for gyrokinetic simulations
Simulations of X-ray synchrotron beams using the EGS4 code system in medical applications
International Nuclear Information System (INIS)
Orion, I.; Henn, A.; Sagi, I.; Dilmanian, F.A.; Pena, L.; Rosenfeld, A.B.
2001-01-01
X-ray synchrotron beams are commonly used in biological and medical research. The availability of intense, polarized low-energy photons from the synchrotron beams provides a high dose transfer to biological materials. The EGS4 code system, which includes the photoelectron angular distribution, electron motion inside a magnetic field, and the LSCAT package, found to be the appropriate Monte Carlo code for synchrotron-produced X-ray simulations. The LSCAT package was developed in 1995 for the EGS4 code to contain the routines to simulate the linear polarization, the bound Compton, and the incoherent scattering functions. Three medical applications were demonstrated using the EGS4 Monte Carlo code as a proficient simulation code system for the synchrotron low-energy X-ray source. (orig.)
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.
Simulation of hydrogen deflagration experiments in the ENACCEF facility using ASTEC code
International Nuclear Information System (INIS)
Povilaitis, Mantas; Urbonavicius, Egidijus; Rimkevicius, Sigitas
2011-01-01
During a hypothetic severe accident in the NPP involving degradation of the core of a light water reactor, hydrogen could be generated and released into the containment atmosphere posing a deflagration or even a detonation risk. In the case of deflagration, the integrity of the containment would be threatened by the increase of the containment atmosphere pressure and temperature. Other risks of containment damage due to turbulent flames exist, caused by high pressure pulses, shock waves and etc. For the simulation of such processes a reliable numerical codes are needed. Despite flame acceleration being largely studied for homogeneous hydrogen - air mixtures, there are still unresolved issues in this research area, e.g., the effect of turbulence level on flame acceleration and quenching. This paper presents simulations of hydrogen deflagration experiments in the ENACCEF facility using ASTEC code, performed in the frames of International Standard Program No. 49 and SARNET2 project. Experiments and simulations were performed with the aim of evaluating the codes' (a number of participants with various codes participated in the project) capabilities to simulate hydrogen combustion. ASTEC code is an integral lumped-parameter approach based nuclear safety analysis code. For the presented simulations, ASTEC modules CPA (containment thermohydromechanics) and FRONT (hydrogen deflagration) were used. Paper present ENACCEF test facility, its nodalisation schemes developed for the calculations, simulated experiments and simulations' results. Brief description of FRONT module is also presented. Calculations' results are compared with experimental results and analyzed. (author)
Cognitive distance, absorptive capacity and group rationality: a simulation study.
Directory of Open Access Journals (Sweden)
Petru Lucian Curşeu
Full Text Available We report the results of a simulation study in which we explore the joint effect of group absorptive capacity (as the average individual rationality of the group members and cognitive distance (as the distance between the most rational group member and the rest of the group on the emergence of collective rationality in groups. We start from empirical results reported in the literature on group rationality as collective group level competence and use data on real-life groups of four and five to validate a mathematical model. We then use this mathematical model to predict group level scores from a variety of possible group configurations (varying both in cognitive distance and average individual rationality. Our results show that both group competence and cognitive distance are necessary conditions for emergent group rationality. Group configurations, in which the groups become more rational than the most rational group member, are groups scoring low on cognitive distance and scoring high on absorptive capacity.
Cognitive distance, absorptive capacity and group rationality: a simulation study.
Curşeu, Petru Lucian; Krehel, Oleh; Evers, Joep H M; Muntean, Adrian
2014-01-01
We report the results of a simulation study in which we explore the joint effect of group absorptive capacity (as the average individual rationality of the group members) and cognitive distance (as the distance between the most rational group member and the rest of the group) on the emergence of collective rationality in groups. We start from empirical results reported in the literature on group rationality as collective group level competence and use data on real-life groups of four and five to validate a mathematical model. We then use this mathematical model to predict group level scores from a variety of possible group configurations (varying both in cognitive distance and average individual rationality). Our results show that both group competence and cognitive distance are necessary conditions for emergent group rationality. Group configurations, in which the groups become more rational than the most rational group member, are groups scoring low on cognitive distance and scoring high on absorptive capacity.
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.)
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.
HELIOS/DRAGON/NESTLE codes' simulation of void reactivity in a CANDU core
International Nuclear Information System (INIS)
Sarsour, H.N.; Rahnema, F.; Mosher, S.; Turinsky, P.J.; Serghiuta, D.; Marleau, G.; Courau, T.
2002-01-01
This paper presents results of simulation of void reactivity in a CANDU core using the NESTLE core simulator, cross sections from the HELIOS lattice physics code in conjunction with incremental cross sections from the DRAGON lattice physics code. First, a sub-region of a CANDU6 core is modeled using the NESTLE core simulator and predictions are contrasted with predictions by the MCNP Monte Carlo simulation code utilizing a continuous energy model. In addition, whole core modeling results are presented using the NESTLE finite difference method (FDM), NESTLE nodal method (NM) without assembly discontinuity factors (ADF), and NESTLE NM with ADF. The work presented in this paper has been performed as part of a project sponsored by the Canadian Nuclear Safety Commission (CNSC). The purpose of the project was to gather information and assess the accuracy of best estimate methods using calculational methods and codes developed independently from the CANDU industry. (author)
Nyx: Adaptive mesh, massively-parallel, cosmological simulation code
Almgren, Ann; Beckner, Vince; Friesen, Brian; Lukic, Zarija; Zhang, Weiqun
2017-12-01
Nyx code solves equations of compressible hydrodynamics on an adaptive grid hierarchy coupled with an N-body treatment of dark matter. The gas dynamics in Nyx use a finite volume methodology on an adaptive set of 3-D Eulerian grids; dark matter is represented as discrete particles moving under the influence of gravity. Particles are evolved via a particle-mesh method, using Cloud-in-Cell deposition/interpolation scheme. Both baryonic and dark matter contribute to the gravitational field. In addition, Nyx includes physics for accurately modeling the intergalactic medium; in optically thin limits and assuming ionization equilibrium, the code calculates heating and cooling processes of the primordial-composition gas in an ionizing ultraviolet background radiation field.
Annual coded wire tag program (Washington) missing production groups : annual report 2000; ANNUAL
International Nuclear Information System (INIS)
Dammers, Wolf; Mills, Robin D.
2002-01-01
The Bonneville Power Administration (BPA) funds the ''Annual Coded-wire Tag Program - Missing Production Groups for Columbia River Hatcheries'' project. The Washington Department of Fish and Wildlife (WDFW), Oregon Department of Fish and Wildlife (ODFW) and the United States Fish and Wildlife Service (USFWS) all operate salmon and steelhead rearing programs in the Columbia River basin. The intent of the funding is to coded-wire tag at least one production group of each species at each Columbia Basin hatchery to provide a holistic assessment of survival and catch distribution over time and to meet various measures of the Northwest Power Planning Council's (NWPPC) Fish and Wildlife Program. The WDFW project has three main objectives: (1) coded-wire tag at least one production group of each species at each Columbia Basin hatchery to enable evaluation of survival and catch distribution over time, (2) recover coded-wire tags from the snouts of fish tagged under objective 1 and estimate survival, contribution, and stray rates for each group, and (3) report the findings under objective 2 for all broods of chinook, and coho released from WDFW Columbia Basin hatcheries. Objective 1 for FY-00 was met with few modifications to the original FY-00 proposal. Under Objective 2, snouts containing coded-wire tags that were recovered during FY-00 were decoded. Under Objective 3, this report summarizes available recovery information through 2000 and includes detailed information for brood years 1989 to 1994 for chinook and 1995 to 1997 for coho
Premar-2: a Monte Carlo code for radiative transport simulation in atmospheric environments
International Nuclear Information System (INIS)
Cupini, E.
1999-01-01
The peculiarities of the PREMAR-2 code, aimed at radiation transport Monte Carlo simulation in atmospheric environments in the infrared-ultraviolet frequency range, are described. With respect to the previously developed PREMAR code, besides plane multilayers, spherical multilayers and finite sequences of vertical layers, each one with its own atmospheric behaviour, are foreseen in the new code, together with the refraction phenomenon, so that long range, highly slanted paths can now be more faithfully taken into account. A zenithal angular dependence of the albedo coefficient has moreover been introduced. Lidar systems, with spatially independent source and telescope, are allowed again to be simulated, and, in this latest version of the code, sensitivity analyses to be performed. According to this last feasibility, consequences on radiation transport of small perturbations in physical components of the atmospheric environment may be analyze and the related effects on searched results estimated. The availability of a library of physical data (reaction coefficients, phase functions and refraction indexes) is required by the code, providing the essential features of the environment of interest needed of the Monte Carlo simulation. Variance reducing techniques have been enhanced in the Premar-2 code, by introducing, for instance, a local forced collision technique, especially apt to be used in Lidar system simulations. Encouraging comparisons between code and experimental results carried out at the Brasimone Centre of ENEA, have so far been obtained, even if further checks of the code are to be performed [it
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
pTSC: Data file editing for the Tokamak Simulation Code
International Nuclear Information System (INIS)
Meiss, J.D.
1987-09-01
The code pTSC is an editor for the data files needed to run the Princeton Tokamak Simulation Code (TSC). pTSC utilizes the Macintosh interface to create a graphical environment for entering the data. As most of the data to run TSC consists of conductor positions, the graphical interface is especially appropriate
EVOLUTION OF GALAXY GROUPS IN THE ILLUSTRIS SIMULATION
Energy Technology Data Exchange (ETDEWEB)
Raouf, Mojtaba; Khosroshahi, Habib G. [School of Astronomy, Institute for Research in Fundamental Sciences (IPM), Tehran, 19395-5746 (Iran, Islamic Republic of); Dariush, A., E-mail: m.raouf@ipm.ir [Institute of Astronomy, University of Cambridge, Madingley Road, Cambridge CB3 0HA (United Kingdom)
2016-06-20
We present the first study of the evolution of galaxy groups in the Illustris simulation. We focus on dynamically relaxed and unrelaxed galaxy groups representing dynamically evolved and evolving galaxy systems, respectively. The evolutionary state of a group is probed from its luminosity gap and separation between the brightest group galaxy and the center of mass of the group members. We find that the Illustris simulation overproduces galaxy systems with a large luminosity gap, known as fossil systems, in comparison to observations and the probed semi-analytical predictions. However, this simulation is just as successful as the probed semi-analytic model in recovering the correlation between luminosity gap and offset of the luminosity centroid. We find evolutionary tracks based on luminosity gap that indicate that a group with a large luminosity gap is rooted in one with a small luminosity gap, regardless of the position of the brightest group galaxy within the halo. This simulation helps to explore, for the first time, the black hole mass and its accretion rate in galaxy groups. For a given stellar mass of the brightest group galaxies, the black hole mass is larger in dynamically relaxed groups with a lower rate of mass accretion. We find this to be consistent with the latest observational studies of radio activity in the brightest group galaxies in fossil groups. We also find that the intragalactic medium in dynamically evolved groups is hotter for a given halo mass than that in evolving groups, again consistent with earlier observational studies.
COUPLED SIMULATION OF GAS COOLED FAST REACTOR FUEL ASSEMBLY WITH NESTLE CODE SYSTEM
Directory of Open Access Journals (Sweden)
Filip Osusky
2018-05-01
Full Text Available The paper is focused on coupled calculation of the Gas Cooled Fast Reactor. The proper modelling of coupled neutronics and thermal-hydraulics is the corner stone for future safety assessment of the control and emergency systems. Nowadays, the system and channel thermal-hydraulic codes are accepted by the national regulatory authorities in European Union for license purposes, therefore the code NESTLE was used for the simulation. The NESTLE code is a coupled multigroup neutron diffusion code with thermal-hydraulic sub-channel code. In the paper, the validation of NESTLE code 5.2.1 installation is presented. The processing of fuel assembly homogeneous parametric cross-section library for NESTLE code simulation is made by the sequence TRITON of SCALE code package system. The simulated case in the NESTLE code is one fuel assembly of GFR2400 concept with reflective boundary condition in radial direction and zero flux boundary condition in axial direction. The results of coupled calculation are presented and are consistent with the GFR2400 study of the GoFastR project.
ICOOL: A Simulation Code for Ionization Cooling of Muon Beams
International Nuclear Information System (INIS)
Fernow, R. C.
1999-01-01
Current ideas [1,2] for designing a high luminosity muon collider require significant cooling of the phase space of the muon beams. The only known method that can cool the beams in a time comparable to the muon lifetime is ionization cooling [3,4]. This method requires directing the particles in the beam at a large angle through a low Z absorber material in a strong focusing magnetic channel and then restoring the longitudinal momentum with an rf cavity. We have developed a new 3-D tracking code ICOOL for examining possible configurations for muon cooling. A cooling system is described in terms of a series of longitudinal regions with associated material and field properties. The tracking takes place in a coordinate system that follows a reference orbit through the system. The code takes into account decays and interactions of ∼50-500 MeV/c muons in matter. Material geometry regions include cylinders and wedges. A number of analytic models are provided for describing the field configurations. Simple diagnostics are built into the code, including calculation of emittances and correlations, longitudinal traces, histograms and scatter plots. A number of auxiliary files can be generated for post-processing analysis by the user
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)
Modeling the reactor core of MNSR to simulate its dynamic behavior using the code PARET
International Nuclear Information System (INIS)
Hainoun, A.; Alhabet, F.
2004-02-01
Using the computer code PARET the core of the MNSR reactor was modelled and the neutronics and thermal hydraulic behaviour of the reactor core for the steady state and selected transients, that deal with step change of reactivity including control rod withdraw starting from steady state at various low power level, were simulated. For this purpose a PARET input model for the core of MNSR reactor has been developed enabling the simulation of neutron kinetic and thermal hydraulic of reactor core including reactivity feedback effects. The neutron kinetic model depends on the point kinetic with 15 groups delayed neutrons including photo neutrons of beryllium reflector. In this regard the effect of photo neutron on the dynamic behaviour has been analysed through two additional calculation. In the first the yield of photo neutrons was neglected completely and in the second its share was added to the sixth group of delayed neutrons. In the thermal hydraulic model the fuel elements with their cooling channels were distributed to 4 different groups with various radial power factors. The pressure lose factors for friction, flow direction change, expansion and contraction were estimated using suitable approaches. The post calculations of the relative neutron flux change and core average temperature were found to be consistent with the experimental measurements. Furthermore, the simulation has indicated the influence of photo neutrons of the Beryllium reflector on the neutron flux behaviour. For the reliability of the results sensitivity analysis was carried out to consider the uncertainty in some important parameters like temperature feedback coefficient and flow velocity. On the other hand the application of PARET in simulation of void formation in the subcooled boiling regime were tested. The calculation indicates the capability of PARET in modelling this phenomenon. However, big discrepancy between calculation results and measurement of axial void distribution were observed
Benchmarking and scaling studies of pseudospectral code Tarang for turbulence simulations
VERMA, MAHENDRA K
2013-09-21
Tarang is a general-purpose pseudospectral parallel code for simulating flows involving fluids, magnetohydrodynamics, and Rayleigh–Bénard convection in turbulence and instability regimes. In this paper we present code validation and benchmarking results of Tarang. We performed our simulations on 10243, 20483, and 40963 grids using the HPC system of IIT Kanpur and Shaheen of KAUST. We observe good ‘weak’ and ‘strong’ scaling for Tarang on these systems.
Benchmarking and scaling studies of pseudospectral code Tarang for turbulence simulations
VERMA, MAHENDRA K; CHATTERJEE, ANANDO; REDDY, K SANDEEP; YADAV, RAKESH K; PAUL, SUPRIYO; CHANDRA, MANI; Samtaney, Ravi
2013-01-01
Tarang is a general-purpose pseudospectral parallel code for simulating flows involving fluids, magnetohydrodynamics, and Rayleigh–Bénard convection in turbulence and instability regimes. In this paper we present code validation and benchmarking results of Tarang. We performed our simulations on 10243, 20483, and 40963 grids using the HPC system of IIT Kanpur and Shaheen of KAUST. We observe good ‘weak’ and ‘strong’ scaling for Tarang on these systems.
International Nuclear Information System (INIS)
Lorenzen, J.
1990-01-01
A new Fuel rod Performance Simulator code FRPS has been developed, tested and benchmarked and is now available in different versions. The user may choose between the batch version INTERPIN producing results in form of listings or beforehand defined plots, or the interactive simulator code SIMSIM which is stepping through a power history under the control of user. Both versions are presently running on minicomputers and PC:s using EGA-Graphics. A third version is the implementation in a Studsvik Compact Simulator with FRPS being one of its various modules receiving the dynamic inputs from the simulator
Application of CORSIKA Simulation Code to Study Lateral and ...
Indian Academy of Sciences (India)
J. Astrophys. ... and iron primary cosmic rays in different energies. ... distance to the shower axis, shape function of particles density, percentage of fluorescent light, ..... (c). Figure 2. Simulation results for proton primaries with other energies.
A new scheme for ATLAS trigger simulation using legacy code
International Nuclear Information System (INIS)
Galster, Gorm; Stelzer, Joerg; Wiedenmann, Werner
2014-01-01
Analyses at the LHC which search for rare physics processes or determine with high precision Standard Model parameters require accurate simulations of the detector response and the event selection processes. The accurate determination of the trigger response is crucial for the determination of overall selection efficiencies and signal sensitivities. For the generation and the reconstruction of simulated event data, the most recent software releases are usually used to ensure the best agreement between simulated data and real data. For the simulation of the trigger selection process, however, ideally the same software release that was deployed when the real data were taken should be used. This potentially requires running software dating many years back. Having a strategy for running old software in a modern environment thus becomes essential when data simulated for past years start to present a sizable fraction of the total. We examined the requirements and possibilities for such a simulation scheme within the ATLAS software framework and successfully implemented a proof-of-concept simulation chain. One of the greatest challenges was the choice of a data format which promises long term compatibility with old and new software releases. Over the time periods envisaged, data format incompatibilities are also likely to emerge in databases and other external support services. Software availability may become an issue, when e.g. the support for the underlying operating system might stop. In this paper we present the encountered problems and developed solutions, and discuss proposals for future development. Some ideas reach beyond the retrospective trigger simulation scheme in ATLAS as they also touch more generally aspects of data preservation.
Overview on pre-harmonization studies conducted by the Working Group on Codes and Standards
International Nuclear Information System (INIS)
Guinovart, J.
1998-01-01
For more than twenty years, the Working Group on Codes and Standards (WGCS) has been an Advisory Expert Group of the European Commission and three subgroups were formed to consider manufacture and inspection, structural mechanics and materials topics. The WGCS seeks to promote studies at the pre-harmonisation level, for the clarification and building of consensus in the European Community concerning technical issues of relevance for the integrity of safety-related components. It deals with pre-standardization process regarding industrial codes whose rules are applicable to design, construction and operation of NPP components in European Community
MCMEG: Simulations of both PDD and TPR for 6 MV LINAC photon beam using different MC codes
International Nuclear Information System (INIS)
Fonseca, T.C.F.; Mendes, B.M.; Lacerda, M.A.S.; Silva, L.A.C.; Paixão, L.
2017-01-01
The Monte Carlo Modelling Expert Group (MCMEG) is an expert network specializing in Monte Carlo radiation transport and the modelling and simulation applied to the radiation protection and dosimetry research field. For the first inter-comparison task the group launched an exercise to model and simulate a 6 MV LINAC photon beam using the Monte Carlo codes available within their laboratories and validate their simulated results by comparing them with experimental measurements carried out in the National Cancer Institute (INCA) in Rio de Janeiro, Brazil. The experimental measurements were performed using an ionization chamber with calibration traceable to a Secondary Standard Dosimetry Laboratory (SSDL). The detector was immersed in a water phantom at different depths and was irradiated with a radiation field size of 10×10 cm 2 . This exposure setup was used to determine the dosimetric parameters Percentage Depth Dose (PDD) and Tissue Phantom Ratio (TPR). The validation process compares the MC calculated results to the experimental measured PDD20,10 and TPR20,10. Simulations were performed reproducing the experimental TPR20,10 quality index which provides a satisfactory description of both the PDD curve and the transverse profiles at the two depths measured. This paper reports in detail the modelling process using MCNPx, MCNP6, EGSnrc and Penelope Monte Carlo codes, the source and tally descriptions, the validation processes and the results. - Highlights: • MCMEG is an expert network specializing in Monte Carlo radiation transport. • MCNPx, MCNP6, EGSnrc and Penelope Monte Carlo codes are used. • Exercise to model and simulate a 6 MV LINAC photon beam using the Monte Carlo codes. • The PDD 20,10 and TPR 20,10 dosimetric parameters were compared with real data. • The paper reports in the modelling process using different Monte Carlo codes.
Assessing the impact of automated coding & grouping technology at St Vincent's Hospital, Sydney.
Howes, M H
1993-12-01
In 1992 the Hospital recognised that the existing casemix data reporting systems were too removed from individual patients to have any meaning for clinicians, analysis of the data was difficult and the processes involved in the DRG assignment were subject to considerable error. Consequently, the Hospital approved the purchase of technology that would facilitate the coding and grouping process. The impact of automated coding and grouping technology is assessed by three methods. Firstly, by looking at by-product information systems, secondly, through subjective responses by coders to a satisfaction questionnaire and, thirdly, by objectively measuring hospital activity and identified coding elements before and after implementation of the 3M technology. It was concluded that while the 3M Coding and Grouping software should not be viewed as a panacea to all coding and documentation ills, objective evidence and subjective comment from the coders indicated an improvement in data quality and more accurate DRG assignment. Development of an in-house casemix information system and a feedback mechanism between coder and clinician had been effected. The product had been used as a training tool for coders and had also proven to be a useful auditing tool. Finally, linkage with other systems and the generation of timely reports had been realised.
Lacan - a global simulation code for laser isotope separation
International Nuclear Information System (INIS)
Goldstein, S.; Quaegebeur, J.P.
1990-01-01
Dimensioning a Laser Isotope Separation (LIS) plant means calculating the values of a large number of parameters in order to optimize some objective function. In such algorithms the calculation of the objective function must be repeated thousands of times, therefore each elementary calculation must consume little time. LACAN uses simple models to describe the elementary physical processes: evaporation, vapour expansion, interaction between photons and atoms, ion extraction etc ... These simple models are derived from refined modeling codes or are empirical. As an example the optimization of the separative work of an uranium facility is discussed
On the adequacy of numerical codes for the simulation of vapour cloud explosions
International Nuclear Information System (INIS)
Wingerden, G.J.M.v.; Berg, A.C.v.d.
1984-01-01
Three spherically symmetric blast simulation codes have been evaluated: a low-flame-speed model (Piston model) and two gasdynamic blast simulation codes (BLAST and CLOUD). Self-similar flow fields in front of constant velocity flames and large- and small-scale spherically symmetric explosions experiments were simulated. The Piston model can be used for the simulation of spherically symmetric explosions at flame speeds -1 whereas BLAST and CLOUD are adequate for flame speeds exceeding 100 ms -1 . An adapted Piston code has been investigated with respect to the capability of simulating blast due to explosions of pancake-shaped clouds. Comparison to an acoustic approach showed that the Piston model can be regarded as an acoustic model with the possibility of handling every imaginable flame path. The research was part of the indirect action research programme on LWR Safety of the Commission of the European Communities. (project 12B, contract 008 SRN)
A one-dimensional transport code for the simulation of D-T burning tokamak plasma
International Nuclear Information System (INIS)
Tone, Tatsuzo; Maki, Koichi; Kasai, Masao; Nishida, Hidetsugu
1980-11-01
A one-dimensional transport code for D-T burning tokamak plasma has been developed, which simulates the spatial behavior of fuel ions(D, T), alpha particles, impurities, temperatures of ions and electrons, plasma current, neutrals, heating of alpha and injected beam particles. The basic transport equations are represented by one generalized equation so that the improvement of models and the addition of new equations may be easily made. A model of burn control using a variable toroidal field ripple is employed. This report describes in detail the simulation model, numerical method and the usage of the code. Some typical examples to which the code has been applied are presented. (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.
Simulation of loss of feedwater transient of MASLWR test facility by MARS-KS code
Energy Technology Data Exchange (ETDEWEB)
Park, Juyeop [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)
2013-05-15
MASLWR test facility is a mock-up of a passive integral type reactor equipped with helical coil steam generator. Since SMART reactor which is being current developed domestically also adopts helical coil steam generator, KINS has joined this ICSP to evaluate performance of domestic regulatory audit thermal-hydraulic code (MARS-KS code) in various respects including wall-to-fluid heat transfer model modification implemented in the code by independent international experiment database. In the ICSP, two types of transient experiments have been focused and they are loss of feedwater transient with subsequent ADS operation and long term cooling (SP-2) and normal operating conditions at different power levels (SP-3). In the present study, KINS simulation results by the MARS-KS code (KS-002 version) for the SP-2 experiment are presented in detail and conclusions on MARS-KS code performance drawn through this simulation is described. Performance of the MARS-KS code is evaluated through the simulation of the loss of feedwater transient of the MASLWR test facility. Steady state run shows helical coil specific heat transfer models implemented in the code is reasonable. However, through the transient run, it is also found that three-dimensional effect within the HPC and axial conduction effect through the HTP are not well reproduced by the code.
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
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.
Applications of the LAHET simulation code to relativistic heavy ion detectors
International Nuclear Information System (INIS)
Waters, L.S.; Gavron, A.
1991-01-01
The Los Alamos High Energy Transport (LAHET) simulation code has been applied to test beam data from the lead/scintillator Participant Calorimeter of BNL AGS experiment E814. The LAHET code treats hadronic interactions with the LANL version of the Oak Ridge code HETC. LAHET has now been expanded to handle hadrons with kinetic energies greater than 5 GeV with the FLUKA code, while HETC is used exclusively below 2.0 GeV. FLUKA is phased in linearly between 2.0 and 5.0 GeV. Transport of electrons and photons is done with EGS4, and an interface to the Los Alamos HMCNP3B library based code is provided to analyze neutrons with kinetic energies less than 20 MeV. Excellent agreement is found between the test data and simulation, and results for 2.46 GeV/c protons and pions are illustrated in this article
Applications of the lahet simulation code to relativistic heavy ion detectors
Energy Technology Data Exchange (ETDEWEB)
Waters, L.; Gavron, A. [Los Alamos National Lab., NM (United States)
1991-12-31
The Los Alamos High Energy Transport (LAHET) simulation code has been applied to test beam data from the lead/scintillator Participant Calorimeter of BNL AGS experiment E814. The LAHET code treats hadronic interactions with the LANL version of the Oak Ridge code HETC. LAHET has now been expanded to handle hadrons with kinetic energies greater than 5 GeV with the FLUKA code, while HETC is used exclusively below 2.0 GeV. FLUKA is phased in linearly between 2.0 and 5.0 GeV. Transport of electrons and photons is done with EGS4, and an interface to the Los Alamos HMCNP3B library based code is provided to analyze neutrons with kinetic energies less than 20 MeV. Excellent agreement is found between the test data and simulation, and results for 2.46 GeV/c protons and pions are illustrated in this article.
Object-Oriented Parallel Particle-in-Cell Code for Beam Dynamics Simulation in Linear Accelerators
International Nuclear Information System (INIS)
Qiang, J.; Ryne, R.D.; Habib, S.; Decky, V.
1999-01-01
In this paper, we present an object-oriented three-dimensional parallel particle-in-cell code for beam dynamics simulation in linear accelerators. A two-dimensional parallel domain decomposition approach is employed within a message passing programming paradigm along with a dynamic load balancing. Implementing object-oriented software design provides the code with better maintainability, reusability, and extensibility compared with conventional structure based code. This also helps to encapsulate the details of communications syntax. Performance tests on SGI/Cray T3E-900 and SGI Origin 2000 machines show good scalability of the object-oriented code. Some important features of this code also include employing symplectic integration with linear maps of external focusing elements and using z as the independent variable, typical in accelerators. A successful application was done to simulate beam transport through three superconducting sections in the APT linac design
Report of the working group on detector simulation
International Nuclear Information System (INIS)
Price, L.E.; Lebrun, P.
1986-01-01
An ad hoc group at Snowmass reviewed the need for detector simulation to support detectors at the SSC. This report first reviews currently available programs for detector simulation, both those written for single specific detectors and those aimed at general utility. It then considers the requirements for detector simulation for the SSC, with particular attention to enhancements that are needed relative to present programs. Finally, a list of recommendations is given
Energy Technology Data Exchange (ETDEWEB)
Muramatsu, Toshiharu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center
1998-08-01
This report explains the numerical methods and the set-up method of input data for a single-phase multi-dimensional thermohydraulics direct numerical simulation code DINUS-3 (Direct Numerical Simulation using a 3rd-order upwind scheme). The code was developed to simulate non-stationary temperature fluctuation phenomena related to thermal striping phenomena, developed at Power Reactor and Nuclear Fuel Development Corporation (PNC). The DINUS-3 code was characterized by the use of a third-order upwind scheme for convection terms in instantaneous Navier-Stokes and energy equations, and an adaptive control system based on the Fuzzy theory to control time step sizes. Author expect this report is very useful to utilize the DINUS-3 code for the evaluation of various non-stationary thermohydraulic phenomena in reactor applications. (author)
A PIC-MCC code for simulation of streamer propagation in air
DEFF Research Database (Denmark)
Chanrion, Olivier Arnaud; Neubert, Torsten
2008-01-01
A particle code has been developed to study the distribution and acceleration of electrons in electric discharges in air. The code can follow the evolution of a discharge from the initial stage of a single free electron in a background electric field to the formation of an electron avalanche...... and its transition into a streamer. The code is in 2D axi-symmetric coordinates, allowing quasi 3D simulations during the initial stages of streamer formation. This is important for realistic simulations of problems where space charge fields are essential such as in streamer formation. The charged...... particles are followed in a Cartesian mesh and the electric field is updated with Poisson's equation from the charged particle densities. Collisional processes between electrons and air molecules are simulated with a Monte Carlo technique, according to cross section probabilities. The code also includes...
International Nuclear Information System (INIS)
Taji, Yukichi
1984-06-01
Development has been made of molecular dynamical codes DGR and GGR to simulate radiation damages yielded in the diamond and graphite structure crystals, respectively. Though the usual molecular dynamical codes deal only with the central forces as the mutual interactions between atoms, the present codes can take account of noncentral forces to represent the effect of the covalent bonds characteristic of diamond or graphite crystals. It is shown that lattice defects yielded in these crystals are stable by themselves in the present method without any supports of virtual surface forces set on the crystallite surfaces. By this effect the behavior of lattice defects has become possible to be simulated in a more realistic manner. Some examples of the simulation with these codes are shown. (author)
Code of Ethics for the American Association of Physicists in Medicine: report of Task Group 109.
Serago, Christopher F; Adnani, Nabil; Bank, Morris I; BenComo, Jose A; Duan, Jun; Fairobent, Lynne; Freedman, D Jay; Halvorsen, Per H; Hendee, William R; Herman, Michael G; Morse, Richard K; Mower, Herbert W; Pfeiffer, Douglas E; Root, William J; Sherouse, George W; Vossler, Matthew K; Wallace, Robert E; Walters, Barbara
2009-01-01
A comprehensive Code of Ethics for the members of the American Association of Physicists in Medicine (AAPM) is presented as the report of Task Group 109 which consolidates previous AAPM ethics policies into a unified document. The membership of the AAPM is increasingly diverse. Prior existing AAPM ethics polices were applicable specifically to medical physicists, and did not encompass other types of members such as health physicists, regulators, corporate affiliates, physicians, scientists, engineers, those in training, or other health care professionals. Prior AAPM ethics policies did not specifically address research, education, or business ethics. The Ethics Guidelines of this new Code of Ethics have four major sections: professional conduct, research ethics, education ethics, and business ethics. Some elements of each major section may be duplicated in other sections, so that readers interested in a particular aspect of the code do not need to read the entire document for all relevant information. The prior Complaint Procedure has also been incorporated into this Code of Ethics. This Code of Ethics (PP 24-A) replaces the following AAPM policies: Ethical Guidelines for Vacating a Position (PP 4-B); Ethical Guidelines for Reviewing the Work of Another Physicist (PP 5-C); Guidelines for Ethical Practice for Medical Physicists (PP 8-D); and Ethics Complaint Procedure (PP 21-A). The AAPM Board of Directors approved this Code or Ethics on July 31, 2008.
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.
Simulation of power maneuvering experiment of MASLWR test facility by MARS-KS code
Energy Technology Data Exchange (ETDEWEB)
Park, Ju Yeop [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)
2013-10-15
In the present study, KINS simulation result by the MARS-KS code (KS-002 version) for the SP-3 experiment is presented in detail and conclusion on MARS-KS code performance drawn through this simulation is described. Performance of the MARS-KS code is evaluated through the simulation of the power maneuvering experiment of the MASLWR test facility. Steady run shows the helical coil specific heat transfer model of the code is reasonable. However, identified discrepancy of the primary mass flowrate at transient run shows code performance for pressure drop needs to be improved considering sensitivity of the flowrate to the pressure drop at natural circulation. Since 2009, IAEA has conducted a research program entitled as ICSP (International Collaborative Standard Problem) on integral PWR design to evaluate current the state of the art of thermal-hydraulic code in simulating natural circulation flow within integral type reactor. In this ICSP, experimental data obtained from MASLWR (Multi-Application Small Light Water Reactor) test facility located at Oregon state university in the US have been simulated by various thermal-hydraulic codes of each participant of the ICSP and compared among others. MASLWR test facility is a mock-up of a passive integral type reactor equipped with helical coil steam generator. Since SMART reactor which is currently being developed in Korea also adopts a helical coil steam generator, Korea Institute of Nuclear Safety (KINS) has joined this ICSP to assess the applicability of a domestic regulatory audit thermal-hydraulic code (i. e. MARS-KS code) for the SMART reactor including wall-to-fluid heat transfer model modification based on independent international experiment data. In the ICSP, two types of transient experiments have been focused and they are loss of feedwater transient with subsequent ADS operation and long term cooling (SP-2) and normal operating conditions at different power levels (SP-3)
An approach for coupled-code multiphysics core simulations from a common input
International Nuclear Information System (INIS)
Schmidt, Rodney; Belcourt, Kenneth; Hooper, Russell; Pawlowski, Roger; Clarno, Kevin; Simunovic, Srdjan; Slattery, Stuart; Turner, John; Palmtag, Scott
2015-01-01
Highlights: • We describe an approach for coupled-code multiphysics reactor core simulations. • The approach can enable tight coupling of distinct physics codes with a common input. • Multi-code multiphysics coupling and parallel data transfer issues are explained. • The common input approach and how the information is processed is described. • Capabilities are demonstrated on an eigenvalue and power distribution calculation. - Abstract: This paper describes an approach for coupled-code multiphysics reactor core simulations that is being developed by the Virtual Environment for Reactor Applications (VERA) project in the Consortium for Advanced Simulation of Light-Water Reactors (CASL). In this approach a user creates a single problem description, called the “VERAIn” common input file, to define and setup the desired coupled-code reactor core simulation. A preprocessing step accepts the VERAIn file and generates a set of fully consistent input files for the different physics codes being coupled. The problem is then solved using a single-executable coupled-code simulation tool applicable to the problem, which is built using VERA infrastructure software tools and the set of physics codes required for the problem of interest. The approach is demonstrated by performing an eigenvalue and power distribution calculation of a typical three-dimensional 17 × 17 assembly with thermal–hydraulic and fuel temperature feedback. All neutronics aspects of the problem (cross-section calculation, neutron transport, power release) are solved using the Insilico code suite and are fully coupled to a thermal–hydraulic analysis calculated by the Cobra-TF (CTF) code. The single-executable coupled-code (Insilico-CTF) simulation tool is created using several VERA tools, including LIME (Lightweight Integrating Multiphysics Environment for coupling codes), DTK (Data Transfer Kit), Trilinos, and TriBITS. Parallel calculations are performed on the Titan supercomputer at Oak
Code modernization and modularization of APEX and SWAT watershed simulation models
SWAT (Soil and Water Assessment Tool) and APEX (Agricultural Policy / Environmental eXtender) are respectively large and small watershed simulation models derived from EPIC Environmental Policy Integrated Climate), a field-scale agroecology simulation model. All three models are coded in FORTRAN an...
Coding Instructions, Worksheets, and Keypunch Sheets for M.E.T.R.O.-APEX Simulation.
Michigan Univ., Ann Arbor. Environmental Simulation Lab.
Compiled in this resource are coding instructions, worksheets, and keypunch sheets for use in the M.E.T.R.O.-APEX simulation, described in detail in documents ED 064 530 through ED 064 550. Air Pollution Exercise (APEX) is a computerized college and professional level "real world" simulation of a community with urban and rural problems, industrial…
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
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 according to the New Romanian Civil Code
Directory of Open Access Journals (Sweden)
G. TIŢA-NICOLESCU
2012-01-01
Full Text Available Simulation arises when for two parties and for the same legal relationship there are two legal acts (more precisely two variants of the same legal act that have different contents, especially essential provisions. One of the two juridical acts is referred to as a public act or the apparent act (but, in fact, simulated, being stated as such by the parties and reflecting the ”official”( but false variant of the agreement between the parties, as it is, reached in front of the Notary, of the lawyer or a private signature act. The other mentioned act is the secret act (referred to as the secret agreement, which represents, in fact, the true agreement between the parties, but it is not included in an official act, being known only by the parties.
Process monitoring and simulation code verification using interactive computer animation
International Nuclear Information System (INIS)
Curtis, J.N.; Beelman, R.J.; Schwieder, D.H.; Stewart, H.D.
1984-01-01
At the Idaho National Engineering Laboratory (INEL), EG and G Idaho, Inc., has developed techniques by which schematics, created for and displayed at color graphics terminals, can be driven by actual or calculated data. These input data cause changes to occur within the displayed schematic. This research is presently being done to develop a prototype to be used in nuclear power plant control rooms. Work stations have already been developed to analyze data that are produced during actual and simulated nuclear reactor experiments
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)
Energy Technology Data Exchange (ETDEWEB)
Tavares, Desirée Yael de Sena; Silva, Adilson Costa da; Lima, Zelmo Rodrigues de, E-mail: zelmolima@yahoo.com.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)
2017-07-01
This work proposes to implement the cell calculation considering the fuel burning using the WIMSD-5B code. The cell calculation procedure allows to determine the nuclear parameters present in the multi-group neutron diffusion equation and for this purpose the neutron transport theory is used in a problem with dimensional reduction, but in contrast is considered a large number of groups associated with the neutron spectrum. There are a variety of reactor physics codes that determine the nuclear parameters by solving the neutron transport equation applied to an equivalent cell representing a fuel element. The WIMSD-5B code is a deterministic code that solves the transport equation using collision probability method. The simulation of fuel burning in the cell calculation took into account different nuclear data libraries. The WIMSD-5B code supports several nuclear data libraries and in the present work the following libraries were used: IAEA, ENDFB-VII.1, JENDL3.2, JEFF3.1 and JEF2.2, all formatted for 69 energy groups. (author)
International Nuclear Information System (INIS)
Tavares, Desirée Yael de Sena; Silva, Adilson Costa da; Lima, Zelmo Rodrigues de
2017-01-01
This work proposes to implement the cell calculation considering the fuel burning using the WIMSD-5B code. The cell calculation procedure allows to determine the nuclear parameters present in the multi-group neutron diffusion equation and for this purpose the neutron transport theory is used in a problem with dimensional reduction, but in contrast is considered a large number of groups associated with the neutron spectrum. There are a variety of reactor physics codes that determine the nuclear parameters by solving the neutron transport equation applied to an equivalent cell representing a fuel element. The WIMSD-5B code is a deterministic code that solves the transport equation using collision probability method. The simulation of fuel burning in the cell calculation took into account different nuclear data libraries. The WIMSD-5B code supports several nuclear data libraries and in the present work the following libraries were used: IAEA, ENDFB-VII.1, JENDL3.2, JEFF3.1 and JEF2.2, all formatted for 69 energy groups. (author)
Gamma irradiator dose mapping simulation using the MCNP code and benchmarking with dosimetry
International Nuclear Information System (INIS)
Sohrabpour, M.; Hassanzadeh, M.; Shahriari, M.; Sharifzadeh, M.
2002-01-01
The Monte Carlo transport code, MCNP, has been applied in simulating dose rate distribution in the IR-136 gamma irradiator system. Isodose curves, cumulative dose values, and system design data such as throughputs, over-dose-ratios, and efficiencies have been simulated as functions of product density. Simulated isodose curves, and cumulative dose values were compared with dosimetry values obtained using polymethyle-methacrylate, Fricke, ethanol-chlorobenzene, and potassium dichromate dosimeters. The produced system design data were also found to agree quite favorably with those of the system manufacturer's data. MCNP has thus been found to be an effective transport code for handling of various dose mapping excercises for gamma irradiators
Formulae for thermal feedback of group constants in digital reactor simulation
International Nuclear Information System (INIS)
Perneczky, L.; Toth, I.; Vigassy, J.
1976-01-01
The problem, how the feedback of the thermohydraulic field to the neutron density in a reactor can be calculated is analysed. After a brief survey of the digital models in reactor simulation the applied model based on the time-dependent two-group diffusion equations is described. Using the reactor physical code system THERESA numerical results for the VVER-440 reactor are presented. (Sz.Z.)
2D and 3D core-collapse supernovae simulation results obtained with the CHIMERA code
Energy Technology Data Exchange (ETDEWEB)
Bruenn, S W; Marronetti, P; Dirk, C J [Physics Department, Florida Atlantic University, 777 W. Glades Road, Boca Raton, FL 33431-0991 (United States); Mezzacappa, A; Hix, W R [Physics Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831-6354 (United States); Blondin, J M [Department of Physics, North Carolina State University, Raleigh, NC 27695-8202 (United States); Messer, O E B [Center for Computational Sciences, Oak Ridge National Laboratory, Oak Ridge, TN 37831-6354 (United States); Yoshida, S, E-mail: bruenn@fau.ed [Max-Planck-Institut fur Gravitationsphysik, Albert Einstein Institut, Golm (Germany)
2009-07-01
Much progress in realistic modeling of core-collapse supernovae has occurred recently through the availability of multi-teraflop machines and the increasing sophistication of supernova codes. These improvements are enabling simulations with enough realism that the explosion mechanism, long a mystery, may soon be delineated. We briefly describe the CHIMERA code, a supernova code we have developed to simulate core-collapse supernovae in 1, 2, and 3 spatial dimensions. We then describe the results of an ongoing suite of 2D simulations initiated from a 12, 15, 20, and 25 M{sub o-dot} progenitor. These have all exhibited explosions and are currently in the expanding phase with the shock at between 5,000 and 20,000 km. We also briefly describe an ongoing simulation in 3 spatial dimensions initiated from the 15 M{sub o-dot} progenitor.
Experience gained in running the EPRI MMS code with an in-house simulation language
International Nuclear Information System (INIS)
Weber, D.S.
1987-01-01
The EPRI Modular Modeling System (MMS) code represents a collection of component models and a steam/water properties package. This code has undergone extensive verification and validation testing. Currently, the code requires a commercially available simulation language to run. The Philadelphia Electric Company (PECO) has been modeling power plant systems for over the past sixteen years. As a result, an extensive number of models have been developed. In addition, an extensive amount of experience has been developed and gained using an in-house simulation language. The objective of this study was to explore the possibility of developing an MMS pre-processor which would allow the use of the MMS package with other simulation languages such as the PECO in-house simulation language
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.
MOCCA code for star cluster simulation: comparison with optical observations using COCOA
Askar, Abbas; Giersz, Mirek; Pych, Wojciech; Olech, Arkadiusz; Hypki, Arkadiusz
2016-02-01
We introduce and present preliminary results from COCOA (Cluster simulatiOn Comparison with ObservAtions) code for a star cluster after 12 Gyr of evolution simulated using the MOCCA code. The COCOA code is being developed to quickly compare results of numerical simulations of star clusters with observational data. We use COCOA to obtain parameters of the projected cluster model. For comparison, a FITS file of the projected cluster was provided to observers so that they could use their observational methods and techniques to obtain cluster parameters. The results show that the similarity of cluster parameters obtained through numerical simulations and observations depends significantly on the quality of observational data and photometric accuracy.
Comparisons of the simulation results using different codes for ADS spallation target
International Nuclear Information System (INIS)
Yu Hongwei; Fan Sheng; Shen Qingbiao; Zhao Zhixiang; Wan Junsheng
2002-01-01
The calculations to the standard thick target were made by using different codes. The simulation of the thick Pb target with length of 60 cm, diameter of 20 cm bombarded with 800, 1000, 1500 and 2000 MeV energetic proton beam was carried out. The yields and the spectra of emitted neutron were studied. The spallation target was simulated by SNSP, SHIELD, DCM/CEM (Dubna Cascade Model /Cascade Evaporation Mode) and LAHET codes. The Simulation Results were compared with experiments. The comparisons show good agreement between the experiments and the SNSP simulated leakage neutron yield. The SHIELD simulated leakage neutron spectra are in good agreement with the LAHET and the DCM/CEM simulated leakage neutron spectra
Benchmark test of drift-kinetic and gyrokinetic codes through neoclassical transport simulations
International Nuclear Information System (INIS)
Satake, S.; Sugama, H.; Watanabe, T.-H.; Idomura, Yasuhiro
2009-09-01
Two simulation codes that solve the drift-kinetic or gyrokinetic equation in toroidal plasmas are benchmarked by comparing the simulation results of neoclassical transport. The two codes are the drift-kinetic δf Monte Carlo code (FORTEC-3D) and the gyrokinetic full- f Vlasov code (GT5D), both of which solve radially-global, five-dimensional kinetic equation with including the linear Fokker-Planck collision operator. In a tokamak configuration, neoclassical radial heat flux and the force balance relation, which relates the parallel mean flow with radial electric field and temperature gradient, are compared between these two codes, and their results are also compared with the local neoclassical transport theory. It is found that the simulation results of the two codes coincide very well in a wide rage of plasma collisionality parameter ν * = 0.01 - 10 and also agree with the theoretical estimations. The time evolution of radial electric field and particle flux, and the radial profile of the geodesic acoustic mode frequency also coincide very well. These facts guarantee the capability of GT5D to simulate plasma turbulence transport with including proper neoclassical effects of collisional diffusion and equilibrium radial electric field. (author)
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.
Annual coded wire tag program (Washington) missing production groups: annual report for 1997; ANNUAL
International Nuclear Information System (INIS)
Byrne, J.; Fuss, H.; Ashbrook, C.
1998-01-01
The Bonneville Power Administration (BPA) funds the ''Annual Coded Wire Tag Program - Missing Production Groups for Columbia River Hatcheries'' project. The Washington Department of Fish and Wildlife (WDFW), Oregon Department of Fish and Wildlife (ODFW) and the United States Fish and Wildlife Service (USFWS) all operate salmon and steelhead rearing programs in the Columbia River basin. The intent of the funding is to coded-wire tag at least one production group of each species at each Columbia Basin hatchery to provide a holistic assessment of survival and catch distribution over time and to meet various measures of the Northwest Power Planning Councils (NWPPC) Fish and Wildlife Program. The WDFW project has three main objectives: (1) coded-wire tag at least one production group of each species at each Columbia Basin hatchery to enable evaluation of survival and catch distribution over time, (2) recover coded-wire tags from the snouts of fish tagged under objective 1 and estimate survival, contribution, and stray rates for each group, and (3) report the findings under objective 2 for all broods of chinook, and coho released from WDFW Columbia Basin hatcheries. Objective 1 for FY-97 was met with few modifications to the original FY-97 proposal. Under Objective 2, snouts containing coded-wire tags that were recovered during FY-97 were decoded. Under Objective 3, survival, contribution and stray rate estimates for the 1991-96 broods of chinook and 1993-96 broods of coho have not been made because recovery data for 1996-97 fisheries and escapement are preliminary. This report summarizes recovery information through 1995
International Nuclear Information System (INIS)
Kasselmann, S.; Druska, C.; Lauer, A.
2010-01-01
The energy spectra of fast and thermal neutrons from fission reactions in the FZJ code TINTE are modelled by two broad energy groups. Present demands for increased numerical accuracy led to the question of how precise the 2-group approximation is compared to a multi-group model. Therefore a new simulation program called MGT (Multi Group TINTE) has recently been developed which is able to handle up to 43 energy groups. Furthermore, an internal spectrum calculation for the determination of cross-sections can be performed for each time step and location within the reactor. In this study the multi-group energy models are compared to former calculations with only two energy groups. Different scenarios (normal operation and design-basis accidents) have been defined for a high temperature pebble bed reactor design with annular core. The effect of an increasing number of energy groups on safety-related parameters like the fuel and coolant temperature, the nuclear heat source or the xenon concentration is studied. It has been found that for the studied scenarios the use of up to 8 energy groups is a good trade-off between precision and a tolerable amount of computing time. (orig.)
Psacoin level 1A intercomparison probabilistic system assessment code (PSAC) user group
International Nuclear Information System (INIS)
Nies, A.; Laurens, J.M.; Galson, D.A.; Webster, S.
1990-01-01
This report describes an international code intercomparison exercise conducted by the NEA Probabilistic System Assessment Code (PSAC) User Group. The PSACOIN Level 1A exercise is the third of a series designed to contribute to the verification of probabilistic codes that may be used in assessing the safety of radioactive waste disposal systems or concepts. Level 1A is based on a more realistic system model than that used in the two previous exercises, and involves deep geological disposal concepts with a relatively complex structure of the repository vault. The report compares results and draws conclusions with regard to the use of different modelling approaches and the possible importance to safety of various processes within and around a deep geological repository. In particular, the relative significance of model uncertainty and data variability is discussed
Santos, José; Monteagudo, Angel
2011-02-21
As the canonical code is not universal, different theories about its origin and organization have appeared. The optimization or level of adaptation of the canonical genetic code was measured taking into account the harmful consequences resulting from point mutations leading to the replacement of one amino acid for another. There are two basic theories to measure the level of optimization: the statistical approach, which compares the canonical genetic code with many randomly generated alternative ones, and the engineering approach, which compares the canonical code with the best possible alternative. Here we used a genetic algorithm to search for better adapted hypothetical codes and as a method to guess the difficulty in finding such alternative codes, allowing to clearly situate the canonical code in the fitness landscape. This novel proposal of the use of evolutionary computing provides a new perspective in the open debate between the use of the statistical approach, which postulates that the genetic code conserves amino acid properties far better than expected from a random code, and the engineering approach, which tends to indicate that the canonical genetic code is still far from optimal. We used two models of hypothetical codes: one that reflects the known examples of codon reassignment and the model most used in the two approaches which reflects the current genetic code translation table. Although the standard code is far from a possible optimum considering both models, when the more realistic model of the codon reassignments was used, the evolutionary algorithm had more difficulty to overcome the efficiency of the canonical genetic code. Simulated evolution clearly reveals that the canonical genetic code is far from optimal regarding its optimization. Nevertheless, the efficiency of the canonical code increases when mistranslations are taken into account with the two models, as indicated by the fact that the best possible codes show the patterns of the
Directory of Open Access Journals (Sweden)
Monteagudo Ángel
2011-02-01
Full Text Available Abstract Background As the canonical code is not universal, different theories about its origin and organization have appeared. The optimization or level of adaptation of the canonical genetic code was measured taking into account the harmful consequences resulting from point mutations leading to the replacement of one amino acid for another. There are two basic theories to measure the level of optimization: the statistical approach, which compares the canonical genetic code with many randomly generated alternative ones, and the engineering approach, which compares the canonical code with the best possible alternative. Results Here we used a genetic algorithm to search for better adapted hypothetical codes and as a method to guess the difficulty in finding such alternative codes, allowing to clearly situate the canonical code in the fitness landscape. This novel proposal of the use of evolutionary computing provides a new perspective in the open debate between the use of the statistical approach, which postulates that the genetic code conserves amino acid properties far better than expected from a random code, and the engineering approach, which tends to indicate that the canonical genetic code is still far from optimal. We used two models of hypothetical codes: one that reflects the known examples of codon reassignment and the model most used in the two approaches which reflects the current genetic code translation table. Although the standard code is far from a possible optimum considering both models, when the more realistic model of the codon reassignments was used, the evolutionary algorithm had more difficulty to overcome the efficiency of the canonical genetic code. Conclusions Simulated evolution clearly reveals that the canonical genetic code is far from optimal regarding its optimization. Nevertheless, the efficiency of the canonical code increases when mistranslations are taken into account with the two models, as indicated by the
Timing group delay and differential code bias corrections for BeiDou positioning
Guo, Fei; Zhang, Xiaohong; Wang, Jinling
2015-05-01
This article first clearly figures out the relationship between parameters of timing group delay (TGD) and differential code bias (DCB) for BDS, and demonstrates the equivalence of TGD and DCB correction models combining theory with practice. The TGD/DCB correction models have been extended to various occasions for BDS positioning, and such models have been evaluated by real triple-frequency datasets. To test the effectiveness of broadcast TGDs in the navigation message and DCBs provided by the Multi-GNSS Experiment (MGEX), both standard point positioning (SPP) and precise point positioning (PPP) tests are carried out for BDS signals with different schemes. Furthermore, the influence of differential code biases on BDS positioning estimates such as coordinates, receiver clock biases, tropospheric delays and carrier phase ambiguities is investigated comprehensively. Comparative analysis show that the unmodeled differential code biases degrade the performance of BDS SPP by a factor of two or more, whereas the estimates of PPP are subject to varying degrees of influences. For SPP, the accuracy of dual-frequency combinations is slightly worse than that of single-frequency, and they are much more sensitive to the differential code biases, particularly for the B2B3 combination. For PPP, the uncorrected differential code biases are mostly absorbed into the receiver clock bias and carrier phase ambiguities and thus resulting in a much longer convergence time. Even though the influence of the differential code biases could be mitigated over time and comparable positioning accuracy could be achieved after convergence, it is suggested to properly handle with the differential code biases since it is vital for PPP convergence and integer ambiguity resolution.
A parallel code named NEPTUNE for 3D fully electromagnetic and pic simulations
International Nuclear Information System (INIS)
Dong Ye; Yang Wenyuan; Chen Jun; Zhao Qiang; Xia Fang; Ma Yan; Xiao Li; Sun Huifang; Chen Hong; Zhou Haijing; Mao Zeyao; Dong Zhiwei
2010-01-01
A parallel code named NEPTUNE for 3D fully electromagnetic and particle-in-cell (PIC) simulations is introduced, which could run on the Linux system with hundreds to thousand CPUs. NEPTUNE is suitable to simulate entire 3D HPM devices; many HPM devices are simulated and designed by using it. In NEPTUNE code, the electromagnetic fields are updated by using the finite-difference in time domain (FDTD) method of solving Maxwell equations and the particles are moved by using Buneman-Boris advance method of solving relativistic Newton-Lorentz equation. Electromagnetic fields and particles are coupled by using liner weighing interpolation PIC method, and the electric filed components are corrected by using Boris method of solve Poisson equation in order to ensure charge-conservation. NEPTUNE code could construct many complicated geometric structures, such as arbitrary axial-symmetric structures, plane transforming structures, slow-wave-structures, coupling holes, foils, and so on. The boundary conditions used in NEPTUNE code are introduced in brief, including perfectly electric conductor boundary, external wave boundary, and particle boundary. Finally, some typical HPM devices are simulated and test by using NEPTUNE code, including MILO, RBWO, VCO, and RKA. The simulation results are with correct and credible physical images, and the parallel efficiencies are also given. (authors)
The use of best estimate codes to improve the simulation in real time
International Nuclear Information System (INIS)
Rivero, N.; Esteban, J. A.; Lenhardt, G.
2007-01-01
Best estimate codes are assumed to be the technology solution providing the most realistic and accurate response. Best estimate technology provides a complementary solution to the conservative simulation technology usually applied to determine plant safety margins and perform security related studies. Tecnatom in the early 90's, within the MAS project, pioneered the initiative to implement best estimate code in its training simulators. Result of this project was the implementation of the first six-equations thermal hydraulic code worldwide (TRAC R T), running in a training environment. To meet real time and other specific training requirements, it was necessary to overcome important difficulties. Tecnatom has just adapted the Global Nuclear Fuel core Design code: PANAC 11, and is about to complete the General Electric TRACG04 thermal hydraulic code adaptation. This technology features a unique solution for nuclear plants aiming at providing the highest fidelity in simulation, enabling to consider the simulator as a multipurpose: engineering and training, simulation platform. Besides, a visual environment designed to optimize the models life cycle, covering both pre and post-processing activities, is in its late development phase. (Author)
Monte Carlo simulations on marker grouping and ordering.
Wu, J; Jenkins, J; Zhu, J; McCarty, J; Watson, C
2003-08-01
Four global algorithms, maximum likelihood (ML), sum of adjacent LOD score (SALOD), sum of adjacent recombinant fractions (SARF) and product of adjacent recombinant fraction (PARF), and one approximation algorithm, seriation (SER), were used to compare the marker ordering efficiencies for correctly given linkage groups based on doubled haploid (DH) populations. The Monte Carlo simulation results indicated the marker ordering powers for the five methods were almost identical. High correlation coefficients were greater than 0.99 between grouping power and ordering power, indicating that all these methods for marker ordering were reliable. Therefore, the main problem for linkage analysis was how to improve the grouping power. Since the SER approach provided the advantage of speed without losing ordering power, this approach was used for detailed simulations. For more generality, multiple linkage groups were employed, and population size, linkage cutoff criterion, marker spacing pattern (even or uneven), and marker spacing distance (close or loose) were considered for obtaining acceptable grouping powers. Simulation results indicated that the grouping power was related to population size, marker spacing distance, and cutoff criterion. Generally, a large population size provided higher grouping power than small population size, and closely linked markers provided higher grouping power than loosely linked markers. The cutoff criterion range for achieving acceptable grouping power and ordering power differed for varying cases; however, combining all situations in this study, a cutoff criterion ranging from 50 cM to 60 cM was recommended for achieving acceptable grouping power and ordering power for different cases.
Divergence-free MHD Simulations with the HERACLES Code
Directory of Open Access Journals (Sweden)
Vides J.
2013-12-01
Full Text Available Numerical simulations of the magnetohydrodynamics (MHD equations have played a significant role in plasma research over the years. The need of obtaining physical and stable solutions to these equations has led to the development of several schemes, all requiring to satisfy and preserve the divergence constraint of the magnetic field numerically. In this paper, we aim to show the importance of maintaining this constraint numerically. We investigate in particular the hyperbolic divergence cleaning technique applied to the ideal MHD equations on a collocated grid and compare it to the constrained transport technique that uses a staggered grid to maintain the property. The methods are implemented in the software HERACLES and several numerical tests are presented, where the robustness and accuracy of the different schemes can be directly compared.
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
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
Prototyping and Simulation of Robot Group Intelligence using Kohonen Networks.
Wang, Zhijun; Mirdamadi, Reza; Wang, Qing
2016-01-01
Intelligent agents such as robots can form ad hoc networks and replace human being in many dangerous scenarios such as a complicated disaster relief site. This project prototypes and builds a computer simulator to simulate robot kinetics, unsupervised learning using Kohonen networks, as well as group intelligence when an ad hoc network is formed. Each robot is modeled using an object with a simple set of attributes and methods that define its internal states and possible actions it may take under certain circumstances. As the result, simple, reliable, and affordable robots can be deployed to form the network. The simulator simulates a group of robots as an unsupervised learning unit and tests the learning results under scenarios with different complexities. The simulation results show that a group of robots could demonstrate highly collaborative behavior on a complex terrain. This study could potentially provide a software simulation platform for testing individual and group capability of robots before the design process and manufacturing of robots. Therefore, results of the project have the potential to reduce the cost and improve the efficiency of robot design and building.
Two-dimensional full-wave code for reflectometry simulations in TJ-II
International Nuclear Information System (INIS)
Blanco, E.; Heuraux, S.; Estrada, T.; Sanchez, J.; Cupido, L.
2004-01-01
A two-dimensional full-wave code in the extraordinary mode has been developed to simulate reflectometry in TJ-II. The code allows us to study the measurement capabilities of the future correlation reflectometer that is being installed in TJ-II. The code uses the finite-difference-time-domain technique to solve Maxwell's equations in the presence of density fluctuations. Boundary conditions are implemented by a perfectly matched layer to simulate free propagation. To assure the stability of the code, the current equations are solved by a fourth-order Runge-Kutta method. Density fluctuation parameters such as fluctuation level, wave numbers, and correlation lengths are extrapolated from those measured at the plasma edge using Langmuir probes. In addition, realistic plasma shape, density profile, magnetic configuration, and experimental setup of TJ-II are included to determine the plasma regimes in which accurate information may be obtained
Modification of PRETOR Code to Be Applied to Transport Simulation in Stellarators
International Nuclear Information System (INIS)
Fontanet, J.; Castejon, F.; Dies, J.; Fontdecaba, J.; Alejaldre, C.
2001-01-01
The 1.5 D transport code PRETOR, that has been previously used to simulate tokamak plasmas, has been modified to perform transport analysis in stellarator geometry. The main modifications that have been introduced in the code are related with the magnetic equilibrium and with the modelling of energy and particle transport. Therefore, PRETOR- Stellarator version has been achieved and the code is suitable to perform simulations on stellarator plasmas. As an example, PRETOR- Stellarator has been used in the transport analysis of several Heliac Flexible TJ-II shots, and the results are compared with those obtained using PROCTR code. These results are also compared with the obtained using the tokamak version of PRETOR to show the importance of the introduced changes. (Author) 18 refs
A simulation of driven reconnection by a high precision MHD code
International Nuclear Information System (INIS)
Kusano, Kanya; Ouchi, Yasuo; Hayashi, Takaya; Horiuchi, Ritoku; Watanabe, Kunihiko; Sato, Tetsuya.
1988-01-01
A high precision MHD code, which has the fourth-order accuracy for both the spatial and time steps, is developed, and is applied to the simulation studies of two dimensional driven reconnection. It is confirm that the numerical dissipation of this new scheme is much less than that of two-step Lax-Wendroff scheme. The effect of the plasma compressibility on the reconnection dynamics is investigated by means of this high precision code. (author)
Simulation of power maneuvering experiment of MASLWR test facility by MARS-KS code
Energy Technology Data Exchange (ETDEWEB)
Park, Ju Yeop [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)
2013-10-15
In this ICSP, experimental data obtained from MASLWR (Mulit-Application Small Light Water Reactor) test facility located at Oregon state university in the US have been simulated by various thermal-hydraulic codes of each participant of the ICSP and compared among others. MASLWR test facility is a mock-up of a passive integral type reactor equipped with helical coil steam generator. Since SMART reactor which is currently being developed in Korea also adopts a helical coil steam generator, Korea Institute of Nuclear Safety (KINS) has joined this ICSP to assess the applicability of a domestic regulatory audit thermal-hydraulic code (i. e. MARS-KS code) for the SMART reactor including wall-to-fluid heat transfer model modification based on independent international experiment data. In the ICSP, two types of transient experiments have been focused and they are 1) loss of feedwater transient with subsequent ADS operation and long term cooling (SP-2) and normal operating conditions at different power levels. In the present study, KINS simulation result by the MARS-KS code (KS-002 version) for the SP-3 experiment is presented in detail and conclusion on MARS-KS code performance drawn through this simulation is described. Performance of the MARS-KS code is evaluated through the simulation of the power maneuvering experiment of the MASLWR test facility. Steady run shows the helical coil specific heat transfer model of the code is reasonable. However, identified discrepancy of the primary mass flowrate at transient run shows code performance for pressure drop needs to be improved considering sensitivity of the flowrate to the pressure drop at natural circulation.
Energy Technology Data Exchange (ETDEWEB)
NONE
2001-01-01
In the report, research results discussed in 1999 fiscal year at Nuclear Code Evaluation Committee of Nuclear Code Research Committee were summarized. Present status of Monte Carlo simulation on nuclear energy study was described. Especially, besides of criticality, shielding and core analyses, present status of applications to risk and radiation damage analyses, high energy transport and nuclear theory calculations of Monte Carlo Method was described. The 18 papers are indexed individually. (J.P.N.)
The Monte-Carlo code DECAY to simulate the decay of baryon and meson resonances
International Nuclear Information System (INIS)
Haenssgen, K.; Ritter, S.
1983-01-01
The code DECAY simulates the decay of unpolarized baryon and meson resonances in the laboratory frame. DECAY treats some resonances among these all baryon resonances of the spin 3/2 + decuplet and all meson resonances of the spin 1 - nonet. A given resonance decays via two or three particle decay steps until all decay products are stable particles. Program summary and code description are given. (author)
Monte-Carlo code PARJET to simulate e+e--annihilation events via QCD jets
International Nuclear Information System (INIS)
Ritter, S.
1983-01-01
The Monte-Carlo code PARJET simulates exclusive hadronic final states produced in e + e - -annihilation via a virtual photon by two steps: (i) the fragmentation of the original quark-antiquark pair into further partons using results of perturbative QCD in the leading logarithmic approximation (LLA), and (ii) the transition of these parton jets into hadrons on the basis of a chain decay model. Program summary and code description are given. (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
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
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)
G4-STORK: A Geant4-based Monte Carlo reactor kinetics simulation code
International Nuclear Information System (INIS)
Russell, Liam; Buijs, Adriaan; Jonkmans, Guy
2014-01-01
Highlights: • G4-STORK is a new, time-dependent, Monte Carlo code for reactor physics applications. • G4-STORK was built by adapting and expanding on the Geant4 Monte Carlo toolkit. • G4-STORK was designed to simulate short-term fluctuations in reactor cores. • G4-STORK is well suited for simulating sub- and supercritical assemblies. • G4-STORK was verified through comparisons with DRAGON and MCNP. - Abstract: In this paper we introduce G4-STORK (Geant4 STOchastic Reactor Kinetics), a new, time-dependent, Monte Carlo particle tracking code for reactor physics applications. G4-STORK was built by adapting and expanding on the Geant4 Monte Carlo toolkit. The toolkit provides the fundamental physics models and particle tracking algorithms that track each particle in space and time. It is a framework for further development (e.g. for projects such as G4-STORK). G4-STORK derives reactor physics parameters (e.g. k eff ) from the continuous evolution of a population of neutrons in space and time in the given simulation geometry. In this paper we detail the major additions to the Geant4 toolkit that were necessary to create G4-STORK. These include a renormalization process that maintains a manageable number of neutrons in the simulation even in very sub- or supercritical systems, scoring processes (e.g. recording fission locations, total neutrons produced and lost, etc.) that allow G4-STORK to calculate the reactor physics parameters, and dynamic simulation geometries that can change over the course of simulation to illicit reactor kinetics responses (e.g. fuel temperature reactivity feedback). The additions are verified through simple simulations and code-to-code comparisons with established reactor physics codes such as DRAGON and MCNP. Additionally, G4-STORK was developed to run a single simulation in parallel over many processors using MPI (Message Passing Interface) pipes
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...
The cultural formation code of successfulness verticals of the U.S. ethnic groups
Directory of Open Access Journals (Sweden)
Liudmyla Petrashko
2010-11-01
Full Text Available In the article there are outlined the prospects of global economic development. There was built an evolutional model of theoretical studies of the phenomenon “culture” in the context of universal, system and value approaches. It gives the brief characteristics of the cultural assimilation model “melting crucible”. There have been determined the indicators of the successfulness verticals of the U.S. ethnic groups and made their assessment. By virtue of the author’s method is given the assessment of the comparative significance of the heterogeneous cultural codes of maternal (immigration and hosting environment of the USA, which gave the possibility to determine the factors that ensure the economic success of the American ethnic groups. The results of the research provide reasoning for the change of traditional vector of the cultures’ typology and confirm the existence of the progressive cultural codes.
International Nuclear Information System (INIS)
Weiss, D.E.; Kalweit, H.W.; Kensek, R.P.
1994-01-01
A simple multilayer slab model of an electron beam using the ITS/TIGER code can consistently account for about 80% of the actual dose delivered by a low voltage electron beam. The difference in calculated values is principally due to the 3D hibachi structure which blocks 22% of the beam. A 3D model was constructed using the ITS/ACCEPT code to improve upon the TIGER simulations. A rectangular source description update to the code and reproduction of all key geometric elements involved, including the hibachi, accounted for 90-95% of the dose received by routine dosimetry
Energy Technology Data Exchange (ETDEWEB)
Picard, Richard Roy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bhat, Kabekode Ghanasham [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2017-07-18
We examine sensitivity analysis and uncertainty quantification for molecular dynamics simulation. Extreme (large or small) output values for the LAMMPS code often occur at the boundaries of input regions, and uncertainties in those boundary values are overlooked by common SA methods. Similarly, input values for which code outputs are consistent with calibration data can also occur near boundaries. Upon applying approaches in the literature for imprecise probabilities (IPs), much more realistic results are obtained than for the complacent application of standard SA and code calibration.
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
International Nuclear Information System (INIS)
Takase, Haruhiko; Senda, Ikuo
1999-01-01
A Toroidally Symmetric Plasma Simulation (TSPS) code has been developed for investigating the position and shape control on tokamak plasmas. The analyses of three-dimensional eddy currents on the conducting components around the plasma and the two-dimensional magneto-hydrodynamic (MHD) equilibrium are taken into account in this code. The code can analyze the plasma position and shape control during the minor disruption in which the deformation of plasma is not negligible. Using the ITER (International Thermonuclear Experimental Reactor) parameters, some examples of calculations are shown in this paper. (author)
The assessment of containment codes by experiments simulating severe accident scenarios
International Nuclear Information System (INIS)
Karwat, H.
1992-01-01
Hitherto, a generally applicable validation matrix for codes simulating the containment behaviour under severe accident conditions did not exist. Past code applications have shown that most problems may be traced back to inaccurate thermalhydraulic parameters governing gas- or aerosol-distribution events. A provisional code-validation matrix is proposed, based on a careful selection of containment experiments performed during recent years in relevant test facilities under various operating conditions. The matrix focuses on the thermalhydraulic aspects of the containment behaviour after severe accidents as a first important step. It may be supplemented in the future by additional suitable tests
Simulation of TRIGA Mark II Benchmark Experiment using WIMSD4 and CITATION codes
International Nuclear Information System (INIS)
Dalle, Hugo Moura; Pereira, Claubia
2000-01-01
This paper presents a simulation of the TRIGA Mark II Benchmark Experiment, Part I: Steady-State Operation and is part of the calculation methodology validation developed to the neutronic calculation of the CDTN's TRIGA IPR - R1 reactor. A version of the WIMSD4, obtained in the Centro de Tecnologia Nuclear, in Cuba, was used in the cells calculation. In the core calculations was adopted the diffusion code CITATION. Was adopted a 3D representation of the core and the calculations were carried out at two energy groups. Many of the experiments were simulated, including, K eff , control rods reactivity worth, fuel elements reactivity worth distribution and the fuel temperature reactivity coefficient. The comparison of the obtained results, with the experimental results, shows differences in the range of the accuracy of the measurements, to the control rods worth and fuel temperature reactivity coefficient, or on an acceptable range, following the literature, to the K eff and fuel elements reactivity worth distribution and the fuel temperature reactivity coefficient. The comparison of the obtained results, with the experimental. results, shows differences in the range of the accuracy of the measurements, to the control rods worth and fuel temperature reactivity coefficient, or in an acceptable range, following the literature, to the K eff and fuel elements reactivity worth distribution. (author)
Calculation of anisotropic few-group constants in asymptotic cells: the code ANICELL
International Nuclear Information System (INIS)
Devenyi, A.
1985-10-01
The theoretical background of the ANICELL computer program together with a user's manual is presented. ANICELL is a nuclear reactor neutron transport code which solves the traditional asymptotic and the so-called tilted flux transport problems in one-dimensional cylindrical geometry using linearly anisotropic scattering. The method of solution used is the first flight collision probability technique. Few-group constants including radial and axial diffusion coefficients for the cell are also prepared by the program. (author)
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)
Yuan, Jian-guo; Zhou, Guang-xiang; Gao, Wen-chun; Wang, Yong; Lin, Jin-zhao; Pang, Yu
2016-01-01
According to the requirements of the increasing development for optical transmission systems, a novel construction method of quasi-cyclic low-density parity-check (QC-LDPC) codes based on the subgroup of the finite field multiplicative group is proposed. Furthermore, this construction method can effectively avoid the girth-4 phenomena and has the advantages such as simpler construction, easier implementation, lower encoding/decoding complexity, better girth properties and more flexible adjustment for the code length and code rate. The simulation results show that the error correction performance of the QC-LDPC(3 780,3 540) code with the code rate of 93.7% constructed by this proposed method is excellent, its net coding gain is respectively 0.3 dB, 0.55 dB, 1.4 dB and 1.98 dB higher than those of the QC-LDPC(5 334,4 962) code constructed by the method based on the inverse element characteristics in the finite field multiplicative group, the SCG-LDPC(3 969,3 720) code constructed by the systematically constructed Gallager (SCG) random construction method, the LDPC(32 640,30 592) code in ITU-T G.975.1 and the classic RS(255,239) code which is widely used in optical transmission systems in ITU-T G.975 at the bit error rate ( BER) of 10-7. Therefore, the constructed QC-LDPC(3 780,3 540) code is more suitable for optical transmission systems.
Harmonization of nuclear codes and standards, pacific nuclear council working and task group report
International Nuclear Information System (INIS)
Dua, S.S.
2006-01-01
Full text: The codes and standards, both at the national and international level, have had a major impact on the industry worldwide and served it well in maintaining the performance and safety of the nuclear reactors and facilities. The codes and standards, in general, are consensus documents and do seek public input at various levels before they are finalized and rolled out for use by the nuclear vendors, consultants, utilities and regulatory bodies. However, the extensive development of prescriptive national standards if unchecked against the global environment and trade agreements (NAFTA, WTO, etc.) can also become barriers and cause difficulties to compete in the world market. During the last decade, the national and international writing standards writing bodies have recognized these issues and are moving more towards the rationalization and harmonization of their standards with the more widely accepted generic standards. The Pacific Nuclear Council (PNC) recognized the need for harmonization of the nuclear codes and standards for its member countries and formed a Task Group to achieve its objectives. The Task Group has a number of members from the PNC member countries. In 2005 PNC further raised the importance of this activity and formed a Working Group to cover a broader scope. The Working Group (WG) mandate is to identify and analyze the different codes and standards introduced to the Pacific Basin region, in order to achieve mutual understanding, harmonization and application in each country. This o requires the WG to develop and encourage the use of reasonably consistent criteria for the design and development, engineering, procurement, fabrication, construction, testing, operations, maintenance, waste management, decommissioning and the management of the commercial nuclear power plants in the Pacific Basin so as to: Promote consistent safety, quality, environmental and management standards for nuclear energy and other peaceful applications of nuclear
Implementing particle-in-cell plasma simulation code on the BBN TC2000
International Nuclear Information System (INIS)
Sturtevant, J.E.; Maccabe, A.B.
1990-01-01
The BBN TC2000 is a multiple instruction, multiple data (MIMD) machine that combines a physically distributed memory with a logically shared memory programming environment using the unique Butterfly switch. Particle-In-Cell (PIC) plasma simulations model the interaction of charged particles with electric and magnetic fields. This paper describes the implementation of both a 1-D electrostatic and a 2 1/2-D electromagnetic PIC (particle-in-cell) plasma simulation code on a BBN TC2000. Performance is compared to implementations of the same code on the shared memory Sequent Balance and distributed memory Intel iPSC hypercube
Interactive fluka: a world wide web version for a simulation code in proton therapy
International Nuclear Information System (INIS)
Garelli, S.; Giordano, S.; Piemontese, G.; Squarcia, S.
1998-01-01
We considered the possibility of using the simulation code FLUKA, in the framework of TERA. We provided a window under World Wide Web in which an interactive version of the code is available. The user can find instructions for the installation, an on-line FLUKA manual and interactive windows for inserting all the data required by the configuration running file in a very simple way. The database choice allows a more versatile use for data verification and update, recall of old simulations and comparison with selected examples. A completely new tool for geometry drawing under Java has also been developed. (authors)
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
Energy Technology Data Exchange (ETDEWEB)
Patra, Anirban [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tome, Carlos [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2016-03-23
This Milestone report shows good progress in interfacing VPSC with the FE codes ABAQUS and MOOSE, to perform component-level simulations of irradiation-induced deformation in Zirconium alloys. In this preliminary application, we have performed an irradiation growth simulation in the quarter geometry of a cladding tube. We have benchmarked VPSC-ABAQUS and VPSC-MOOSE predictions with VPSC-SA predictions to verify the accuracy of the VPSCFE interface. Predictions from the FE simulations are in general agreement with VPSC-SA simulations and also with experimental trends.
Modelling of Ohmic discharges in ADITYA tokamak using the Tokamak Simulation Code
International Nuclear Information System (INIS)
Bandyopadhyay, I; Ahmed, S M; Atrey, P K; Bhatt, S B; Bhattacharya, R; Chaudhury, M B; Deshpande, S P; Gupta, C N; Jha, R; Joisa, Y Shankar; Kumar, Vinay; Manchanda, R; Raju, D; Rao, C V S; Vasu, P
2004-01-01
Several Ohmic discharges of the ADITYA tokamak are simulated using the Tokamak Simulation Code (TSC), similar to that done earlier for the TFTR tokamak. Unlike TFTR, the dominant radiation process in ADITYA is through impurity line radiation. TSC can follow the experimental plasma current and position to very good accuracy. The thermal transport model of TSC including impurity line radiation gives a good match of the simulated results with experimental data for the Ohmic flux consumption, electron temperature and Z eff . Even the simulated magnetic probe signals are in reasonably good agreement with the experimental values
Modelling of Ohmic discharges in ADITYA tokamak using the Tokamak Simulation Code
Energy Technology Data Exchange (ETDEWEB)
Bandyopadhyay, I; Ahmed, S M; Atrey, P K; Bhatt, S B; Bhattacharya, R; Chaudhury, M B; Deshpande, S P; Gupta, C N; Jha, R; Joisa, Y Shankar; Kumar, Vinay; Manchanda, R; Raju, D; Rao, C V S; Vasu, P [Institute for Plasma Research, Bhat, Gandhinagar 382428 (India)
2004-09-01
Several Ohmic discharges of the ADITYA tokamak are simulated using the Tokamak Simulation Code (TSC), similar to that done earlier for the TFTR tokamak. Unlike TFTR, the dominant radiation process in ADITYA is through impurity line radiation. TSC can follow the experimental plasma current and position to very good accuracy. The thermal transport model of TSC including impurity line radiation gives a good match of the simulated results with experimental data for the Ohmic flux consumption, electron temperature and Z{sub eff}. Even the simulated magnetic probe signals are in reasonably good agreement with the experimental values.
Energy Technology Data Exchange (ETDEWEB)
Liu, Z. X., E-mail: zxliu316@ipp.ac.cn; Xia, T. Y.; Liu, S. C.; Ding, S. Y. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Xu, X. Q.; Joseph, I.; Meyer, W. H. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Gao, X.; Xu, G. S.; Shao, L. M.; Li, G. Q.; Li, J. G. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)
2014-09-15
Experimental measurements of edge localized modes (ELMs) observed on the EAST experiment are compared to linear and nonlinear theoretical simulations of peeling-ballooning modes using the BOUT++ code. Simulations predict that the dominant toroidal mode number of the ELM instability becomes larger for lower current, which is consistent with the mode structure captured with visible light using an optical CCD camera. The poloidal mode number of the simulated pressure perturbation shows good agreement with the filamentary structure observed by the camera. The nonlinear simulation is also consistent with the experimentally measured energy loss during an ELM crash and with the radial speed of ELM effluxes measured using a gas puffing imaging diagnostic.
International Nuclear Information System (INIS)
Serfontein, Dawid E.; Mulder, Eben J.; Reitsma, Frederik
2014-01-01
A computer code was developed for the semi-automatic translation of input models for the VSOP-A diffusion neutronics simulation code to the format of the newer VSOP 99/05 code. In this paper, this algorithm is presented as a generic method for producing codes for the automatic translation of input models from the format of one code version to another, or even to that of a completely different code. Normally, such translations are done manually. However, input model files, such as for the VSOP codes, often are very large and may consist of many thousands of numeric entries that make no particular sense to the human eye. Therefore the task, of for instance nuclear regulators, to verify the accuracy of such translated files can be very difficult and cumbersome. This may cause translation errors not to be picked up, which may have disastrous consequences later on when a reactor with such a faulty design is built. Therefore a generic algorithm for producing such automatic translation codes may ease the translation and verification process to a great extent. It will also remove human error from the process, which may significantly enhance the accuracy and reliability of the process. The developed algorithm also automatically creates a verification log file which permanently record the names and values of each variable used, as well as the list of meanings of all the possible values. This should greatly facilitate reactor licensing applications
Energy Technology Data Exchange (ETDEWEB)
Serfontein, Dawid E., E-mail: Dawid.Serfontein@nwu.ac.za [School of Mechanical and Nuclear Engineering, North West University (PUK-Campus), PRIVATE BAG X6001 (Internal Post Box 360), Potchefstroom 2520 (South Africa); Mulder, Eben J. [School of Mechanical and Nuclear Engineering, North West University (South Africa); Reitsma, Frederik [Calvera Consultants (South Africa)
2014-05-01
A computer code was developed for the semi-automatic translation of input models for the VSOP-A diffusion neutronics simulation code to the format of the newer VSOP 99/05 code. In this paper, this algorithm is presented as a generic method for producing codes for the automatic translation of input models from the format of one code version to another, or even to that of a completely different code. Normally, such translations are done manually. However, input model files, such as for the VSOP codes, often are very large and may consist of many thousands of numeric entries that make no particular sense to the human eye. Therefore the task, of for instance nuclear regulators, to verify the accuracy of such translated files can be very difficult and cumbersome. This may cause translation errors not to be picked up, which may have disastrous consequences later on when a reactor with such a faulty design is built. Therefore a generic algorithm for producing such automatic translation codes may ease the translation and verification process to a great extent. It will also remove human error from the process, which may significantly enhance the accuracy and reliability of the process. The developed algorithm also automatically creates a verification log file which permanently record the names and values of each variable used, as well as the list of meanings of all the possible values. This should greatly facilitate reactor licensing applications.
Inclusion of models to describe severe accident conditions in the fuel simulation code DIONISIO
Energy Technology Data Exchange (ETDEWEB)
Lemes, Martín; Soba, Alejandro [Sección Códigos y Modelos, Gerencia Ciclo del Combustible Nuclear, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina); Daverio, Hernando [Gerencia Reactores y Centrales Nucleares, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina); Denis, Alicia [Sección Códigos y Modelos, Gerencia Ciclo del Combustible Nuclear, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina)
2017-04-15
The simulation of fuel rod behavior is a complex task that demands not only accurate models to describe the numerous phenomena occurring in the pellet, cladding and internal rod atmosphere but also an adequate interconnection between them. In the last years several models have been incorporated to the DIONISIO code with the purpose of increasing its precision and reliability. After the regrettable events at Fukushima, the need for codes capable of simulating nuclear fuels under accident conditions has come forth. Heat removal occurs in a quite different way than during normal operation and this fact determines a completely new set of conditions for the fuel materials. A detailed description of the different regimes the coolant may exhibit in such a wide variety of scenarios requires a thermal-hydraulic formulation not suitable to be included in a fuel performance code. Moreover, there exist a number of reliable and famous codes that perform this task. Nevertheless, and keeping in mind the purpose of building a code focused on the fuel behavior, a subroutine was developed for the DIONISIO code that performs a simplified analysis of the coolant in a PWR, restricted to the more representative situations and provides to the fuel simulation the boundary conditions necessary to reproduce accidental situations. In the present work this subroutine is described and the results of different comparisons with experimental data and with thermal-hydraulic codes are offered. It is verified that, in spite of its comparative simplicity, the predictions of this module of DIONISIO do not differ significantly from those of the specific, complex codes.
Monte Carlo simulation of a coded-aperture thermal neutron camera
International Nuclear Information System (INIS)
Dioszegi, I.; Salwen, C.; Forman, L.
2011-01-01
We employed the MCNPX Monte Carlo code to simulate image formation in a coded-aperture thermal-neutron camera. The camera, developed at Brookhaven National Laboratory (BNL), consists of a 20 x 17 cm"2 active area "3He-filled position-sensitive wire chamber in a cadmium enclosure box. The front of the box is a coded-aperture cadmium mask (at present with three different resolutions). We tested the detector experimentally with various arrangements of moderated point-neutron sources. The purpose of using the Monte Carlo modeling was to develop an easily modifiable model of the device to predict the detector's behavior using different mask patterns, and also to generate images of extended-area sources or large numbers (up to ten) of them, that is important for nonproliferation and arms-control verification, but difficult to achieve experimentally. In the model, we utilized the advanced geometry capabilities of the MCNPX code to simulate the coded aperture mask. Furthermore, the code simulated the production of thermal neutrons from fission sources surrounded by a thermalizer. With this code we also determined the thermal-neutron shadow cast by the cadmium mask; the calculations encompassed fast- and epithermal-neutrons penetrating into the detector through the mask. Since the process of signal production in "3He-filled position-sensitive wire chambers is well known, we omitted this part from our modeling. Simplified efficiency values were used for the three (thermal, epithermal, and fast) neutron-energy regions. Electronic noise and the room's background were included as a uniform irradiation component. We processed the experimental- and simulated-images using identical LabVIEW virtual instruments. (author)
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.
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)
Directory of Open Access Journals (Sweden)
Hongwei ZHAO
2014-09-01
Full Text Available In this paper, the capacity of the BICM system over AWGN channels is first analyzed; the curves of BICM capacity versus SNR are also got by the Monte-Carlo simulations===?=== and compared with the curves of the CM capacity. Based on the analysis results, we simulate the error performances of BICM system with LDPC codes. Simulation results show that the capacity of BICM system with LDPC codes is enormously influenced by the mapping methods. Given a certain modulation method, the BICM system can obtain about 2-3 dB gain with Gray mapping compared with Non-Gray mapping. Meanwhile, the simulation results also demonstrate the correctness of the theory analysis.
A FEW ASPECTS REGARDING THE SIMULATION OF CONTRACT IN THE ROMANIAN CIVIL CODE
Directory of Open Access Journals (Sweden)
Tudor Vlad RĂDULESCU
2017-05-01
Full Text Available The article aims to analyze some key aspects of simulation in contracts, as regulated by the Romanian Civil Code . The process of simulation will be explained, based on the provisions of the previous Civil Code, but also with reference to the relevant provisions of the legislation of some European countries. The analyse will focus on the apparent act, and also on the secret one and a special emphasis on intention to simulate, animo simulandi, the key aspect of the matter. Also the effects of the simulation will be reviewed, both from the point of view of the parties and that of third parties, the concept of third parties having another meaning in this procedure.
MCMEG: Simulations of both PDD and TPR for 6 MV LINAC photon beam using different MC codes
Fonseca, T. C. F.; Mendes, B. M.; Lacerda, M. A. S.; Silva, L. A. C.; Paixão, L.; Bastos, F. M.; Ramirez, J. V.; Junior, J. P. R.
2017-11-01
The Monte Carlo Modelling Expert Group (MCMEG) is an expert network specializing in Monte Carlo radiation transport and the modelling and simulation applied to the radiation protection and dosimetry research field. For the first inter-comparison task the group launched an exercise to model and simulate a 6 MV LINAC photon beam using the Monte Carlo codes available within their laboratories and validate their simulated results by comparing them with experimental measurements carried out in the National Cancer Institute (INCA) in Rio de Janeiro, Brazil. The experimental measurements were performed using an ionization chamber with calibration traceable to a Secondary Standard Dosimetry Laboratory (SSDL). The detector was immersed in a water phantom at different depths and was irradiated with a radiation field size of 10×10 cm2. This exposure setup was used to determine the dosimetric parameters Percentage Depth Dose (PDD) and Tissue Phantom Ratio (TPR). The validation process compares the MC calculated results to the experimental measured PDD20,10 and TPR20,10. Simulations were performed reproducing the experimental TPR20,10 quality index which provides a satisfactory description of both the PDD curve and the transverse profiles at the two depths measured. This paper reports in detail the modelling process using MCNPx, MCNP6, EGSnrc and Penelope Monte Carlo codes, the source and tally descriptions, the validation processes and the results.
Energy Technology Data Exchange (ETDEWEB)
Ud-Din Khan, Salah [Chinese Academy of Sciences, Hefei (China). Inst. of Plasma Physics; King Saud Univ., Riyadh (Saudi Arabia). Sustainable Energy Technologies Center; Peng, Minjun [Harbin Engineering Univ. (China). College of Nuclear Science and Technology; Yuntao, Song; Ud-Din Khan, Shahab [Chinese Academy of Sciences, Hefei (China). Inst. of Plasma Physics; Haider, Sajjad [King Saud Univ., Riyadh (Saudi Arabia). Sustainable Energy Technologies Center
2017-02-15
The objective is to analyze the safety of small modular nuclear reactors of 220 MWe power. Reactivity initiated accidents (RIA) were investigated by neutron kinetic/thermal hydraulic (NK/TH) coupling approach and thermal hydraulic code i.e., RELAP5. The results obtained by these approaches were compared for validation and accuracy of simulation. In the NK/TH coupling technique, three codes (HELIOS, REMARK, THEATRe) were used. These codes calculate different parameters of the reactor core (fission power, reactivity, fuel temperature and inlet/outlet temperatures). The data exchanges between the codes were assessed by running the codes simultaneously. The results obtained from both (NK/TH coupling) and RELAP5 code analyses complement each other, hence confirming the accuracy of simulation.
Accurate simulation of ionisation chamber response with the Monte Carlo code PENELOPE
International Nuclear Information System (INIS)
Sempau, Josep; Andreo, Pedro
2011-01-01
Ionisation chambers (IC) are routinely used in hospitals for the dosimetry of the photon and electron beams used for radiotherapy treatments. The determination of absorbed dose to water from the absorbed dose to the air filling the cavity requires the introduction of stopping power ratios and perturbation factors, which account for the disturbance caused by the presence of the chamber. Although this may seem a problem readily amenable to Monte Carlo simulation, the fact is that the accurate determination of IC response has been, for various decades, one of the most important challenges of the simulation of electromagnetic showers. The main difficulty stems from the use of condensed history techniques for electron and positron transport. This approach, which involves grouping a large number of interactions into a single artificial event, is known to produce the so-called interface effects when particles travel across surfaces separating different media. These effects can be sizeable when the electron step length is not negligible compared to the size of the region being crossed, as it is the case with the cavity of an IC. The artefact, which becomes apparent when the chamber response shows a marked dependence on the adopted step size, can be palliated with the use of sophisticated electron transport algorithms. These topics are discussed in the context of the transport model implemented in the PENELOPE code. The degree of violation of the Fano theorem for a simple, planar geometry, is used as a measure of the stability of the algorithm with respect to variations of the electron step length, thus assessing the 'quality' of its condensed history scheme. It is shown that, with a suitable choice of transport parameters, PENELOPE simulates IC response with an accuracy of the order of 0.1%.
Accurate simulation of ionization chamber response with the Monte Carlo code PENELOPE
International Nuclear Information System (INIS)
Sempau, Josep
2010-01-01
Full text. Ionization chambers (IC) are routinely used in hospitals for the dosimetry of the photon and electron beams used for radiotherapy treatments. The determination of absorbed dose to water from the absorbed dose to the air filling the cavity requires the introduction of stopping power ratios and perturbation factors, which account for the disturbance caused by the presence of the chamber. Although this may seem a problem readily amenable to Monte Carlo simulation, the fact is that the accurate determination of IC response has been, during the last 20 years, one of the most important challenges of the simulation of electromagnetic showers. The main difficulty stems from the use of condensed history techniques for electron and positron transport. This approach, which involves grouping a large number of interactions into a single artificial event, is known to produce the so-called interface effects when particles travel across surfaces separating different media. These effects are extremely important when the electron step length is not negligible compared to the size of the region being crossed, as it is the case with the cavity of an IC. The artifact, which becomes apparent when the chamber response shows a marked dependence on the adopted step size, can be palliated with the use of sophisticated electron transport algorithms. These topics will be discussed in the context of the transport model implemented in the Penelope code. The degree of violation of the Fano theorem for a simple, planar geometry, will be used as a measure of the stability of the algorithm with respect to variations of the electron step length, thus assessing the 'quality' of its condensed history scheme. It will be shown that, with a suitable choice of transport parameters, Penelope can simulate IC response with an accuracy of the order of 0.1%. (author)
TIMS-1: a processing code for production of group constants of heavy resonant nuclei
International Nuclear Information System (INIS)
Takano, Hideki; Ishiguro, Yukio; Matsui, Yasushi.
1980-09-01
The TIMS-1 code calculates the infinitely dilute group cross sections and the temperature dependent self-shielding factors for arbitrary values of σ 0 and R, where σ 0 is the effective background cross section of potential scattering and R the ratio of the atomic number densities for two resonant nuclei if any. This code is specifically programmed to use the evaluated nuclear data file of ENDF/B or JENDL as input data. In the unresolved resonance region, the resonance parameters and the level spacings are generated by using Monte Carlo method from the Porter-Thomas and Wigner distributions respectively. The Doppler broadened cross sections are calculated on the ultra-fine lethargy meshes of about 10 -3 -- 10 -5 using the generated and resolved resonance parameters. The effective group constants are calculated by solving the neutron slowing down equation with the use of the recurrence formula for the neutron slowing down source. The output of the calculated results is given in a format being consistent with the JAERI-Fast set (JFS) or the Standard Reactor Analysis Code (SRAC) library. Both FACOM 230/75 and M200 versions of TIMS-1 are available. (author)
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
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
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 codes to evcaluate dose conversion coefficients for hadrons over 10 GeV
International Nuclear Information System (INIS)
Sato, T.; Tsuda, S.; Sakamoto, Y.; Yamaguchi, Y.; Niita, K.
2002-01-01
The conversion coefficients from fluence to effective dose for high energy hadrons are indispensable for various purposes such as accelerator shielding design and dose evaluation in space mission. Monte Carlo calculation code HETC-3STEP was used to evaluate dose conversion coefficients for neutrons and protons up to 10 GeV with an anthropomorphic model. The scaling model was incorporated in the code for simulation of high energy nuclear reactions. However, the secondary particle energy spectra predicted by the model were not smooth for nuclear reactions over several GeV. We attempted, therefore, to simulate transportation of such high energy particles by two newly developed Monte Carlo simulation codes: one is HETC-3STEP including the model used in EVENTQ instead of the scaling model, and the other is NMTC/JAM. By comparing calculated cross sections by these codes with experimental data for high energy nuclear reactions, it was found that NMTC/JAM had a better agreement with the data. We decided, therefore, to adopt NMTC/JAM for evaluation of dose conversion coefficients for hadrons with energies over 10 GeV. The effective dose conversion coefficients for high energy neutrons and protons evaluated by NMTC/JAM were found to be close to those by the FLUKA code
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
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
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.
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
Virtual Gaming Simulation in Nursing Education: A Focus Group Study.
Verkuyl, Margaret; Hughes, Michelle; Tsui, Joyce; Betts, Lorraine; St-Amant, Oona; Lapum, Jennifer L
2017-05-01
The use of serious gaming in a virtual world is a novel pedagogical approach in nursing education. A virtual gaming simulation was implemented in a health assessment class that focused on mental health and interpersonal violence. The study's purpose was to explore students' experiences of the virtual gaming simulation. Three focus groups were conducted with a convenience sample of 20 first-year nursing students after they completed the virtual gaming simulation. Analysis yielded five themes: (a) Experiential Learning, (b) The Learning Process, (c) Personal Versus Professional, (d) Self-Efficacy, and (e) Knowledge. Virtual gaming simulation can provide experiential learning opportunities that promote engagement and allow learners to acquire and apply new knowledge while practicing skills in a safe and realistic environment. [J Nurs Educ. 2017;56(5):274-280.]. Copyright 2017, SLACK Incorporated.
77 FR 70421 - GPS Satellite Simulator Control Working Group Meeting
2012-11-26
... DEPARTMENT OF DEFENSE Department of the Air Force GPS Satellite Simulator Control Working Group Meeting AGENCY: Space and Missile Systems Center, Global Positioning Systems (GPS) Directorate, Department of the Air Force, DoD. ACTION: Meeting Notice. SUMMARY: This meeting notice is to inform GPS...
Constrained Local UniversE Simulations: a Local Group factory
Carlesi, Edoardo; Sorce, Jenny G.; Hoffman, Yehuda; Gottlöber, Stefan; Yepes, Gustavo; Libeskind, Noam I.; Pilipenko, Sergey V.; Knebe, Alexander; Courtois, Hélène; Tully, R. Brent; Steinmetz, Matthias
2016-05-01
Near-field cosmology is practised by studying the Local Group (LG) and its neighbourhood. This paper describes a framework for simulating the `near field' on the computer. Assuming the Λ cold dark matter (ΛCDM) model as a prior and applying the Bayesian tools of the Wiener filter and constrained realizations of Gaussian fields to the Cosmicflows-2 (CF2) survey of peculiar velocities, constrained simulations of our cosmic environment are performed. The aim of these simulations is to reproduce the LG and its local environment. Our main result is that the LG is likely a robust outcome of the ΛCDMscenario when subjected to the constraint derived from CF2 data, emerging in an environment akin to the observed one. Three levels of criteria are used to define the simulated LGs. At the base level, pairs of haloes must obey specific isolation, mass and separation criteria. At the second level, the orbital angular momentum and energy are constrained, and on the third one the phase of the orbit is constrained. Out of the 300 constrained simulations, 146 LGs obey the first set of criteria, 51 the second and 6 the third. The robustness of our LG `factory' enables the construction of a large ensemble of simulated LGs. Suitable candidates for high-resolution hydrodynamical simulations of the LG can be drawn from this ensemble, which can be used to perform comprehensive studies of the formation of the LG.
Simulation of the $\\beta$ setup of the SSD group
Barron, Jared
2017-01-01
The objective of this study was to first implement a simulation in GEANT4 of the setup used by the SSD group to conduct measurements of the response of silicon detectors to ionizing radiation. This simulation was then compared to theoretical results on energy deposition in silicon sensors of varying thickness by ionizing radiation, as well as measurements of the energy deposition distribution taken using the apparatus being simulated. Simulation was also used to calibrate the trigger, estimating that a peak trigger signal of 1V corresponds to an energy deposited of 0.744 MeV. It was found that for the most probable value (MPV) of the energy deposition distribution, simulation coincides with theory within 5%, but the full-width at half-maximum (FWHM) is underestimated by 15%. Measurements were conducted on a 280$\\mu$m sensor and compared to simulation. The MPV was 10% lower in observation than simulation, and the FWHM was 10-15% higher.
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.
International Nuclear Information System (INIS)
Farvacque, M.; Faydide, B.; Dufeil, Ph.; Raimond, E.
2003-01-01
The use of Cathare in the simulators of pressurized water reactors has been effective since the beginning of the nineties. Scar project is the second stage of the Cathare strategy for the simulators, its main objective is the extension of the field of simulation to the accident situations in cold shutdown states. Work was carried out in 3 major areas: modelling, optimization and integration in the simulator. Throughout the project, the developments were part of a 3 stages validation strategy: -) elementary tests of the developments of new model on the N4 (1450 MW PWR); -) analytical tests and systems to ensure non regression of the validation of the physical laws of the Cathare code during the modifications carried out within the optimization stage; and -) overall tests of the SIPA-CP1 (900 MW PWR) simulator, controlled automatically by programmed scenarios including the transients which are carried out in PWR, the transients of the Regulatory Guides and the accident transients
Morality in group and family therapies: multiperson therapies and the 1992 ethics code.
Lakin, M
1994-11-01
Although virtually every psychotherapeutic approach or orientation has adapted group and family therapy to its conceptions of psychological dysfunctions and how to treat them, levels of training of practitioners in all of these approaches are often insufficient to meet the requirements of ethically as well as technically responsible conduct of treatment for persons in groups and families. The new ethics code (American Psychological Association [APA], 1992) does include a few issues specific to multiperson therapies, but other issues critical to the competent practice of group and family therapy remain unaddressed. The result can be confusing to those applying standards for individual therapy to multiperson therapies. It is argued that the classical ethical concerns of psychotherapists, informed consent, confidentiality, countertransference reactions, aand intrusions of therapist values, require special sensitivity to how they are expressed in mulitperson therapies. Practitioners of group and family therapies must be better sensitized to the technical distinctions and the associated ethical vulnerabilities of the modalities they use. Future planning for revision of the APA ethics code should take these factors into account.
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)
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
An efficient simulation method of a cyclotron sector-focusing magnet using 2D Poisson code
Energy Technology Data Exchange (ETDEWEB)
Gad Elmowla, Khaled Mohamed M; Chai, Jong Seo, E-mail: jschai@skku.edu; Yeon, Yeong H; Kim, Sangbum; Ghergherehchi, Mitra
2016-10-01
In this paper we discuss design simulations of a spiral magnet using 2D Poisson code. The Independent Layers Method (ILM) is a new technique that was developed to enable the use of two-dimensional simulation code to calculate a non-symmetric 3-dimensional magnetic field. In ILM, the magnet pole is divided into successive independent layers, and the hill and valley shape around the azimuthal direction is implemented using a reference magnet. The normalization of the magnetic field in the reference magnet produces a profile that can be multiplied by the maximum magnetic field in the hill magnet, which is a dipole magnet made of the hills at the same radius. Both magnets are then calculated using the 2D Poisson SUPERFISH code. Then a fully three-dimensional magnetic field is produced using TOSCA for the original spiral magnet, and the comparison of the 2D and 3D results shows a good agreement between both.
PRIAM: A self consistent finite element code for particle simulation in electromagnetic fields
International Nuclear Information System (INIS)
Le Meur, G.; Touze, F.
1990-06-01
A 2 1/2 dimensional, relativistic particle simulation code is described. A short review of the used mixed finite element method is given. The treatment of the driving terms (charge and current densities), initial, boundary conditions are exposed. Graphical results are shown
Multi-dimensional free-electron laser simulation codes: a comparison study
Biedron, S G; Dejus, Roger J; Faatz, B; Freund, H P; Milton, S V; Nuhn, H D; Reiche, S
2000-01-01
A self-amplified spontaneous emission (SASE) free-electron laser (FEL) is under construction at the Advanced Photon Source (APS). Five FEL simulation codes were used in the design phase: GENESIS, GINGER, MEDUSA, RON, and TDA3D. Initial comparisons between each of these independent formulations show good agreement for the parameters of the APS SASE FEL.
Multi-dimensional free-electron laser simulation codes: a comparison study
International Nuclear Information System (INIS)
Biedron, S. G.; Chae, Y. C.; Dejus, R. J.; Faatz, B.; Freund, H. P.; Milton, S. V.; Nuhn, H.-D.; Reiche, S.
1999-01-01
A self-amplified spontaneous emission (SASE) free-electron laser (FEL) is under construction at the Advanced Photon Source (APS). Five FEL simulation codes were used in the design phase: GENESIS, GINGER, MEDUSA, RON, and TDA3D. Initial comparisons between each of these independent formulations show good agreement for the parameters of the APS SASE FEL
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)
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…
International Nuclear Information System (INIS)
Soba, A; Denis, A; Lemes, M; Gonzalez, M E
2012-01-01
During the latest ten years the Codes and Models Section of the Nuclear Fuel Cycle Department has been developing the DIONISIO code, which simulates most of the main phenomena that take place within a fuel rod during the normal operation of a nuclear reactor: temperature distribution, thermal expansion, elastic and plastic strain, creep, irradiation growth, pellet-cladding mechanical interaction, fission gas release, swelling and densification. Axial symmetry is assumed and cylindrical finite elements are used to discretized the domain. The code has a modular structure and contains more than 40 interconnected models. A group of subroutines, designed to extend the application range of the fuel performance code DIONISIO to high burn up, has recently been included in the code. The new calculation tools, which are tuned for UO 2 fuels in LWR conditions, predict the radial distribution of power density, burnup and concentration of diverse nuclides within the pellet. New models of porosity and fission gas release in the rim, as well as the influence of the microstructure of this zone on the thermal conductivity of the pellet, are presently under development. A considerable computational challenge was the inclusion of the option of simulating the whole bar, by dividing it in a number of axial segments, at the user's choice, and solving in each segment the complete problem. All the general rod parameters (pressure, fission gas release, volume, etc.) are evaluated at the end of every time step. This modification allows taking into account the axial variation of the linear power and, consequently, evaluating the dependence of all the significant rod parameters with that coordinate. DIONISIO was elected for participating in the FUMEX III project of codes intercomparison, organized by IAEA, from 2008 to 2011. The results of the simulations performed within this project were compared with more than 30 experiments that involve more than 150 irradiated rods. The high number
Simulation of multibunch motion with the Headtail code and application to the CERN SPS and LHC
Mounet, N; Rumolo, G
2011-01-01
Multibunch instabilities due to beam-coupling impedance can be a critical limitation for synchrotrons operating with many bunches. It is particularly true for the LHC under nominal conditions, where according to theoretical predictions the 2808 bunches rely entirely on the performance of the transverse feedback system to remain stable. To study these instabilities, the HEADTAIL code has been extended to simulate the motion of many bunches under the action of wake fields. All the features already present in the single-bunch version of the code, such as synchrotron motion, chromaticity, amplitude detuning due to octupoles and the ability to load any kind of wake fields through tables, have remained available. This new code has been then parallelized in order to track thousands of bunches in a reasonable amount of time. The code was benchmarked against theory and exhibited a good agreement. We also show results for bunch trains in the LHC and compare them with beam-based measurements.
A New Code SORD for Simulation of Polarized Light Scattering in the Earth Atmosphere
Korkin, Sergey; Lyapustin, Alexei; Sinyuk, Aliaksandr; Holben, Brent
2016-01-01
We report a new publicly available radiative transfer (RT) code for numerical simulation of polarized light scattering in plane-parallel atmosphere of the Earth. Using 44 benchmark tests, we prove high accuracy of the new RT code, SORD (Successive ORDers of scattering). We describe capabilities of SORD and show run time for each test on two different machines. At present, SORD is supposed to work as part of the Aerosol Robotic NETwork (AERONET) inversion algorithm. For natural integration with the AERONET software, SORD is coded in Fortran 90/95. The code is available by email request from the corresponding (first) author or from ftp://climate1.gsfc.nasa.gov/skorkin/SORD/.
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
Uncertainty and sensitivity analysis in the scenario simulation with RELAP/SCDAP and MELCOR codes
International Nuclear Information System (INIS)
Garcia J, T.; Cardenas V, J.
2015-09-01
A methodology was implemented for analysis of uncertainty in simulations of scenarios with RELAP/SCDAP V- 3.4 bi-7 and MELCOR V-2.1 codes, same that are used to perform safety analysis in the Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS). The uncertainty analysis methodology chosen is a probabilistic method of type Propagation of uncertainty of the input parameters to the departure parameters. Therefore, it began with the selection of the input parameters considered uncertain and are considered of high importance in the scenario for its direct effect on the output interest variable. These parameters were randomly sampled according to intervals of variation or probability distribution functions assigned by expert judgment to generate a set of input files that were run through the simulation code to propagate the uncertainty to the output parameters. Then, through the use or ordered statistical and formula Wilks, was determined that the minimum number of executions required to obtain the uncertainty bands that include a population of 95% at a confidence level of 95% in the results is 93, is important to mention that in this method that number of executions does not depend on the number of selected input parameters. In the implementation routines in Fortran 90 that allowed automate the process to make the uncertainty analysis in transients for RELAP/SCDAP code were generated. In the case of MELCOR code for severe accident analysis, automation was carried out through complement Dakota Uncertainty incorporated into the Snap platform. To test the practical application of this methodology, two analyzes were performed: the first with the simulation of closing transient of the main steam isolation valves using the RELAP/SCDAP code obtaining the uncertainty band of the dome pressure of the vessel; while in the second analysis, the accident simulation of the power total loss (Sbo) was carried out with the Macarol code obtaining the uncertainty band for the
Simulation of LOCA power transients of CANDU6 by SCAN/RELAP-CANDU coupled code system
International Nuclear Information System (INIS)
Hong, In Seob; Kim, Chang Hyo; Hwang, Su Hyun; Kim, Man Woong; Chung, Bub Dong
2004-01-01
As can be seen in the standalone application of RELAP-CANDU for LOCA analysis of CANDU-PHWR, the system thermal-hydraulic code alone cannot predict the transient behavior accurately. Therefore, best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. The purpose of this research is to develop and test a coupled neutronics and thermal-hydraulics analysis code, SCAN (SNU CANDU-PHWR Neutronics) and RELAP-CANDU, for transient analysis of CANDU-PHWR's. For this purpose, a spatial kinetics calculation module of SCAN, a 3-D CANDU-PHWR neutronics design and analysis code, is dynamically coupled with RELAP-CANDU, the system thermal-hydraulic code for CANDU-PHWR. The performance of the coupled code system is examined by simulation of reactor power transients caused by a hypothetical Loss Of Coolant Accident (LOCA) in Wolsong units, which involves the insertion of positive void reactivity into the core in the course of transients. Specifically, a 40% Reactor Inlet Header (RIH) break LOCA was assumed for the test of the SCAN/RELAP-CANDU coupled code system analysis
On RELAP5-simulated High Flux Isotope Reactor reactivity transients: Code change and application
International Nuclear Information System (INIS)
Freels, J.D.
1993-01-01
This paper presents a new and innovative application for the RELAP5 code (hereafter referred to as ''the code''). The code has been used to simulate several transients associated with the (presently) draft version of the High-Flux Isotope Reactor (HFIR) updated safety analysis report (SAR). This paper investigates those thermal-hydraulic transients induced by nuclear reactivity changes. A major goal of the work was to use an existing RELAP5 HFIR model for consistency with other thermal-hydraulic transient analyses of the SAR. To achieve this goal, it was necessary to incorporate a new self-contained point kinetics solver into the code because of a deficiency in the point-kinetics reactivity model of the Mod 2.5 version of the code. The model was benchmarked against previously analyzed (known) transients. Given this new code, four event categories defined by the HFIR probabilistic risk assessment (PRA) were analyzed: (in ascending order of severity) a cold-loop pump start; run-away shim-regulating control cylinder and safety plate withdrawal; control cylinder ejection; and generation of an optimum void in the target region. All transients are discussed. Results of the bounding incredible event transient, the target region optimum void, are shown. Future plans for RELAP5 HFIR applications and recommendations for code improvements are also discussed
[Variation of CAG repeats in coding region of ATXN2 gene in different ethnic groups].
Chen, Xiao-Chen; Sun, Hao; Mi, Dong-Qing; Huang, Xiao-Qin; Lin, Ke-Qin; Yi, Wen; Yu, Liang; Shi, Lei; Shi, Li; Yang, Zhao-Qing; Chu, Jia-You
2011-04-01
Toinvestigate CAG repeats variation of ATXN2 gene coding region in six ethnic groups that live in comparatively different environments, to evaluate whether these variations are under positive selection, and to find factors driving selection effects, 291 unrelated healthy individuals were collected from six ethnic groups and their STR geneotyping was performed. The frequencies of alleles and genotypes were counted and thereby Slatkin's linearized Fst values were calculated. The UPGMA tree against this gene was constructed. The MDS analysis among these groups was carried out as well. The results from the linearized Fst values indicated that there were significant evolutionary differences of the STR in ATXN2 gene between Hui and Yi groups, but not among the other 4 groups. Further analysis was performed by combining our data with published data obtained from other groups. These results indicated that there were significant differences between Japanese and other groups including Hui, Hani, Yunnan Mongolian, and Inner Mongolian. Both Hui and Mongolian from Inner Mongolia were significantly different from Han. In conclusion, the six ethnic groups had their own distribution characterizations of allelic frequencies of ATXN2 STR, and the potential cause of frequency changes in rare alleles could be the consequence of positive selection.
Harmonization of nuclear codes and standards. Pacific nuclear council working and task group report
International Nuclear Information System (INIS)
Dua, Shami
2008-01-01
Nuclear codes and standards have been an integral part of the nuclear industry since its inception. As the industry came into the main stream over the 2nd half of the 20th century, a number of national and international standards were developed to support a specific nuclear reactor concept. These codes and standards have been a key component of the industry to maintain its focus on nuclear safety, reliability and quality. Both national and international standards have served the industry well in obtaining public, shareholder and regulatory acceptance. The existing suite of national and international standards is required to support the emerging nuclear renaissance. However as noted above under Pacific Nuclear Council (PNC), Manufacturing Design Evaluation Program (MDEP) and SMiRT discussions, the time has come now for the codes and standards writing bodies and the industry to take the next step to examine the relevance of existing suite in view of current needs and challenges. This review must account for the changing global environment including global supply chain and regulatory framework, resources, deregulation, free trade, and industry need for competitiveness and performance excellence. The Task Group (TG) has made limited progress in this review period as no additional information on the listing of codes and standards have been received from the members. However, TG Chair has been successful in obtaining considerable interest from some additional individuals from the member countries. It is important that PNC management seek additional participation from the member countries and asks for their active engagement in the Working Group (WG) TG activities to achieve its mandate and deliverables. The harmonization of codes and standards is a key area for the emerging nuclear renaissance and as noted by a number of international organizations (refer to MDEP action noted above) that these tasks can not be completed unless we have the right level of resources and
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)
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
Finite element methods in a simulation code for offshore wind turbines
Kurz, Wolfgang
1994-06-01
Offshore installation of wind turbines will become important for electricity supply in future. Wind conditions above sea are more favorable than on land and appropriate locations on land are limited and restricted. The dynamic behavior of advanced wind turbines is investigated with digital simulations to reduce time and cost in development and design phase. A wind turbine can be described and simulated as a multi-body system containing rigid and flexible bodies. Simulation of the non-linear motion of such a mechanical system using a multi-body system code is much faster than using a finite element code. However, a modal representation of the deformation field has to be incorporated in the multi-body system approach. The equations of motion of flexible bodies due to deformation are generated by finite element calculations. At Delft University of Technology the simulation code DUWECS has been developed which simulates the non-linear behavior of wind turbines in time domain. The wind turbine is divided in subcomponents which are represented by modules (e.g. rotor, tower etc.).
IDRIFF two-phase simulation code and its application to the study of a pressurizer
International Nuclear Information System (INIS)
Sollychin, R.; Garland, W.J.; Chang, J.S.
1987-01-01
The simulation code IDRIFF (Integrated Drift-flux Formulation) has been developed as a convenient tool in two-phase flow analysis, which demands the following two conflicting requirements: (a) provision for detailed information on local phenomena in the flow;(b) fast calculation of averaged values of parameters for engineering type flow problems. A small scale pressurizer made of a glass tank and its associated systems were set-up to simulate the behavior of nuclear power plant pressurizer. Flow-pattern observation in the pressurizer at quasi-steady-state, and measurement of pressure, temperature and void fraction at certain fixed locations during both quasi-steady-state and transient experiments are obtained. The IDRIFF code is then applied to supplement the empirical experiment in generating a complete data base, so that extensive theoretical and empirical analyses of the pressurizer behaviour can be systematically performed or verified. The technique of applying the IDRIFF code to simulate both the quasi-steady-state and transient experiment is discussed in detail in the paper. The result of the simulation is in good agreement with measurements taken during the experiment. Analysis of both the empirical and numerical data results in: (1) relationships among void fraction, heater power and steam-bleed flow;(2) a pressurizer flow-regime map and (3) constitutive equations for bubble rising flow and droplet drop flow. This strongly suggests that the approach of extrapolating information obtained from empirical experiment by numerical simulation is a useful method in two-phase flow analysis
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.
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)
Integrated fast ignition simulation of cone-guided target with three codes
Energy Technology Data Exchange (ETDEWEB)
Sakagami, H. [Hyogo Univ., Computer Engineering, Himeji, Hyogo (Japan); Johzaki, T.; Nagatomo, H.; Mima, K. [Osaka Univ., Institute of Laser Engineering, Suita, Osaka (Japan)
2004-07-01
It was reported that the fuel core was heated up to {approx} 0.8 keV in the fast ignition experiments with cone-guided targets, but they could not theoretically explain heating mechanisms and achievement of such high temperature. Thus simulations should play an important role in estimating the scheme performance, and we must simulate each phenomenon with individual codes and integrate them under the Fast Ignition Integrated Interconnecting code project. In the previous integrated simulations, fast electrons generated by the laser-plasma interaction were too hot to efficiently heat the core and we got only a 0.096 keV temperature rise. Including the density gap at the contact surface between the cone tip and the imploded plasma, the period of core heating became longer and the core was heated by 0.162 keV, about 69% higher increment compared with ignoring the density gap effect. (authors)
Use of sensitivity-information for the adaptive simulation of thermo-hydraulic system codes
International Nuclear Information System (INIS)
Kerner, Alexander M.
2011-01-01
Within the scope of this thesis the development of methods for online-adaptation of dynamical plant simulations of a thermal-hydraulic system code to measurement data is depicted. The described approaches are mainly based on the use of sensitivity-information in different areas: statistical sensitivity measures are used for the identification of the parameters to be adapted and online-sensitivities for the parameter adjustment itself. For the parameter adjustment the method of a ''system-adapted heuristic adaptation with partial separation'' (SAHAT) was developed, which combines certain variants of parameter estimation and control with supporting procedures to solve the basic problems. The applicability of the methods is shown by adaptive simulations of a PKL-III experiment and by selected transients in a nuclear power plant. Finally the main perspectives for the application of a tracking simulator on a system code are identified.
A System Structure for a VHTR-SI Process Dynamic Simulation Code
International Nuclear Information System (INIS)
Chang, Jiwoon; Shin, Youngjoon; Kim, Jihwan; Lee, Kiyoung; Lee, Wonjae; Chang, Jonghwa; Youn, Cheung
2008-01-01
The VHTR-SI process dynamic simulation code embedded in a mathematical solution engine is an application software system that simulates the dynamic behavior of the VHTR-SI process. Also, the software system supports a user friendly graphical user interface (GUI) for user input/out. Structured analysis techniques were developed in the late 1970s by Yourdon, DeMarco, Gane and Sarson for applying a systematic approach to a systems analysis. It included the use of data flow diagrams and data modeling and fostered the use of an implementation-independent graphical notation for a documentation. In this paper, we present a system structure for a VHRT-SI process dynamic simulation code by using the methodologies of structured analysis
One-dimensional transport code for one-group problems in plane geometry
International Nuclear Information System (INIS)
Bareiss, E.H.; Chamot, C.
1970-09-01
Equations and results are given for various methods of solution of the one-dimensional transport equation for one energy group in plane geometry with inelastic scattering and an isotropic source. After considerable investigation, a matrix method of solution was found to be faster and more stable than iteration procedures. A description of the code is included which allows for up to 24 regions, 250 points, and 16 angles such that the product of the number of angles and the number of points is less than 600
International Nuclear Information System (INIS)
Mori, Takamasa; Nakagawa, Masayuki; Sasaki, Makoto.
1988-11-01
We have developed a group of computer codes to realize the accurate transport calculation by using the multi-group double-differential form cross section. This type of cross section can correctly take account of the energy-angle correlated reaction kinematics. Accordingly, the transport phenomena in materials with highly anisotropic scattering are accurately calculated by using this cross section. They include the following four codes or code systems: PROF-DD : a code system to generate the multi-group double-differential form cross section library by processing basic nuclear data file compiled in the ENDF / B-IV or -V format, ANISN-DD : a one-dimensional transport code based on the discrete ordinate method, DOT-DD : a two-dimensional transport code based on the discrete ordinate method, MORSE-DD : a three-dimensional transport code based on the Monte Carlo method. In addition to these codes, several auxiliary codes have been developed to process calculated results. This report describes the calculation algorithm employed in these codes and how to use them. (author)
Phase 1 Validation Testing and Simulation for the WEC-Sim Open Source Code
Ruehl, K.; Michelen, C.; Gunawan, B.; Bosma, B.; Simmons, A.; Lomonaco, P.
2015-12-01
WEC-Sim is an open source code to model wave energy converters performance in operational waves, developed by Sandia and NREL and funded by the US DOE. The code is a time-domain modeling tool developed in MATLAB/SIMULINK using the multibody dynamics solver SimMechanics, and solves the WEC's governing equations of motion using the Cummins time-domain impulse response formulation in 6 degrees of freedom. The WEC-Sim code has undergone verification through code-to-code comparisons; however validation of the code has been limited to publicly available experimental data sets. While these data sets provide preliminary code validation, the experimental tests were not explicitly designed for code validation, and as a result are limited in their ability to validate the full functionality of the WEC-Sim code. Therefore, dedicated physical model tests for WEC-Sim validation have been performed. This presentation provides an overview of the WEC-Sim validation experimental wave tank tests performed at the Oregon State University's Directional Wave Basin at Hinsdale Wave Research Laboratory. Phase 1 of experimental testing was focused on device characterization and completed in Fall 2015. Phase 2 is focused on WEC performance and scheduled for Winter 2015/2016. These experimental tests were designed explicitly to validate the performance of WEC-Sim code, and its new feature additions. Upon completion, the WEC-Sim validation data set will be made publicly available to the wave energy community. For the physical model test, a controllable model of a floating wave energy converter has been designed and constructed. The instrumentation includes state-of-the-art devices to measure pressure fields, motions in 6 DOF, multi-axial load cells, torque transducers, position transducers, and encoders. The model also incorporates a fully programmable Power-Take-Off system which can be used to generate or absorb wave energy. Numerical simulations of the experiments using WEC-Sim will be
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)
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).
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.
Comparison of the MAAP4 code with the station blackout simulation in the IIST facility
International Nuclear Information System (INIS)
Robert E Henry; Christopher E Henry; Chan Y Paik; George M Hauser
2005-01-01
Full text of publication follows: The Modular Accident Analysis Program (MAAP) is an integral system model to assess challenges to the reactor core, Reactor Coolant System (RCS) and containment for accident conditions. MAAP4 is the current version used by the MAAP Users Group to assess the responses to a spectrum of accident conditions. Benchmarking of the MAAP code with integral system experiments has been a continuing effort by MAAP developers and users. Several of these have been configured into dynamic benchmarks and are included in Volume III (Benchmarking) of the MAAP4 Users Manual (EPRI, 2004). One such integral experiment is the INER integral system test (IIST) constructed at the Institute of Nuclear Energy Research in Taiwan. This experimental facility is a reduced height, reduced pressure representation of a 3-loop PWR and has been used to examine several different types of accident sequences. One of these is a station blackout simulation with loss of auxiliary feedwater at the time that the transient is initiated. This is an important integral experiment to be compared with the MAAP4 code models. A parameter file (those values representing the system design and boundary experimental conditions) has been developed for the IIST facility and an input deck has been configured to represent a station blackout sequence with instantaneous loss of auxiliary feedwater. Of importance in this benchmark is (a) the rate at which the secondary side inventory is depleted, (b) the depletion of water within the reactor pressure vessel and (c) the time at which the top of the reactor core is uncovered. Comparisons have been made with these three different intervals and there is good agreement between the timing of these events for the MAAP4 benchmark. This is important since this reference sequence represents a set of boundary conditions that is continually with subsequent analyses being perturbations on this type of accident sequence. The good agreement between MAAP4 and
Overall simulation of a HTGR plant with the gas adapted MANTA code
International Nuclear Information System (INIS)
Emmanuel Jouet; Dominique Petit; Robert Martin
2005-01-01
Full text of publication follows: AREVA's subsidiary Framatome ANP is developing a Very High Temperature Reactor nuclear heat source that can be used for electricity generation as well as cogeneration including hydrogen production. The selected product has an indirect cycle architecture which is easily adapted to all possible uses of the nuclear heat source. The coupling to the applications is implemented through an Intermediate Heat exchanger. The system code chosen to calculate the steady-state and transient behaviour of the plant is based on the MANTA code. The flexible and modular MANTA code that is originally a system code for all non LOCA PWR plant transients, has been the subject of new developments to simulate all the forced convection transients of a nuclear plant with a gas cooled High Temperature Reactor including specific core thermal hydraulics and neutronics modelizations, gas and water steam turbomachinery and control structure. The gas adapted MANTA code version is now able to model a total HTGR plant with a direct Brayton cycle as well as indirect cycles. To validate these new developments, a modelization with the MANTA code of a real plant with direct Brayton cycle has been performed and steady-states and transients compared with recorded thermal hydraulic measures. Finally a comparison with the RELAP5 code has been done regarding transient calculations of the AREVA indirect cycle HTR project plant. Moreover to improve the user-friendliness in order to use MANTA as a systems conception, optimization design tool as well as a plant simulation tool, a Man- Machine-Interface is available. Acronyms: MANTA Modular Advanced Neutronic and Thermal hydraulic Analysis; HTGR High Temperature Gas-Cooled Reactor. (authors)
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
Improved core-edge tokamak transport simulations with the CORSICA 2 code
International Nuclear Information System (INIS)
Tarditi, A.; Cohen, R.H.; Crotinger, J.A.
1996-01-01
The CORSICA 2 code models the nonlinear transport between the core and the edge of a tokamak plasma. The code couples a 2D axisymmetric edge/SOL model (UEDGE) to a 1D model for the radial core transport in toroidal flux coordinates (the transport module from the CORSICA 1 code). The core density and temperature profiles are joined to the flux-surface average profiles from the 2D code sufficiently inside the magnetic separatrix, at a flux surface on which the edge profiles are approximately constant. In the present version of the code, the deuterium density and electron and ion temperatures are coupled. The electron density is determined by imposing quasi-neutrality, both in the core and in the edge. The model allows the core-edge coupling of multiple ion densities while retaining a single temperature (corresponding to the equilibration value) for the all ion species. Applications of CORSICA 2 to modeling the DIII-D tokamak are discussed. This work will focus on the simulation of the L-H transition, coupling a single ion species (deuterium) and the two (electron and ion) temperatures. These simulations will employ a new self-consistent model for the L-H transition that is being implemented in the UEDGE code. Applications to the modeling of ITER ignition scenarios are also discussed. This will involve coupling a second density species (the thermal alphas), bringing the total number of coupled variables up to four. Finally, the progress in evolving the magnetic geometry is discussed. Currently, this geometry is calculated by CORSICA's MHD equilibrium module (TEQ) at the beginning of the run and fixed thereafter. However, CORSICA 1 can evolve this geometry quasistatically, and this quasistatic treatment is being extended to include the edge/SOL geometry. Recent improvements for code speed-up are also presented
HEP Software Foundation Community White Paper Working Group - Detector Simulation
Energy Technology Data Exchange (ETDEWEB)
Apostolakis, J; et al.
2018-03-12
A working group on detector simulation was formed as part of the high-energy physics (HEP) Software Foundation's initiative to prepare a Community White Paper that describes the main software challenges and opportunities to be faced in the HEP field over the next decade. The working group met over a period of several months in order to review the current status of the Full and Fast simulation applications of HEP experiments and the improvements that will need to be made in order to meet the goals of future HEP experimental programmes. The scope of the topics covered includes the main components of a HEP simulation application, such as MC truth handling, geometry modeling, particle propagation in materials and fields, physics modeling of the interactions of particles with matter, the treatment of pileup and other backgrounds, as well as signal processing and digitisation. The resulting work programme described in this document focuses on the need to improve both the software performance and the physics of detector simulation. The goals are to increase the accuracy of the physics models and expand their applicability to future physics programmes, while achieving large factors in computing performance gains consistent with projections on available computing resources.
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)
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
Directory of Open Access Journals (Sweden)
Juan P. Marcolongo
2018-03-01
Full Text Available 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.
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.
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
Application of the TRAC-PD2 code to the simulation of the CANON experiment
International Nuclear Information System (INIS)
Neves Conti, T. das; Freitas, R.L.
1985-01-01
A comparison between the TRAC -PD2 code calculations and results from the CANON experiment is presented. The CANON experiment simulates the loss of coolant accident through the depressurization of a horizontal tube containing water at different temperatures. The experiment consist of the instantaneous rupture at one end of the tubing and the corresponding pressure and void fraction measurements during the transient. The comparison shows that the TRAC-PD2 code predicts satisfactorily the pressure and void fraction evolution in the CANON experiment. (F.C.) [pt
Evaluation of the Trac-PF1 code for simulating the Neptun reflooding experiment
International Nuclear Information System (INIS)
Pontedeiro, A.C.; Galetti, M.R.S.
1991-01-01
The present work presents an assessment of the TRAC-BF1 code using the results of the NEPTUN experiment which simulates the reflooding in a loss-of-coolant accident (LOCA) in a PWR. The NEPTUN experiment is composed of an array of electrically-heated tubes where the reflooding condition can be tested. Two types of tests results are presented and compared with the values obtained with the TRAC-BF1 code. From this comparison it is concluded that TRAC is suitable for verifying accident analysis. (author)
Draft-Report on the RTOP-Code Simulations in the Fumex-3 Exercises
International Nuclear Information System (INIS)
Likhanskii, V.
2013-01-01
The RTOP-code is used for prediction of the following main parameters of a fuel rod during irradiation: - internal gas pressure in fuel rods - mechanical stresses in cladding and fuel pellets due to PCMI. Simulation of fuel behavior by the RTOP-code is based on various physical models. 1. Thermal models. 2. Evolution of Burnup and Pu distribution in the fuel rod during irradiation. 3. Fission gas release models. 4. Models of microstructure evolution of the fuel. 5. Mechanical stresses models and models for description of plastic deformations of fuel and cladding. (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.
DELOCA, a code for simulation of CANDU fuel channel in thermal transients
International Nuclear Information System (INIS)
Mihalache, M.; Florea, Silviu; Ionescu, V.; Pavelescu, M.
2005-01-01
Full text: In certain LOCA scenarios into the CANDU fuel channel, the ballooning of the pressure tube and the contact with the calandria tube can occur. After the contact moment, a radial heat transfer from cooling fluid to moderator arises through the contact area. If the temperature of channel walls increases, the contact area is drying, the heat transfer becomes inefficiently and the fuel channel could lose its integrity. DELOCA code was developed to simulate the mechanical behaviour of pressure tube during pre-contact transition, and mechanical and thermal behaviour of pressure tube and calandria tube after the contact between the two tubes. The code contains a few models: the creep of Zr-2.5%Nb alloy, the heat transfer by conduction through the cylindrical walls, channel failure criteria and calculus of heat transfer at the calandria tube - moderator interface. This code evaluates the contact and channel failure moments. This code was systematically verified by Contact1 and Cathena codes. This paper presents the results obtained at different temperature increasing rates. In addition, the contact moment for a RIH 5% postulated accident was calculated. The Cathena thermo-hydraulic code provided the input data. (authors)
DELOCA, a code for simulation of CANDU fuel channel in thermal transients
International Nuclear Information System (INIS)
Mihalache, M.; Florea, Silviu; Ionescu, V.; Pavelescu, M.
2005-01-01
In certain LOCA scenarios into the CANDU fuel channel, the ballooning of the pressure tube and the contact with the calandria tube can occur. After the contact moment, a radial heat transfer from cooling fluid to moderator arises through the contact area. If the temperature of channel walls increases, the contact area is drying, the heat transfer becomes inefficiently and the fuel channel could lose its integrity. DELOCA code was developed to simulate the mechanical behaviour of pressure tube during pre-contact transition, and mechanical and thermal behaviour of pressure tube and calandria tube after the contact between the two tubes. The code contains a few models: the creep of Zr-2.5%Nb alloy, the heat transfer by conduction through the cylindrical walls, channel failure criteria and calculus of heat transfer at the calandria tube - moderator interface. This code evaluates the contact and channel failure moments. This code was systematically verified by Contact1 and Cathena codes. This paper presents the results obtained at different temperature increasing rates. In addition, the contact moment for a RIH 5% postulated accident was calculated. The Cathena thermo-hydraulic code provided the input data. (authors)
Use of a hybrid code for global-scale plasma simulation
International Nuclear Information System (INIS)
Swift, D.W.
1996-01-01
This paper presents a demonstration of the use of a hybrid code to model the Earth's magnetosphere on a global scale. The typical hybrid code calculates the interaction of fully kinetic ions and a massless electron fluid with the magnetic field. This code also includes a fluid ion component to approximate the cold ionospheric plasma that spatially overlaps with the discrete particle component. Other innovative features of the code include a numerically generated curvilinear coordinate system and subcycling of the magnetic field update to the particle push. These innovations allow the code to accommodate disparate time and distance scales. The demonstration is a simulation of the noon meridian plane of the magnetosphere. The code exhibits the formation of fast and slow-mode shocks and tearing reconnection at the magnetopause. New results include particle acceleration in the cusp and nearly field aligned currents linking the cusp and polar ionosphere. The paper also describes a density depletion instability and measures to avoid it. 27 refs., 4 figs
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
Energy Technology Data Exchange (ETDEWEB)
Schultz, Peter Andrew
2011-12-01
The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. Achieving the objective of modeling the performance of a disposal scenario requires describing processes involved in waste form degradation and radionuclide release at the subcontinuum scale, beginning with mechanistic descriptions of chemical reactions and chemical kinetics at the atomic scale, and upscaling into effective, validated constitutive models for input to high-fidelity continuum scale codes for coupled multiphysics simulations of release and transport. Verification and validation (V&V) is required throughout the system to establish evidence-based metrics for the level of confidence in M&S codes and capabilities, including at the subcontiunuum scale and the constitutive models they inform or generate. This Report outlines the nature of the V&V challenge at the subcontinuum scale, an approach to incorporate V&V concepts into subcontinuum scale modeling and simulation (M&S), and a plan to incrementally incorporate effective V&V into subcontinuum scale M&S destined for use in the NEAMS Waste IPSC work flow to meet requirements of quantitative confidence in the constitutive models informed by subcontinuum scale phenomena.
International Nuclear Information System (INIS)
Douglass, M.; Bezak, E.
2010-01-01
Full text: Radiobiology science is important for cancer treatment as it improves our understanding of radiation induced cell death. Monte Carlo simulations playa crucial role in developing improved knowledge of cellular processes. By model Ii ng the cell response to radiation damage and verifying with experimental data, understanding of cell death through direct radiation hits and bystander effects can be obtained. A Monte Carlo input code was developed using 'Geant4' to simulate cellular level radiation interactions. A physics list which enables physically accurate interactions of heavy ions to energies below 100 e V was implemented. A simple biological cell model was also implemented. Each cell consists of three concentric spheres representing the nucleus, cytoplasm and the membrane. This will enable all critical cell death channels to be investigated (i.e. membrane damage, nucleus/DNA). The current simulation has the ability to predict the positions of ionization events within the individual cell components on I micron scale. We have developed a Geant4 simulation for investigation of radiation damage to cells on sub-cellular scale (∼I micron). This code currently allows the positions of the ionisation events within the individual components of the cell enabling a more complete picture of cell death to be developed. The next stage will include expansion of the code to utilise non-regular cell lattice. (author)
International Nuclear Information System (INIS)
Schultz, Peter Andrew
2011-01-01
The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M and S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. Achieving the objective of modeling the performance of a disposal scenario requires describing processes involved in waste form degradation and radionuclide release at the subcontinuum scale, beginning with mechanistic descriptions of chemical reactions and chemical kinetics at the atomic scale, and upscaling into effective, validated constitutive models for input to high-fidelity continuum scale codes for coupled multiphysics simulations of release and transport. Verification and validation (V and V) is required throughout the system to establish evidence-based metrics for the level of confidence in M and S codes and capabilities, including at the subcontiunuum scale and the constitutive models they inform or generate. This Report outlines the nature of the V and V challenge at the subcontinuum scale, an approach to incorporate V and V concepts into subcontinuum scale modeling and simulation (M and S), and a plan to incrementally incorporate effective V and V into subcontinuum scale M and S destined for use in the NEAMS Waste IPSC work flow to meet requirements of quantitative confidence in the constitutive models informed by subcontinuum scale phenomena.
SimCommSys: taking the errors out of error-correcting code simulations
Directory of Open Access Journals (Sweden)
Johann A. Briffa
2014-06-01
Full Text Available In this study, we present SimCommSys, a simulator of communication systems that we are releasing under an open source license. The core of the project is a set of C + + libraries defining communication system components and a distributed Monte Carlo simulator. Of principal interest is the error-control coding component, where various kinds of binary and non-binary codes are implemented, including turbo, LDPC, repeat-accumulate and Reed–Solomon. The project also contains a number of ready-to-build binaries implementing various stages of the communication system (such as the encoder and decoder, a complete simulator and a system benchmark. Finally, SimCommSys also provides a number of shell and python scripts to encapsulate routine use cases. As long as the required components are already available in SimCommSys, the user may simulate complete communication systems of their own design without any additional programming. The strict separation of development (needed only to implement new components and use (to simulate specific constructions encourages reproducibility of experimental work and reduces the likelihood of error. Following an overview of the framework, we provide some examples of how to use the framework, including the implementation of a simple codec, the specification of communication systems and their simulation.
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)
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)
Development of dynamic simulation code for fuel cycle of fusion reactor
International Nuclear Information System (INIS)
Aoki, Isao; Seki, Yasushi; Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan
1999-02-01
A dynamic simulation code for fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during 2 days pulse operation cycles. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the fuel burn and the function of exhaust, purification, and supply. The processing constants of subsystem for steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using this code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)
Capability of the RELAP5 code to simulate natural circulation behaviour in test facilities
International Nuclear Information System (INIS)
Mangal, Amit; Jain, Vikas; Nayak, A.K.
2011-01-01
In the present study, one of the extensively used best estimate code RELAP5 has been used for simulation of steady state, transient and stability behavior of natural circulation based experimental facilities, such as the High-Pressure Natural Circulation Loop (HPNCL) and the Parallel Channel Loop (PCL) installed and operating at BARC. The test data have been generated for a range of pressure, power and subcooling conditions. The computer code RELAP5/MOD3.2 was applied to predict the transient natural circulation characteristics under single-phase and two-phase conditions, thresholds of flow instability, amplitude and frequency of flow oscillations for different operating conditions of the loops. This paper presents the effect of nodalisation in prediction of natural circulation behavior in test facilities and a comparison of experimental data in with that of code predictions. The errors associated with the predictions are also characterized
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)
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)
Rabie, M.; Franck, C. M.
2016-06-01
We present a freely available MATLAB code for the simulation of electron transport in arbitrary gas mixtures in the presence of uniform electric fields. For steady-state electron transport, the program provides the transport coefficients, reaction rates and the electron energy distribution function. The program uses established Monte Carlo techniques and is compatible with the electron scattering cross section files from the open-access Plasma Data Exchange Project LXCat. The code is written in object-oriented design, allowing the tracing and visualization of the spatiotemporal evolution of electron swarms and the temporal development of the mean energy and the electron number due to attachment and/or ionization processes. We benchmark our code with well-known model gases as well as the real gases argon, N2, O2, CF4, SF6 and mixtures of N2 and O2.
Simulation of heat and mass transfer in boiling water with the Melodif code
International Nuclear Information System (INIS)
Freydier, P.; Chen, O.; Olive, J.; Simonin, O.
1991-04-01
The Melodif code is developed at Electricite de France, Research and Development Division. It is an eulerian two dimensional code for the simulation of turbulent two phase flows (a three dimensional code derived from Melodif, ASTRID, is currently being prepared). Melodif is based on the two fluid model, solving the equations of conservation for mass, momentum and energy, for both phases. In such a two fluid model, the description of interfacial transfers between phases is a crucial issue. The model used applies to a dominant continuous phase, and a dispersed phase. A good description of interfacial momentum transfer exists in the standard MELODIF code: the drag force, the apparent mass force... are taken into account. An important factor for interfacial transfers is the interfacial area per volume unit. With the assumption of spherical gas bubbles, an equation has been written for this variable. In the present wok, a model has been tested for interfacial heat and mass transfer in the case of boiling water: it is assumed that mass transfer is controlled by heat transfer through the latent massic energy taken in the phase that vaporizes (or condenses). This heat and mass transfer model has been tested in various configurations: - a cylinder with water flowing inside, is being heated. Boiling takes place near the wall, while bubbles migrating to the core of the flow recondense. This roughly simulates a sub-cooled boiling phenomenon. - a box containing liquid water is depressurized. Boiling takes place in the whole volume of the fluid. The Melodif code can simulate this configuration due to the implicitation of the relation between interphase mass transfer and the pressure variable
An Enhanced GINGER Simulation Code with Harmonic Emission and HDF5 IO Capabilities
International Nuclear Information System (INIS)
Fawley, William M.
2006-01-01
GINGER [1] is an axisymmetric, polychromatic (r-z-t) FEL simulation code originally developed in the mid-1980's to model the performance of single-pass amplifiers. Over the past 15 years GINGER's capabilities have been extended to include more complicated configurations such as undulators with drift spaces, dispersive sections, and vacuum chamber wakefield effects; multi-pass oscillators; and multi-stage harmonic cascades. Its coding base has been tuned to permit running effectively on platforms ranging from desktop PC's to massively parallel processors such as the IBM-SP. Recently, we have made significant changes to GINGER by replacing the original predictor-corrector field solver with a new direct implicit algorithm, adding harmonic emission capability, and switching to the HDF5 IO library [2] for output diagnostics. In this paper, we discuss some details regarding these changes and also present simulation results for LCLS SASE emission at λ = 0.15 nm and higher harmonics
A high-resolution code for large eddy simulation of incompressible turbulent boundary layer flows
Cheng, Wan
2014-03-01
We describe a framework for large eddy simulation (LES) of incompressible turbulent boundary layers over a flat plate. This framework uses a fractional-step method with fourth-order finite difference on a staggered mesh. We present several laminar examples to establish the fourth-order accuracy and energy conservation property of the code. Furthermore, we implement a recycling method to generate turbulent inflow. We use the stretched spiral vortex subgrid-scale model and virtual wall model to simulate the turbulent boundary layer flow. We find that the case with Reθ ≈ 2.5 × 105 agrees well with available experimental measurements of wall friction, streamwise velocity profiles and turbulent intensities. We demonstrate that for cases with extremely large Reynolds numbers (Reθ = 1012), the present LES can reasonably predict the flow with a coarse mesh. The parallel implementation of the LES code demonstrates reasonable scaling on O(103) cores. © 2013 Elsevier Ltd.
TMCC: a transient three-dimensional neutron transport code by the direct simulation method - 222
International Nuclear Information System (INIS)
Shen, H.; Li, Z.; Wang, K.; Yu, G.
2010-01-01
A direct simulation method (DSM) is applied to solve the transient three-dimensional neutron transport problems. DSM is based on the Monte Carlo method, and can be considered as an application of the Monte Carlo method in the specific type of problems. In this work, the transient neutronics problem is solved by simulating the dynamic behaviors of neutrons and precursors of delayed neutrons during the transient process. DSM gets rid of various approximations which are always necessary to other methods, so it is precise and flexible in the requirement of geometric configurations, material compositions and energy spectrum. In this paper, the theory of DSM is introduced first, and the numerical results obtained with the new transient analysis code, named TMCC (Transient Monte Carlo Code), are presented. (authors)
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...
Simulation of the KAERI PASCAL Test with MARS-KS and TRACE Codes
Energy Technology Data Exchange (ETDEWEB)
Lee, Kyung Won; Cheong, Aeju; Shin, Andong; Cho, Min Ki [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)
2016-10-15
In order to validate the operational performance of the PAFS, KAERI has performed the experimental investigation using the PASCAL (PAFS Condensing heat removal Assessment Loop) facility. In this study, we simulated the KAERI PASCAL SS-540-P1 test with MARS-KS V1.4 and TRACE V5.0 p4 codes to assess the code predictability for the condensation heat transfer inside the passive auxiliary feedwater system. We simulated the KAERI PASCAL SS-540-P1 test with MARS-KS V1.4 and TRACE V5.0 p4 codes to assess the code predictability for the condensation heat transfer inside the passive auxiliary feedwater system. The calculated results of heat flux, inner wall surface temperature of the condensing tube, fluid temperature, and steam mass flow rate are compared with the experimental data. The result shows that the MARS-KS generally under-predict the heat fluxes. The TRACE over-predicts the heat flux at tube inlet region and under-predicts it at tube outlet region. The TRACE prediction shows larger amount of steam condensation by about 3% than the MARS-KS prediction.
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)
Development of 3D multi-group neutron diffusion code for hexagonal geometry
International Nuclear Information System (INIS)
Sun Wei; Wang Kan; Ni Dongyang; Li Qing
2013-01-01
Based on the theory of new flux expansion nodal method to solve the neutron diffusion equations, the intra-nodal fluence rate distribution was expanded in a series of analytic basic functions for each group. In order to improve the accuracy of calculation result, continuities of neutron fluence rate and current were utilized across the nodal surfaces. According to the boundary conditions, the iteration method was adopted to solve the diffusion equation, where inner iteration speedup method is Gauss-Seidel method and outer is Lyusternik-Wagner. A new speedup method (one-outer-iteration and multi-inner-iteration method) was proposed according to the characteristic that the convergence speed of multiplication factor is faster than that of neutron fluence rate and the update of inner iteration matrix is slow. Based on the proposed model, the code HANDF-D was developed and tested by 3D two-group vver440 benchmark, experiment 2 of HFETR, 3D four-group thermal reactor benchmark, and 3D seven-group fast reactor benchmark. The numerical results show that HANDF-D can predict accurately the multiplication factor and nodal powers. (authors)
Further development of the V-code for recirculating linear accelerator simulations
Energy Technology Data Exchange (ETDEWEB)
Franke, Sylvain; Ackermann, Wolfgang; Weiland, Thomas [Institut fuer Theorie Elektromagnetischer Felder, Technische Universitaet Darmstadt (Germany); Eichhorn, Ralf; Hug, Florian; Kleinmann, Michaela; Platz, Markus [Institut fuer Kernphysik, Technische Universitaet Darmstadt (Germany)
2011-07-01
The Superconducting Darmstaedter LINear Accelerator (S-DALINAC) installed at the institute of nuclear physics (IKP) at TU Darmstadt is designed as a recirculating linear accelerator. The beam is first accelerated up to 10 MeV in the injector beam line. Then it is deflected by 180 degrees into the main linac. The linac section with eight superconducting cavities is passed up to three times, providing a maximal energy gain of 40 MeV on each passage. Due to this recirculating layout it is complicated to find an accurate setup for the various beam line elements. Fast online beam dynamics simulations can advantageously assist the operators because they provide a more detailed insight into the actual machine status. In this contribution further developments of the moment based simulation tool V-code which enables to simulate recirculating machines are presented together with simulation results.
LACAN Code for global simulation of SILVA laser isotope separation process
International Nuclear Information System (INIS)
Quaegebeur, J.P.; Goldstein, S.
1991-01-01
Functions used for the definition of a SILVA separator require quite a lot of dimensional and operating parameters. Sizing a laser isotope separation plant needs the determination of these parameters for optimization. In the LACAN simulation code, each elementary physical process is described by simplified models. An example is given for a uranium isotope separation plant whose separation power is optimized with 6 parameters [fr
A quick and easy improvement of Monte Carlo codes for simulation
Lebrere, A.; Talhi, R.; Tripathy, M.; Pyée, M.
The simulation of trials of independent random variables of given distribution is a critical element of running Monte-Carlo codes. This is usually performed by using pseudo-random number generators (and in most cases linearcongruential ones). We present here an alternative way to generate sequences with given statistical properties. This sequences are purely deterministic and are given by closed formulae, and can give in some cases better results than classical generators.
Verification of simulation model with COBRA-IIIP code by confrontment of experimental results
International Nuclear Information System (INIS)
Silva Galetti, M.R. da; Pontedeiro, A.C.; Oliveira Barroso, A.C. de
1985-01-01
It is presented an evaluation of the COBRA IIIP/MIT code (of thermal hydraulic analysis by subchannels), comparing their results with experimental data obtained in stationary and transient regimes. It was done a study to calculate the spatial and temporal critical heat flux. It is presented a sensitivity study of simulation model related to the turbulent mixture and the number of axial intervals. (M.C.K.) [pt
Simulation calculations using the code Geant III for the EUROGAM device
Energy Technology Data Exchange (ETDEWEB)
Beck, F A; Curien, D; Duchene, G; France, G de; Wei, L [Strasbourg-1 Univ., 67 (France). Centre de Recherches Nucleaires
1992-08-01
Simulation calculations are good tools to determine, at a low cost, the characteristics of a detector. It enables to change the geometry of the counter in an iterative way to optimize its response leading to the best performances for the whole multi-detector device. This kind of calculations have been performed using the Geant III code for the EUROGAM device. (author). 3 tabs., 5 figs.
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)
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.
Massey, J. L.
1976-01-01
The very low error probability obtained with long error-correcting codes results in a very small number of observed errors in simulation studies of practical size and renders the usual confidence interval techniques inapplicable to the observed error probability. A natural extension of the notion of a 'confidence interval' is made and applied to such determinations of error probability by simulation. An example is included to show the surprisingly great significance of as few as two decoding errors in a very large number of decoding trials.
Monte Carlo simulation of dose calculation in voxel and geometric phantoms using GEANT4 code
International Nuclear Information System (INIS)
Martins, Maximiano C.; Santos, Denison de S.; Queiroz Filho, Pedro P. de; Silva, Rosana de S. e; Begalli, Marcia
2009-01-01
Monte Carlo simulation techniques have become a valuable tool for scientific purposes. In radiation protection many quantities are obtained by means of the simulation of particles passing through human body models, also known as phantoms, allowing the calculation of doses deposited in an individual's organs exposed to ionizing radiation. These information are very useful from the medical viewpoint, as they are used in the planning of external beam radiotherapy and brachytherapy treatments. The goal of this work is the implementation of a voxel phantom and a geometrical phantom in the framework of the Geant4 tool kit, aiming at a future use of this code by professionals in the medical area. (author)
Numerical simulation of a short RFQ resonator using the MAFIA codes
International Nuclear Information System (INIS)
Wang, H.; Ben-Zvi, I.; Jain, A.; Paul, P.; Lombardi, A.
1991-01-01
The electrical characteristics of a short (2βλ=0.4 m) resonator with large modulation (m=4) have been studied using the three dimensional codes, MAFIA. The complete resonator, including the modulated electrodes and a complex support structure, has been simulated using ∼ 350,000 mesh points. Important characteristics studied include the resonant frequency, electric and magnetic fields distributions, quality factor and stored energy. The results of the numerical simulations are compared with the measurements of an actual resonator and analytical approximations. 7 refs., 3 figs., 1 tab
Particle-in-Cell Code BEAMPATH for Beam Dynamics Simulations in Linear Accelerators and Beamlines
International Nuclear Information System (INIS)
Batygin, Y.
2004-01-01
A code library BEAMPATH for 2 - dimensional and 3 - dimensional space charge dominated beam dynamics study in linear particle accelerators and beam transport lines is developed. The program is used for particle-in-cell simulation of axial-symmetric, quadrupole-symmetric and z-uniform beams in a channel containing RF gaps, radio-frequency quadrupoles, multipole lenses, solenoids and bending magnets. The programming method includes hierarchical program design using program-independent modules and a flexible combination of modules to provide the most effective version of the structure for every specific case of simulation. Numerical techniques as well as the results of beam dynamics studies are presented
Particle-in-Cell Code BEAMPATH for Beam Dynamics Simulations in Linear Accelerators and Beamlines
Energy Technology Data Exchange (ETDEWEB)
Batygin, Y.
2004-10-28
A code library BEAMPATH for 2 - dimensional and 3 - dimensional space charge dominated beam dynamics study in linear particle accelerators and beam transport lines is developed. The program is used for particle-in-cell simulation of axial-symmetric, quadrupole-symmetric and z-uniform beams in a channel containing RF gaps, radio-frequency quadrupoles, multipole lenses, solenoids and bending magnets. The programming method includes hierarchical program design using program-independent modules and a flexible combination of modules to provide the most effective version of the structure for every specific case of simulation. Numerical techniques as well as the results of beam dynamics studies are presented.
YT: A Multi-Code Analysis Toolkit for Astrophysical Simulation Data
Energy Technology Data Exchange (ETDEWEB)
Turk, Matthew J.; /San Diego, CASS; Smith, Britton D.; /Michigan State U.; Oishi, Jeffrey S.; /KIPAC, Menlo Park /Stanford U., Phys. Dept.; Skory, Stephen; Skillman, Samuel W.; /Colorado U., CASA; Abel, Tom; /KIPAC, Menlo Park /Stanford U., Phys. Dept.; Norman, Michael L.; /aff San Diego, CASS
2011-06-23
The analysis of complex multiphysics astrophysical simulations presents a unique and rapidly growing set of challenges: reproducibility, parallelization, and vast increases in data size and complexity chief among them. In order to meet these challenges, and in order to open up new avenues for collaboration between users of multiple simulation platforms, we present yt (available at http://yt.enzotools.org/) an open source, community-developed astrophysical analysis and visualization toolkit. Analysis and visualization with yt are oriented around physically relevant quantities rather than quantities native to astrophysical simulation codes. While originally designed for handling Enzo's structure adaptive mesh refinement data, yt has been extended to work with several different simulation methods and simulation codes including Orion, RAMSES, and FLASH. We report on its methods for reading, handling, and visualizing data, including projections, multivariate volume rendering, multi-dimensional histograms, halo finding, light cone generation, and topologically connected isocontour identification. Furthermore, we discuss the underlying algorithms yt uses for processing and visualizing data, and its mechanisms for parallelization of analysis tasks.
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)
yt: A MULTI-CODE ANALYSIS TOOLKIT FOR ASTROPHYSICAL SIMULATION DATA
International Nuclear Information System (INIS)
Turk, Matthew J.; Norman, Michael L.; Smith, Britton D.; Oishi, Jeffrey S.; Abel, Tom; Skory, Stephen; Skillman, Samuel W.
2011-01-01
The analysis of complex multiphysics astrophysical simulations presents a unique and rapidly growing set of challenges: reproducibility, parallelization, and vast increases in data size and complexity chief among them. In order to meet these challenges, and in order to open up new avenues for collaboration between users of multiple simulation platforms, we present yt (available at http://yt.enzotools.org/) an open source, community-developed astrophysical analysis and visualization toolkit. Analysis and visualization with yt are oriented around physically relevant quantities rather than quantities native to astrophysical simulation codes. While originally designed for handling Enzo's structure adaptive mesh refinement data, yt has been extended to work with several different simulation methods and simulation codes including Orion, RAMSES, and FLASH. We report on its methods for reading, handling, and visualizing data, including projections, multivariate volume rendering, multi-dimensional histograms, halo finding, light cone generation, and topologically connected isocontour identification. Furthermore, we discuss the underlying algorithms yt uses for processing and visualizing data, and its mechanisms for parallelization of analysis tasks.
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
A wide-range model of two-group gross sections in the dynamics code HEXTRAN
International Nuclear Information System (INIS)
Kaloinen, E.; Peltonen, J.
2002-01-01
In dynamic analyses the thermal hydraulic conditions within the reactor core may have a large variation, which sets a special requirement on the modeling of cross sections. The standard model in the dynamics code HEXTRAN is the same as in the static design code HEXBU-3D/MODS. It is based on a linear and second order fitting of two-group cross sections on fuel and moderator temperature, moderator density and boron density. A new, wide-range model of cross sections developed in Fortum Nuclear Services for HEXBU-3D/MOD6 has been included as an option into HEXTRAN. In this model the nodal cross sections are constructed from seven state variables in a polynomial of more than 40 terms. Coefficients of the polynomial are created by a least squares fitting to the results of a large number of fuel assembly calculations. Depending on the choice of state variables for the spectrum calculations, the new cross section model is capable to cover local conditions from cold zero power to boiling at full power. The 5. dynamic benchmark problem of AER is analyzed with the new option and results are compared to calculations with the standard model of cross sections in HEXTRAN (Authors)
Medland, A. J.; Zhu, Guowang; Gao, Jian; Sun, Jian
1996-03-01
Feature conversion, also called feature transformation and feature mapping, is defined as the process of converting features from one view of an object to another view of the object. In a relatively simple implementation, for each application the design features are automatically converted into features specific for that application. All modifications have to be made via the design features. This is the approach that has attracted most attention until now. In the ideal situation, however, conversions directly from application views to the design view, and to other applications views, are also possible. In this paper, some difficulties faced in feature conversion are discussed. A new representation scheme of feature-based parts models has been proposed for the purpose of one-way feature conversion. The parts models consist of five different levels of abstraction, extending from an assembly level and its attributes, single parts and their attributes, single features and their attributes, one containing the geometric reference element and finally one for detailed geometry. One implementation of feature conversion for rotational components within GT (Group Technology) has already been undertaken using an automated coding procedure operating on a design-feature database. This database has been generated by a feature-based design system, and the GT coding scheme used in this paper is a specific scheme created for a textile machine manufacturing plant. Such feature conversion techniques presented here are only in their early stages of development and further research is underway.
High-fidelity simulation among bachelor students in simulation groups and use of different roles.
Thidemann, Inger-Johanne; Söderhamn, Olle
2013-12-01
Cost limitations might challenge the use of high-fidelity simulation as a teaching-learning method. This article presents the results of a Norwegian project including two simulation studies in which simulation teaching and learning were studied among students in the second year of a three-year bachelor nursing programme. The students were organised into small simulation groups with different roles; nurse, physician, family member and observer. Based on experiences in different roles, the students evaluated the simulation design characteristics and educational practices used in the simulation. In addition, three simulation outcomes were measured; knowledge (learning), Student Satisfaction and Self-confidence in Learning. The simulation was evaluated to be a valuable teaching-learning method to develop professional understanding and insight independent of roles. Overall, the students rated the Student Satisfaction and Self-confidence in Learning as high. Knowledge about the specific patient focus increased after the simulation activity. Students can develop practical, communication and collaboration skills, through experiencing the nurse's role. Assuming the observer role, students have the potential for vicarious learning, which could increase the learning value. Both methods of learning (practical experience or vicarious learning) may bridge the gap between theory and practice and contribute to the development of skills in reflective and critical thinking. Copyright © 2012 Elsevier Ltd. All rights reserved.
Schnabel, M; Mann, D; Efe, T; Schrappe, M; V Garrel, T; Gotzen, L; Schaeg, M
2004-10-01
The introduction of the German Diagnostic Related Groups (D-DRG) system requires redesigning administrative patient management strategies. Wrong coding leads to inaccurate grouping and endangers the reimbursement of treatment costs. This situation emphasizes the roles of documentation and coding as factors of economical success. The aims of this study were to assess the quantity and quality of initial documentation and coding (ICD-10 and OPS-301) and find operative strategies to improve efficiency and strategic means to ensure optimal documentation and coding quality. In a prospective study, documentation and coding quality were evaluated in a standardized way by weekly assessment. Clinical data from 1385 inpatients were processed for initial correctness and quality of documentation and coding. Principal diagnoses were found to be accurate in 82.7% of cases, inexact in 7.1%, and wrong in 10.1%. Effects on financial returns occurred in 16%. Based on these findings, an optimized, interdisciplinary, and multiprofessional workflow on medical documentation, coding, and data control was developed. Workflow incorporating regular assessment of documentation and coding quality is required by the DRG system to ensure efficient accounting of hospital services. Interdisciplinary and multiprofessional cooperation is recognized to be an important factor in establishing an efficient workflow in medical documentation and coding.
Epp: A C++ EGSnrc user code for x-ray imaging and scattering simulations
International Nuclear Information System (INIS)
Lippuner, Jonas; Elbakri, Idris A.; Cui Congwu; Ingleby, Harry R.
2011-01-01
Purpose: Easy particle propagation (Epp) is a user code for the EGSnrc code package based on the C++ class library egspp. A main feature of egspp (and Epp) is the ability to use analytical objects to construct simulation geometries. The authors developed Epp to facilitate the simulation of x-ray imaging geometries, especially in the case of scatter studies. While direct use of egspp requires knowledge of C++, Epp requires no programming experience. Methods: Epp's features include calculation of dose deposited in a voxelized phantom and photon propagation to a user-defined imaging plane. Projection images of primary, single Rayleigh scattered, single Compton scattered, and multiple scattered photons may be generated. Epp input files can be nested, allowing for the construction of complex simulation geometries from more basic components. To demonstrate the imaging features of Epp, the authors simulate 38 keV x rays from a point source propagating through a water cylinder 12 cm in diameter, using both analytical and voxelized representations of the cylinder. The simulation generates projection images of primary and scattered photons at a user-defined imaging plane. The authors also simulate dose scoring in the voxelized version of the phantom in both Epp and DOSXYZnrc and examine the accuracy of Epp using the Kawrakow-Fippel test. Results: The results of the imaging simulations with Epp using voxelized and analytical descriptions of the water cylinder agree within 1%. The results of the Kawrakow-Fippel test suggest good agreement between Epp and DOSXYZnrc. Conclusions: Epp provides the user with useful features, including the ability to build complex geometries from simpler ones and the ability to generate images of scattered and primary photons. There is no inherent computational time saving arising from Epp, except for those arising from egspp's ability to use analytical representations of simulation geometries. Epp agrees with DOSXYZnrc in dose calculation, since
Holmes, Shawn Yvette
A simulation was created to emulate two Racial Ethical Sensitivity Test (REST) videos (Brabeck et al., 2000). The REST is a reliable assessment for ethical sensitivity to racial and gender intolerant behaviors in educational settings. Quantitative and qualitative analysis of the REST was performed using the Quick-REST survey and an interview protocol. The purpose of this study was to affect science educator ability to recognize instances of racial and gender intolerant behaviors by levering immersive qualities of simulations. The fictitious Hazelton High School virtual environment was created by the researcher and compared with the traditional REST. The study investigated whether computer simulations can influence the ethical sensitivity of preservice and inservice science teachers to racial and gender intolerant behaviors in school settings. The post-test only research design involved 32 third-year science education students enrolled in science education classes at several southeastern universities and 31 science teachers from the same locale, some of which were part of an NSF project. Participant samples were assigned to the video control group or the simulation experimental group. This resulted in four comparison group; preservice video, preservice simulation, inservice video and inservice simulation. Participants experienced two REST scenarios in the appropriate format then responded to Quick-REST survey questions for both scenarios. Additionally, the simulation groups answered in-simulation and post-simulation questions. Nonparametric analysis of the Quick-REST ascertained differences between comparison groups. Cronbach's alpha was calculated for internal consistency. The REST interview protocol was used to analyze recognition of intolerant behaviors in the in-simulation prompts. Post-simulation prompts were analyzed for emergent themes concerning effect of the simulation on responses. The preservice video group had a significantly higher mean rank score than
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.)
Design and implementation of a software tool intended for simulation and test of real time codes
International Nuclear Information System (INIS)
Le Louarn, C.
1986-09-01
The objective of real time software testing is to show off processing errors and unobserved functional requirements or timing constraints in a code. In the perspective of safety analysis of nuclear equipments of power plants testing should be carried independently from the physical process (which is not generally available), and because casual hardware failures must be considered. We propose here a simulation and test tool, integrally software, with large interactive possibilities for testing assembly code running on microprocessor. The OST (outil d'aide a la simulation et au Test de logiciels temps reel) simulates code execution and hardware or software environment behaviour. Test execution is closely monitored and many useful informations are automatically saved. The present thesis work details, after exposing methods and tools dedicated to real time software, the OST system. We show the internal mechanisms and objects of the system: particularly ''events'' (which describe evolutions of the system under test) and mnemonics (which describe the variables). Then, we detail the interactive means available to the user for constructing the test data and the environment of the tested software. Finally, a prototype implementation is presented along with the results of the tests carried out. This demonstrates the many advantages of the use of an automatic tool over a manual investigation. As a conclusion, further developments, nececessary to complete the final tool are rewieved [fr
Energy Technology Data Exchange (ETDEWEB)
Kida, Takashi; Umeda, Miki; Sugikawa, Susumu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2003-03-01
MOX dissolution using silver-mediated electrochemical method will be employed for the preparation of plutonium nitrate solution in the criticality safety experiments in the Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF). A simulation code for the MOX dissolution has been developed for the operating support. The present report describes the outline of the simulation code, a comparison with the experimental data and a parameter study on the MOX dissolution. The principle of this code is based on the Zundelevich's model for PuO{sub 2} dissolution using Ag(II). The influence of nitrous acid on the material balance of Ag(II) is taken into consideration and the surface area of MOX powder is evaluated by particle size distribution in this model. The comparison with experimental data was carried out to confirm the validity of this model. It was confirmed that the behavior of MOX dissolution could adequately be simulated using an appropriate MOX dissolution rate constant. It was found from the result of parameter studies that MOX particle size was major governing factor on the dissolution rate. (author)
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
Simulations of inspiraling and merging double neutron stars using the Spectral Einstein Code
Haas, Roland; Ott, Christian D.; Szilagyi, Bela; Kaplan, Jeffrey D.; Lippuner, Jonas; Scheel, Mark A.; Barkett, Kevin; Muhlberger, Curran D.; Dietrich, Tim; Duez, Matthew D.; Foucart, Francois; Pfeiffer, Harald P.; Kidder, Lawrence E.; Teukolsky, Saul A.
2016-06-01
We present results on the inspiral, merger, and postmerger evolution of a neutron star-neutron star (NSNS) system. Our results are obtained using the hybrid pseudospectral-finite volume Spectral Einstein Code (SpEC). To test our numerical methods, we evolve an equal-mass system for ≈22 orbits before merger. This waveform is the longest waveform obtained from fully general-relativistic simulations for NSNSs to date. Such long (and accurate) numerical waveforms are required to further improve semianalytical models used in gravitational wave data analysis, for example, the effective one body models. We discuss in detail the improvements to SpEC's ability to simulate NSNS mergers, in particular mesh refined grids to better resolve the merger and postmerger phases. We provide a set of consistency checks and compare our results to NSNS merger simulations with the independent bam code. We find agreement between them, which increases confidence in results obtained with either code. This work paves the way for future studies using long waveforms and more complex microphysical descriptions of neutron star matter in SpEC.
Korall, Petra; Pryer, Kathleen M; Metzgar, Jordan S; Schneider, Harald; Conant, David S
2006-06-01
Tree ferns are a well-established clade within leptosporangiate ferns. Most of the 600 species (in seven families and 13 genera) are arborescent, but considerable morphological variability exists, spanning the giant scaly tree ferns (Cyatheaceae), the low, erect plants (Plagiogyriaceae), and the diminutive endemics of the Guayana Highlands (Hymenophyllopsidaceae). In this study, we investigate phylogenetic relationships within tree ferns based on analyses of four protein-coding, plastid loci (atpA, atpB, rbcL, and rps4). Our results reveal four well-supported clades, with genera of Dicksoniaceae (sensu ) interspersed among them: (A) (Loxomataceae, (Culcita, Plagiogyriaceae)), (B) (Calochlaena, (Dicksonia, Lophosoriaceae)), (C) Cibotium, and (D) Cyatheaceae, with Hymenophyllopsidaceae nested within. How these four groups are related to one other, to Thyrsopteris, or to Metaxyaceae is weakly supported. Our results show that Dicksoniaceae and Cyatheaceae, as currently recognised, are not monophyletic and new circumscriptions for these families are needed.
International Nuclear Information System (INIS)
Kim, Sung Hyun; Moon, Chan Ki; Park, Sung Baek; Na, Man Gyun
2013-01-01
The RELAP5/MOD3 code introduced in cooperation with U. S. NRC has been utilized mainly for validation calculation of accident analysis submitted by licensee in Korea. The Korea Institute of Nuclear Safety has built a verification system of LWR accident analysis with RELAP5/MOD3 code engine. Therefore, the simulator replicates the design basis accident and its results are compared with RELAP5/MOD3 code results that will have important implications in the verification of the simulator in the future. The SGTR simulations were performed by the simulator and its results were compared with ones by RELAP5/MOD3 code in this study. Thus, the results of this study can be used as materials to build the verification system of the nuclear power plant simulator. We tried to compare with RELAP5/MOD3 verification code by replicating major parameters of steam generator tube rupture using the simulator for OPR-1000 in Yonggwang training center. By comparing the changes in temperature, pressure and inventory of the reactor coolant system and main steam system during the SGTR, it was confirmed that the main behaviors of SGTR which the simulator and RELAP5/MOD3 code showed are similar. However, the behavior of SG pressure and level that are important parameters to diagnose the accident were a little different. We estimated that RELAP5/MOD3 code was not reflected the major control systems in detail, such as FWCS, SBCS and PPCS. The different behaviors of SG level and pressure in this study should be needed an additional review. As a result of the comparison, the major simulation parameters behavior by RELAP5/MOD3 code agreed well with the one by the simulator. Therefore, it is thought that RELAP5/MOD3 code is used as a tool for validation of NPP simulator in the near future through this study
Plasma burn-through simulations using the DYON code and predictions for ITER
International Nuclear Information System (INIS)
Kim, Hyun-Tae; Sips, A C C; De Vries, P C
2013-01-01
This paper will discuss simulations of the full ionization process (i.e. plasma burn-through), fundamental to creating high temperature plasma. By means of an applied electric field, the gas is partially ionized by the electron avalanche process. In order for the electron temperature to increase, the remaining neutrals need to be fully ionized in the plasma burn-through phase, as radiation is the main contribution to the electron power loss. The radiated power loss can be significantly affected by impurities resulting from interaction with the plasma facing components. The DYON code is a plasma burn-through simulator developed at Joint European Torus (JET) (Kim et al and EFDA-JET Contributors 2012 Nucl. Fusion 52 103016, Kim, Sips and EFDA-JET Contributors 2013 Nucl. Fusion 53 083024). The dynamic evolution of the plasma temperature and plasma densities including the impurity content is calculated in a self-consistent way using plasma wall interaction models. The recent installation of a beryllium wall at JET enabled validation of the plasma burn-through model in the presence of new, metallic plasma facing components. The simulation results of the plasma burn-through phase show a consistent good agreement against experiments at JET, and explain differences observed during plasma initiation with the old carbon plasma facing components. In the International Thermonuclear Experimental Reactor (ITER), the allowable toroidal electric field is restricted to 0.35 (V m −1 ), which is significantly lower compared to the typical value (∼1 (V m −1 )) used in the present devices. The limitation on toroidal electric field also reduces the range of other operation parameters during plasma formation in ITER. Thus, predictive simulations of plasma burn-through in ITER using validated model is of crucial importance. This paper provides an overview of the DYON code and the validation, together with new predictive simulations for ITER using the DYON code. (paper)
SPACE code simulation of ATLAS DVI line break accident test (SB DVI 08 Test)
Energy Technology Data Exchange (ETDEWEB)
Lim, Sang Gyu [KHNP, Daejeon (Korea, Republic of)
2012-10-15
APR1400 has adopted new safety design features which are 4 mechanically independent DVI (Direct Vessel Injection) systems and fluidic device in the safety injection tanks (SITs). Hence, DVI line break accident has to be evaluated as one of the small break LOCA (SBLOCA) to ensure the safety of APR1400. KAERI has been performed for DVI line break test (SB DVI 08) using ATLAS (Advanced Thermal Hydraulic Test Loop for Accident Simulation) facility which is an integral effect test facility for APR1400. The test result shows that the core collapsed water level decreased before a loop seal clearance, so that a core uncover occurred. At this time, the peak cladding temperature (PCT) is rapidly increased even though the emergency core cooling (ECC) water is injected from safety injection pump (SIP). This test result is useful for supporting safety analysis using thermal hydraulic safety analysis code and increases the understanding of SBLOCA phenomena in APR1400. The SBLOCA evaluation methodology for APR1400 is now being developed using SPACE code. The object of the development of this methodology is to set up a conservative evaluation methodology in accordance with appendix K of 10 CFR 50. ATLAS SB DVI 08 test is selected for the evaluation of SBLOCA methodology using SPACE code. Before applying the conservative models and correlations, benchmark calculation of the test is performed with the best estimate models and correlations to verify SPACE code capability. This paper deals with benchmark calculations results of ATLAS SB DVI 08 test. Calculation results of the major hydraulics variables are compared with measured data. Finally, this paper carries out the SPACE code performances for simulating the integral effect test of SBLOCA.
DeLyser, Dydia; Potter, Amy E.
2013-01-01
This article describes experiential-learning approaches to conveying the work and rewards involved in qualitative research. Seminar students interviewed one another, transcribed or took notes on those interviews, shared those materials to create a set of empirical materials for coding, developed coding schemes, and coded the materials using those…
49 CFR 173.52 - Classification codes and compatibility groups of explosives.
2010-10-01
... containing both an explosive substance and flammable liquid or gel J 1.1J1.2J 1.3J Article containing both an... classification codes for substances and articles described in the first column of table 1. Table 2 shows the... possible classification codes for explosives. Table 1—Classification Codes Description of substances or...
International Nuclear Information System (INIS)
Lee, Joon H.; Siegel, Malcolm Dean; Arguello, Jose Guadalupe Jr.; Webb, Stephen Walter; Dewers, Thomas A.; Mariner, Paul E.; Edwards, Harold Carter; Fuller, Timothy J.; Freeze, Geoffrey A.; Jove-Colon, Carlos F.; Wang, Yifeng
2011-01-01
This report describes a gap analysis performed in the process of developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with rigorous verification, validation, and software quality requirements. The gap analyses documented in this report were are performed during an initial gap analysis to identify candidate codes and tools to support the development and integration of the Waste IPSC, and during follow-on activities that delved into more detailed assessments of the various codes that were acquired, studied, and tested. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. The gap analysis indicates that significant capabilities may already exist in the existing THC codes although there is no single code able to fully account for all physical and chemical processes involved in a waste disposal system. Large gaps exist in modeling chemical processes and their couplings with other processes. The coupling of chemical processes with flow transport and mechanical deformation remains challenging. The data for extreme environments (e.g., for elevated temperature and high ionic strength media) that are
Energy Technology Data Exchange (ETDEWEB)
Lee, Joon H.; Siegel, Malcolm Dean; Arguello, Jose Guadalupe, Jr.; Webb, Stephen Walter; Dewers, Thomas A.; Mariner, Paul E.; Edwards, Harold Carter; Fuller, Timothy J.; Freeze, Geoffrey A.; Jove-Colon, Carlos F.; Wang, Yifeng
2011-03-01
This report describes a gap analysis performed in the process of developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with rigorous verification, validation, and software quality requirements. The gap analyses documented in this report were are performed during an initial gap analysis to identify candidate codes and tools to support the development and integration of the Waste IPSC, and during follow-on activities that delved into more detailed assessments of the various codes that were acquired, studied, and tested. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. The gap analysis indicates that significant capabilities may already exist in the existing THC codes although there is no single code able to fully account for all physical and chemical processes involved in a waste disposal system. Large gaps exist in modeling chemical processes and their couplings with other processes. The coupling of chemical processes with flow transport and mechanical deformation remains challenging. The data for extreme environments (e.g., for elevated temperature and high ionic strength media) that are
Simulations of corrosion product transfer with the OSCAR V1.2 code
International Nuclear Information System (INIS)
Dacquait, F.; Francescatto, J.; Broutin, F.; Genin, J.B.; Benier, G.; Girard, M.; You, D.; Ranchoux, G.; Bonnefon, J.; Bachet, M.; Riot, G.
2012-09-01
Activated Corrosion Products (ACPs) generate a radiation field in PWRs, which is the major contributor to the dose absorbed by nuclear power plant staff working during shutdown operations and maintenance. Therefore, a thorough understanding of the mechanisms that control the corrosion product transfer is of the highest importance. Since the 1970's, the R and D strategy in France has been based on experiments in test loops representative of PWR conditions, on in-situ gamma spectrometry measurements of the PWR primary system contamination and on simulation code development. The simulation of corrosion product transfers in PWR primary circuits is a major challenge since it involves many physical and chemical phenomena including: corrosion, dissolution, precipitation, erosion, deposition, convection, activation... In addition to the intrinsic difficulty of multi-physics modelling, the primary systems present severe operating conditions (300 deg. C, 150 bar, neutron flux, fluid velocity up to 15 m.s -1 and very low corrosion product concentrations). The purpose of the OSCAR code, developed by the CEA in cooperation with EDF and AREVA NP, is to predict the PWR primary system contamination by corrosion and fission products. The OSCAR code is considered to be not only a tool for numerical simulations and predictions (operational practices improvements and new-built PWRs design) but also one that might combine and organise all new knowledge useful to progress on contamination. The OSCAR code for Products of Corrosion, OSCAR PC, allows researchers to analyse the corrosion product behaviour and to calculate the ACP volume and surface activities of the primary and auxiliary systems. In the new version, OSCAR PC V1.2, the corrosion product transfer in the particulate form is enhanced and a new feature is the possibility to simulate cold shutdowns. In order to validate this version, the contamination transfer has been simulated in 5 French PWRs with different operating and
Energy Technology Data Exchange (ETDEWEB)
Kim, Myung Hyun [Kyunghee University, Seoul (Korea, Republic of)
1996-07-01
A development project of 3-dimensional kinetics code for ALMR has three level of works. In the first level, a multi-group, nodal kinetics code for the HEX-Z geometry has been developed. A code showed very good results for the static analysis as well as for the kinetics problems. At the second level, a core thermal-hydraulic analysis code was developed for the temperature feedback calculation in ALMR transients analysis. This code is coupled with kinetics code. A sodium property table was programmed and tested to the KAERI data and thermal feedback model was developed and coupled in code. Benchmarking of T/H calculation has been performed and showed fairly good results. At the third level of research work, reactivity feedback model for structure thermal expansion is developed and added to the code. At present, basic model was studied. However, code development in now on going. Benchmarking of this model developed can not be done because of lack of data. 31 refs., 17 tabs., 38 figs. (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)
Comparison of Geant4-DNA simulation of S-values with other Monte Carlo codes
International Nuclear Information System (INIS)
André, T.; Morini, F.; Karamitros, M.; Delorme, R.; Le Loirec, C.; Campos, L.; Champion, C.; Groetz, J.-E.; Fromm, M.; Bordage, M.-C.; Perrot, Y.; Barberet, Ph.
2014-01-01
Monte Carlo simulations of S-values have been carried out with the Geant4-DNA extension of the Geant4 toolkit. The S-values have been simulated for monoenergetic electrons with energies ranging from 0.1 keV up to 20 keV, in liquid water spheres (for four radii, chosen between 10 nm and 1 μm), and for electrons emitted by five isotopes of iodine (131, 132, 133, 134 and 135), in liquid water spheres of varying radius (from 15 μm up to 250 μm). The results have been compared to those obtained from other Monte Carlo codes and from other published data. The use of the Kolmogorov–Smirnov test has allowed confirming the statistical compatibility of all simulation results
Simulations of the neutronic REP behaviour using the codes DRAGON/DONJON
International Nuclear Information System (INIS)
Le Mer, J.
2007-01-01
Neutron flux calculation is necessary to understand how a nuclear reactor works. This flux is derived from the transport equation on the whole core. Because of its really complex structure and the angular dependence of the transport equation, it is impossible to compute the flux directly and several neutronic calculation codes must be used to solve the equation for different discretizations which require different modelisations. This chain of successive models, known as a calculation scheme, compute the neutron flux of a reactor from its geometry, its isotopic compositions and a cross-section library. Pressurised light Water Reactor (PWR) are the most common nuclear reactor used today. It is necessary for each neutronic code to be validated for this type of reactor. The goal of this work is to create a complete calculation scheme which can be applied to the evolution of the core of a pressurised light water nuclear reactor using the lattice code DRAGON and the reactor code DONJON. Each step of this scheme will be validated by comparisons with other codes or with experimental results. The unit cell calculation will be computed for a benchmark submitted by R. Mosteller. The assembly calculations will be used to compare the results given by DRAGON, APOLLO2 and MCNP for an assembly used by EDF for code testing. The core calculations will show that the codes DRAGON and DONJON can produce accurate macroscopic results for a real core. Those studies will be used to show the effects of many factors on the flux distribution including the cross section library, the number of energy groups, spatial discretization of the unit cell, the tracking model, the self-shielding of the resonant isotopes or the burnup steps. (author)
Simulation of natural convection cooling phenomena for research reactors using the code PARET
International Nuclear Information System (INIS)
Hainoun, A.; Al-Habit, E.
2006-01-01
This study deals with testing the capacity of the code PARET to simulate natural circulation phenomena under different boundary conditions in addition to assessment of some new options related to simulation of control rod movement and the reactivity effect of thermal expansion fuel elements. the experiments of the simple thermal hydraulic loop of Missouri University about natural circulation phenomena in narrow parallel channel were used to validate the code. The results indicate good agreements regarding the evolution of coolant velocity and clad temperature. In particular the heat transfer coefficient of natural convection has been calculated in good agreement with the experiment. On the other hand, the core of MNSR reactor has been modelled to stimulate the reactor dynamic behaviour under natural circulation condition for different initial power level. The observed oscillations during the initial phase vanish gradually with passing time. In this context three experiment of step reactivity insertion were calculated using two different options of boundary conditions, either using initial velocity or pressure drop along the core. The results indicate good agreement with the experiments regarding the evolution of relative power. The validations included also sensitivity analysis against some important parameters like initial velocity and radial distance of fuel rod. The new option for simulation of control rod movement was also tested. For this purpose the MNSR experiment of all control rod withdraw was selected. This means control rod velocity was estimated using experimental measurement. The simulation result of relative power evolution shows good agreement with the experiment during the first phase of the transient. However, an increased deviation is observed in the following phase due to the effect of closed hydrodynamics loop, which can be modelled with the code PARET. (Authors)
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)
BASHAN: A few-group three-dimensional diffusion code with burnup and fuel management features
International Nuclear Information System (INIS)
Pearce, D.F.
1970-12-01
The diffusion equation for a two or three-dimensional, two-group or multi-group downscatter problem is solved by conventional finite difference techniques. An x-y-z geometry is assumed with an 'in-channel' mesh point representation. Options are available which allow representation of a soluble poison dispersed throughout the reactor, and also absorber rods in specified channels. The power distribution and multiplication factor k eff are calculated and a point rating map is used to advance the irradiation at each mesh point by a specified time-step so that burnup is followed. Fuel changes may be made so that radial shuffling and axial shuffling fuel management schemes can be studies. The code has been written in FORTRAN S2 for an IBM 7030 (STRETCH) computer which, with a fast store of 80,000 locations, allows problems of up to 15,000 mesh points to be dealt with. Conversion to FORTRAN IV for IBM 360 has now been completed. (author)
International Nuclear Information System (INIS)
Williamson, R.L.
2011-01-01
Highlights: → The ABAQUS thermomechanics code is enhanced to enable simulation of nuclear fuel behavior. → Comparisons are made between discrete and smeared fuel pellet analysis. → Multidimensional and multipellet analysis is important for accurate prediction of PCMI. → Fully coupled thermomechanics results in very smooth prediction of fuel-clad gap closure. → A smeared-pellet approximation results in significant underprediction of clad radial displacements and plastic strain. - Abstract: A powerful multidimensional fuels performance analysis capability, applicable to both steady and transient fuel behavior, is developed based on enhancements to the commercially available ABAQUS general-purpose thermomechanics code. Enhanced capabilities are described, including: UO 2 temperature and burnup dependent thermal properties, solid and gaseous fission product swelling, fuel densification, fission gas release, cladding thermal and irradiation creep, cladding irradiation growth, gap heat transfer, and gap/plenum gas behavior during irradiation. This new capability is demonstrated using a 2D axisymmetric analysis of the upper section of a simplified multipellet fuel rod, during both steady and transient operation. Comparisons are made between discrete and smeared-pellet simulations. Computational results demonstrate the importance of a multidimensional, multipellet, fully-coupled thermomechanical approach. Interestingly, many of the inherent deficiencies in existing fuel performance codes (e.g., 1D thermomechanics, loose thermomechanical coupling, separate steady and transient analysis, cumbersome pre- and post-processing) are, in fact, ABAQUS strengths.
The priority cases of the FUMEX-III exercises simulated with the TRANSURANUS code
International Nuclear Information System (INIS)
Boneva, S.
2011-01-01
The FUMEX-III project provides a good basis for testing common code priorities and the needs for further developments. The GAIN experiment contains results on four Gd 2 O 3 doped UO 2 rods and offers good opportunities for testing of the fuel performance codes in the case of Gd-doped fuel. A good agreement between the TRANSURANUS calculations and the measurements is achieved for the fuel and the cladding deformation. The FUMEX-III priority cases cover two rods from the GINNA reactor experiment: rod2 with fuel solid pellets, and rod4 with annular pellets and standard Zircaloy-4 cladding. Both rods were irradiated 5 cycles up to 52MWd/kgU. The simulations of the GINNA and US PWR experiments are part of the ongoing validation of the TRANSURANUS code - for different pellet design. The simulations of irradiation transients reveal the need for improving the fission gas release model, including burst release and release from the high burn-up structure
Modification and application of the ATHLET-SC code to trans-critical simulations
International Nuclear Information System (INIS)
Fu, S.-W.; Zhou, C.; Xu, Z.-H.; Liu, X.-J.; Yang, Y.-H.; Cheng, H.
2011-01-01
In the simulation of trans-critical transients of Supercritical water cooled reactor (SCWR), calculation will terminate because of the sudden change in void fraction across the critical point. To solve this problem, a pseudo two-phase method is proposed with a virtual region of latent heat at pseudo-critical temperatures. A smooth variation of void fraction can be realized by using liquid-field conservation equations at temperatures lower than the pseudo-critical temperature, and vapor-field conservation equations at temperatures higher than the pseudo-critical temperature. Using this method, the system code ATHLET is modified to ATHLET-SC mod 2 on the basic of the previous modified version ATHLET-SC by Shanghai Jiao Tong University. The results of tests are verified that the calculation error with the pseudo two-phase method for supercritical fluid is acceptable, when the virtual region of latent heat is kept small. Moreover, the ATHLET-SC mod 2 code is used to simulate the pressurization and depressurization process of a single flow channel with the pressure transition as well as blowdown process. The results indicate a good applicability of the modified code. (author)
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)
BRICTEST: a code for charge breeding simulations in RF quadrupolar field
International Nuclear Information System (INIS)
Variale, V.; Claudione, M.
2005-01-01
In the framework of the SPES project (Study for Production of Exotic Species), funded by Istituto Nazionale Fisica Nucleare (INFN) at the Laboratori Nazionali Legnaro (LNL) (Padua) for Radioactive Ion Beam (RIB) production, an R and D experiment of a charge breeder device, called BRIC (BReeding Ion Charge), is in progress at LNL. BRIC is an Electron Beam Ion Source (EBIS) type ion charge state breeder in which a radio frequency (RF) quadrupolar field has been superimposed in the trapped ion region to introduce a selective containment with the aim of increasing the wanted ion trapping efficiency. A code that studies the motion and the ion charge state evolution in the trap region of the BRIC device has been recently developed in the Bari INFN section. That code has the aim of showing if, in the presence of an axial magnetic field and electron beam space charge force, the RF quadrupole field can still give a selective ion containment in the EBIS trap region. The code, furthermore, should allow choosing the RF quadrupole parameters to optimize the ion charge containment efficiency. In this paper the main feature of the code, named BRICTEST, and the simulation test will be presented and shortly discussed
Look at nuclear artillery yield options using JANUS, a wargame simulation code
International Nuclear Information System (INIS)
Andre, C.G.
1982-01-01
JANUS, a two-sided, interactive wargame simulation code, was used to explore how using each of several different yield options in a nuclear artillery shell might affect a tactical battlefield simulation. In a general sense, the results or outcomes of these simulations support the results or outcomes of previous studies. In these simulations the Red player knew of the anticipated nuclear capability of the Blue player. Neither side experienced a decisive win over the other, and both continued fighting and experienced losses that, under most historical circumstances, would have been termed unacceptable - that is, something else would have happened (the attack would have been called off). During play, each side had only fragmentary knowledge of the remaining resources on the other side - thus each side desired to continue fighting on the basis of known information. We found that the anticipated use of nuclear weapons by either side affects the character of a game significantly and that, if the employment of nuclear weapons is to have a decided effect on the progress and outcome of a battle, each side will have to have an adequate number of nuclear weapons. In almost all the simulations we ran using JANUS, enhanced radiation (ER) weapons were more effective than 1-kt fission weapons in imposing overall losses on Red. The typical visibility in the JANUS simulation limited each side's ability to acquire units deep into enemy territory and so the 10-kt fission weapon was not useful against enemy tanks that were not engaged in battle
International Nuclear Information System (INIS)
Hong, In Seob; Suk, Ho Chun; Kim, Soon Young; Jo, Chang Keun
2002-06-01
The major objective of this research is to validate the incremental cross section property of DRAGON code in connection with WIMS-AECL/DRAGON/RFSP code system with ENDF/B-VI library and full 2G calculation model. The direct comparison between the incremental cross section results calculated by DRAGON with ENDF/B-VI and ENDF/B-V and MULTICELL with ENDF/B-V indicate that there are not much differences between the incremental cross sections of DRAGON with ENDF/B-V and ENDF/B-VI, but there exists large discrepancies between the results of DRAGON and those of MULTICELL. In the analysis of the difference between calculated and measured reactivity worths of various types of control devices during Phase-B Post-Simulation of Wolsong Units 2, 3 and 4, WIMS-AECL/DRAGON/RFSP analysis well agrees with those of previous WIMS-AECL /MULTICELL/RFSP analysis within very small differences. From those results, we can conclude that DRAGON code can be used as a general purpose incremental cross section generation tool for not only the natural uranium fuel but also slightly enriched fuel such as RU or SEU, to cover the shortcomings of natural uranium based MULTICELL code
Energy Technology Data Exchange (ETDEWEB)
Ivanov, Vladimir [Scientific and Engineering Centre for Nuclear and Radiation Safety (SES NRS), Moscow (Russian Federation); Bousquet, Jeremy [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)
2016-11-15
This work aims to assess the capabilities of reactor physics codes (initially validated for thermal reactors) to simulate fast sodium cooled reactors. The BFS-62-3A critical experiment from the BN-600 Hybrid Core Benchmark Analyses was chosen for the investigation. Monte-Carlo codes (KENO from SCALE and SERPENT 2.1.23) and the deterministic diffusion code DYN3D-MG are applied to calculate the neutronic parameters. It was found that the multiplication factor and reactivity effects calculated by KENO and SERPENT using the ENDF/B-VII.0 continuous energy library are in a good agreement with each other and with the measured benchmark values. Few-groups macroscopic cross sections, required for DYN3D-MG, were prepared in applying different methods implemented in SCALE and SERPENT. The DYN3D-MG results of a simplified benchmark show reasonable agreement with results from Monte-Carlo calculations and measured values. The former results are used to justify DYN3D-MG implementation for sodium cooled fast reactors coupled deterministic analysis.
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
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.
CASINO, a code for simulation of charged particles in an axisymmetric Tokamak
International Nuclear Information System (INIS)
Dillner, Oe.
1992-01-01
The present report comprises a documentation of CASINO, a simulation code developed as a means for the study of high energy charged particles in an axisymmetric Tokamak. The background of the need for such a numerical tool is presented. In the description of the numerical model used for the orbit integration, the method using constants of motion, the Lao-Hirsman geometry for the flux surfaces and a method for reducing the necessary number of particles is elucidated. A brief outline of the calculational sequence is given as a flow chart. The essential routines and functions as well as the common blocks are briefly described. The input and output routines are shown. Finally the documentation is completed by a short discussion of possible extensions of the code and a test case. (au)
Simulation of density curve for slim borehole using the Monte Carlo code MCNPX
International Nuclear Information System (INIS)
Souza, Edmilson Monteiro de; Silva, Ademir Xavier da; Lopes, Ricardo Tadeu; Lima, Inaya C.B.; Rocha, Paula L.F.
2010-01-01
Borehole logging for formation density has been an important geophysical measurement in oil industry. For calibration of the Gamma Ray nuclear logging tool, numerous rock models of different lithology and densities are necessary. However, the full success of this calibration process is determined by a reliable benchmark, where the complete and precise chemical composition of the standards is necessary. Simulations using the Monte Carlo MCNP have been widely employed in well logging application once it serves as a low-cost substitute for experimental test pits, as well as a means for obtaining data that are difficult to obtain experimentally. Considering this, the purpose of this work is to use the code MCNP to obtain density curves for slim boreholes using Gamma Ray logging tools. For this, a Slim Density Gamma Probe, named TRISOND R , and a 100 mCi Cs-137 gamma source has been modeled with the new version of MCNP code MCNPX. (author)
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)
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
Simulation of density curve for slim borehole using the Monte Carlo code MCNPX
Energy Technology Data Exchange (ETDEWEB)
Souza, Edmilson Monteiro de; Silva, Ademir Xavier da; Lopes, Ricardo Tadeu, E-mail: emonteiro@nuclear.ufrj.b, E-mail: ademir@nuclear.ufrj.b, E-mail: ricardo@lin.ufrj.b [Coordenacao dos Programas de Pos-Graduacao de Engenharia (COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear; Correa, Samanda Cristine Arruda, E-mail: scorrea@nuclear.ufrj.b [Centro Universitario Estadual da Zona Oeste (CCMAT/UEZO), Rio de Janeiro, RJ (Brazil); Lima, Inaya C.B., E-mail: inaya@lin.ufrj.b [Universidade Estadual do Rio de Janeiro (IPRJ/UERJ) Nova Friburgo, Rio de Janeiro, RJ (Brazil). Instituto Politecnico do Rio de Janeiro; Rocha, Paula L.F., E-mail: ferrucio@acd.ufrj.b [Universidade Federal do Rio de Janeiro (UFRJ) RJ (Brazil). Dept. de Geologia
2010-07-01
Borehole logging for formation density has been an important geophysical measurement in oil industry. For calibration of the Gamma Ray nuclear logging tool, numerous rock models of different lithology and densities are necessary. However, the full success of this calibration process is determined by a reliable benchmark, where the complete and precise chemical composition of the standards is necessary. Simulations using the Monte Carlo MCNP have been widely employed in well logging application once it serves as a low-cost substitute for experimental test pits, as well as a means for obtaining data that are difficult to obtain experimentally. Considering this, the purpose of this work is to use the code MCNP to obtain density curves for slim boreholes using Gamma Ray logging tools. For this, a Slim Density Gamma Probe, named TRISOND{sup R}, and a 100 mCi Cs-137 gamma source has been modeled with the new version of MCNP code MCNPX. (author)
Simulation code for the interaction of 14 MeV neutrons on cells
Energy Technology Data Exchange (ETDEWEB)
Nenot, M.L.; Alard, J.P.; Dionet, C.; Arnold, J.; Tchirkov, A.; Meunier, H.; Bodez, V.; Rapp, M.; Verrelle, P
2002-07-01
The structure of the survival curve of melanoma cells irradiated by 14 MeV neutrons displays unusual features at very low dose rate where a marked increase in cell killings at 0.05 Gy is followed by a plateau for survival from 0.1 to 0.32 Gy. In parallel a simulation code was constructed for the interaction of 14 MeV neutrons with cellular cultures. The code describes the interaction of the neutrons with the atomic nuclei of the cellular medium and of the external medium (flask culture and culture medium), and is used to compute the deposited energy into the cell volume. It was found that the large energy transfer events associated with heavy charged recoil can occur and that a large part of the energy deposition events are due to recoil protons emitted from the external medium. It is suggested that such events could partially explain the experimental results. (author)
Premar-2: a Monte Carlo code for radiative transport simulation in atmospheric environments
Energy Technology Data Exchange (ETDEWEB)
Cupini, E. [ENEA, Centro Ricerche Ezio Clementel, Bologna, (Italy). Dipt. Innovazione
1999-07-01
The peculiarities of the PREMAR-2 code, aimed at radiation transport Monte Carlo simulation in atmospheric environments in the infrared-ultraviolet frequency range, are described. With respect to the previously developed PREMAR code, besides plane multilayers, spherical multilayers and finite sequences of vertical layers, each one with its own atmospheric behaviour, are foreseen in the new code, together with the refraction phenomenon, so that long range, highly slanted paths can now be more faithfully taken into account. A zenithal angular dependence of the albedo coefficient has moreover been introduced. Lidar systems, with spatially independent source and telescope, are allowed again to be simulated, and, in this latest version of the code, sensitivity analyses to be performed. According to this last feasibility, consequences on radiation transport of small perturbations in physical components of the atmospheric environment may be analyze and the related effects on searched results estimated. The availability of a library of physical data (reaction coefficients, phase functions and refraction indexes) is required by the code, providing the essential features of the environment of interest needed of the Monte Carlo simulation. Variance reducing techniques have been enhanced in the Premar-2 code, by introducing, for instance, a local forced collision technique, especially apt to be used in Lidar system simulations. Encouraging comparisons between code and experimental results carried out at the Brasimone Centre of ENEA, have so far been obtained, even if further checks of the code are to be performed. [Italian] Nel presente rapporto vengono descritte le principali caratteristiche del codice di calcolo PREMAR-2, che esegue la simulazione Montecarlo del trasporto della radiazione elettromagnetica nell'atmosfera, nell'intervallo di frequenza che va dall'infrarosso all'ultravioletto. Rispetto al codice PREMAR precedentemente sviluppato, il codice
Study of MHD stability beta limit in LHD by hierarchy integrated simulation code
International Nuclear Information System (INIS)
Sato, M.; Watanabe, K.Y.; Nakamura, Y.
2008-10-01
The beta limit by the ideal MHD instabilities (so-called 'MHD stability beta limit') for helical plasmas is studied by a hierarchy integrated simulation code. A numerical model for the effect of the MHD instabilities is introduced such that the pressure profile is flattened around the rational surface due to the MHD instabilities. The width of the flattening of the pressure gradient is determined from the width of the eigenmode structure of the MHD instabilities. It is assumed that there is the upper limit of the mode number of the MHD instabilities which directly affect the pressure gradient. The upper limit of the mode number is determined using a recent high beta experiment in the Large Helical Device (LHD). The flattening of the pressure gradient is calculated by the transport module in a hierarchy integrated code. The achievable volume averaged beta value in the LHD is expected to be beyond 6%. (author)
Atucha II NPP full scope simulator modelling with the thermal hydraulic code TRACRT
International Nuclear Information System (INIS)
Alonso, Pablo Rey; Ruiz, Jose Antonio; Rivero, Norberto
2011-01-01
In February 2010 NA-SA (Nucleoelectrica Argentina S.A.) awarded Tecnatom the Atucha II full scope simulator project. NA-SA is a public company owner of the Argentinean nuclear power plants. Atucha II is due to enter in operation shortly. Atucha II NPP is a PHWR type plant cooled by the water of the Parana River and has the same design as the Atucha I unit, doubling its power capacity. Atucha II will produce 745 MWe utilizing heavy water as coolant and moderator, and natural uranium as fuel. A plant singular feature is the permanent core refueling. TRAC R T is the first real time thermal hydraulic six-equations code used in the training simulation industry for NSSS modeling. It is the result from adapting to real time the best estimate code TRACG. TRAC R T is based on first principle conservation equations for mass, energy and momentum for liquid and steam phases, with two phase flows under non homogeneous and non equilibrium conditions. At present, it has been successfully implemented in twelve full scope replica simulators in different training centers throughout the world. To ease the modeling task, TRAC R T includes a graphical pre-processing tool designed to optimize this process and alleviate the burden of entering alpha numerical data in an input file. (author)
Hydraulic Simulation of In-vessel Downstream Effect Test Using MARS-KS Code
Energy Technology Data Exchange (ETDEWEB)
Bang, Young Seok; Lee, Joon Soo; Ryu, Seung Hoon [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)
2015-05-15
In-vessel downstream effect test (IDET) has been required to evaluate the effect of debris on long term core cooling following a loss of coolant accident (LOCA) in support of resolution of Generic Safety Issue (GSI) 191. Head loss induced by debris (fiber and particle) accumulated on prototypical fuel assembly (FA) should be compared with the available driving head to the core for the various combinations of LOCA and Emergency Core Cooling System (ECCS) injection. The actual simulation was conducted using MARS-KS code. Also the influence of small difference in gap size which was found in the actual experiment is evaluated using the present model. A simple model to determine the form loss factors of FA and gap in clean state and the debris laden state is discussed based on basic fluid mechanics. Those form loss factors were applied to the hydraulic simulation of a selected IDET using MARS-KS code. The result indicated that the present model can be applied to IDET simulation. The pressure drop influenced by small difference in gap size can be evaluated by the present model with practical assumption.
New electromagnetic particle simulation code for the analysis of spacecraft-plasma interactions
International Nuclear Information System (INIS)
Miyake, Yohei; Usui, Hideyuki
2009-01-01
A novel particle simulation code, the electromagnetic spacecraft environment simulator (EMSES), has been developed for the self-consistent analysis of spacecraft-plasma interactions on the full electromagnetic (EM) basis. EMSES includes several boundary treatments carefully coded for both longitudinal and transverse electric fields to satisfy perfect conductive surface conditions. For the longitudinal component, the following are considered: (1) the surface charge accumulation caused by impinging or emitted particles and (2) the surface charge redistribution, such that the surface becomes an equipotential. For item (1), a special treatment has been adopted for the current density calculated around the spacecraft surface, so that the charge accumulation occurs exactly on the surface. As a result, (1) is realized automatically in the updates of the charge density and the electric field through the current density. Item (2) is achieved by applying the capacity matrix method. Meanwhile, the transverse electric field is simply set to zero for components defined inside and tangential to the spacecraft surfaces. This paper also presents the validation of EMSES by performing test simulations for spacecraft charging and peculiar EM wave modes in a plasma sheath.
Simulation of the spectrum (Co-60), Theratron Equinox, using the code Penelope
International Nuclear Information System (INIS)
Quispe V, N. Y.; Ballon P, C. I.; Vega R, J. L. J.; Santos F, C.
2017-10-01
Using the code Penelope (Penetration and Energy Loss of Positrons and Electrons) V. 2008, the spectrum of the Theratron Equinox cobalt unit, currently used at the Goyeneche Hospital in Arequipa (Peru), was obtained in the radiotherapy service. The Penmain program was used to obtain the spectrum that, together with the PENGEOM package included in the Penelope code, allowed to build complex structures with, in this case, the cobalt unit head essentially comprising the cobalt source and its collimators. The dose-to-depth percentage curves were also obtained in different sizes of irradiated fields of 5 x 5, 10 x 10 and 15 x 15 cm 2 for the cobalt spectrum obtained, in which is observed that there is greater dispersion for fields greater and more time of simulation was needed, being concordance of the results of the simulation, when comparing the experimentally obtained data of the dose with the ionization chamber in a water tank. The spectrum obtained was validated with the data of the ionization chamber in the determination of dose-to-depth percentage curves; it can be used as a reference to optimize the radiotherapy planning system in the simulation with equivalent body materials. (Author)
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/.
2015-11-01
Memorandum Simulation of Weld Mechanical Behavior to Include Welding-Induced Residual Stress and Distortion: Coupling of SYSWELD and Abaqus Codes...Weld Mechanical Behavior to Include Welding-Induced Residual Stress and Distortion: Coupling of SYSWELD and Abaqus Codes by Charles R. Fisher...Welding- Induced Residual Stress and Distortion: Coupling of SYSWELD and Abaqus Codes 5a. CONTRACT NUMBER N/A 5b. GRANT NUMBER N/A 5c
MOND simulation suggests an origin for some peculiarities in the Local Group
Bílek, M.; Thies, I.; Kroupa, P.; Famaey, B.
2018-06-01
Context. The Milky Way (MW) and Andromeda (M 31) galaxies possess rotating planes of satellites. The formation of these planes has not been explained satisfactorily so far. It has been suggested that the MW and M 31 satellites are ancient tidal dwarf galaxies; this might explain their configuration. This suggestion gained support by an analytic backward-calculation of the relative MW-M 31 orbit in the MOND modified dynamics paradigm. The result implied that the galaxies experienced a close flyby 7-11 Gyr ago. Aims: Here we explore the Local Group history in MOND in more detail using a simplified first-ever self-consistent simulation. We describe the features induced by the encounter in the simulation and identify possible real counterparts of these features. Methods: The initial conditions were set to eventually roughly reproduce the observed MW and M 31 masses, effective radii, separation, relative velocity, and disk inclinations. We used the publicly available adaptive-mesh-refinement code Phantom of RAMSES. Results: Matter was transferred from the MW to M 31 along a tidal tail in the simulation. The encounter induced the formation of several structures resembling the peculiarities of the Local Group. Most notably are that 1) a rotating planar structure formed around M 31 from the transferred material. It had a size similar to the observed satellite plane and was oriented edge-on to the simulated MW, just as the real plane. 2) The same structure also resembled the tidal features observed around M 31 by its size and morphology. 3) A warp in the MW developed with an amplitude and orientation similar to that observed. 4) A cloud of particles formed around the simulated MW, with the extent of the actual MW satellite system. The encounter did not end by merging in a Hubble time. The simulated stellar disks also thickened as a result of the encounter. Conclusions: The simulation demonstrated that MOND might explain many peculiarities of the Local Group; this needs to
International Nuclear Information System (INIS)
Kim, Jung Do; Gil, Choong Sub; Lee, Jong Tai; Hwang, Won Guk
1992-01-01
A one-group cross section data base of the ORIGEN2 computer code was developed for research reactor applications. For this, ENDF/B-IV and -V data were processed using the NJOY code system into 69-group data. The burnup dependent weighting spectra for KMRR were calculated with the WIMS-KAERI computer code, and then the 69-group data were collapsed to one-group using the spectra. The ORlGEN2-predicted burnup-dependent actinide compositions of KMRR spent fuel using the newly developed data base show a good agreement with the results of detailed multigroup transport calculation. In addition, the burnup characteristics of KMRR spent fuel was analyzed with the new data base. (Author)
Energy Technology Data Exchange (ETDEWEB)
Kim, Jung Do; Gil, Choong Sub; Lee, Jong Tai [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Hwang, Won Guk [Kyung Hee University, Seoul (Korea, Republic of)
1992-03-01
A one-group cross section data base of the ORIGEN2 computer code was developed for research reactor applications. For this, ENDF/B-IV and -V data were processed using the NJOY code system into 69-group data. The burnup dependent weighting spectra for KMRR were calculated with the WIMS-KAERI computer code, and then the 69-group data were collapsed to one-group using the spectra. The ORlGEN2-predicted burnup-dependent actinide compositions of KMRR spent fuel using the newly developed data base show a good agreement with the results of detailed multigroup transport calculation. In addition, the burnup characteristics of KMRR spent fuel was analyzed with the new data base. (Author).
International Nuclear Information System (INIS)
Nelson, Andrew F.; Wetzstein, M.; Naab, T.
2009-01-01
We continue our presentation of VINE. In this paper, we begin with a description of relevant architectural properties of the serial and shared memory parallel computers on which VINE is intended to run, and describe their influences on the design of the code itself. We continue with a detailed description of a number of optimizations made to the layout of the particle data in memory and to our implementation of a binary tree used to access that data for use in gravitational force calculations and searches for smoothed particle hydrodynamics (SPH) neighbor particles. We describe the modifications to the code necessary to obtain forces efficiently from special purpose 'GRAPE' hardware, the interfaces required to allow transparent substitution of those forces in the code instead of those obtained from the tree, and the modifications necessary to use both tree and GRAPE together as a fused GRAPE/tree combination. We conclude with an extensive series of performance tests, which demonstrate that the code can be run efficiently and without modification in serial on small workstations or in parallel using the OpenMP compiler directives on large-scale, shared memory parallel machines. We analyze the effects of the code optimizations and estimate that they improve its overall performance by more than an order of magnitude over that obtained by many other tree codes. Scaled parallel performance of the gravity and SPH calculations, together the most costly components of most simulations, is nearly linear up to at least 120 processors on moderate sized test problems using the Origin 3000 architecture, and to the maximum machine sizes available to us on several other architectures. At similar accuracy, performance of VINE, used in GRAPE-tree mode, is approximately a factor 2 slower than that of VINE, used in host-only mode. Further optimizations of the GRAPE/host communications could improve the speed by as much as a factor of 3, but have not yet been implemented in VINE
Global and kinetic MHD simulation by the Gpic-MHD code
International Nuclear Information System (INIS)
Naitou, Hiroshi; Yamada, Yusuke; Kajiwara, Kenji; Lee, Wei-li; Tokuda, Shinji; Yagi, Masatoshi
2011-01-01
In order to implement large-scale and high-beta tokamak simulation, a new algorithm of the electromagnetic gyrokinetic PIC (particle-in-cell) code was proposed and installed on the Gpic-MHD code [Gyrokinetic PIC code for magnetohydrodynamic (MHD) simulation]. In the new algorithm, the vortex equation and the generalized ohm's law along the magnetic field are derived from the basic equations of the gyrokinetic Vlasov, Poisson, and Ampere system and are used to describe the spatio-temporal evolution of the field quantities of the electrostatic potential φ and the longitudinal component of the vector potential A z . Particle information is mainly used to estimate second order moments in the generalized ohm's law. Because the lower order moments of the charge density and the longitudinal current density are not used explicitly to determine φ and A z , the numerical noise induced by the discreteness of particle quantities reduces drastically. Another advantage of the algorithm is that the longitudinal induced electric field, E Tz =-∂A z /∂t, is explicitly estimated by the generalized ohm's law and used in the equations of motion. The particle velocities along the magnetic field are used (v z -formulation) instead of generalized momentums (p z -formulation), hence there is no problem of 'cancellation', which appear when estimating A z from the Ampere's law in the p z -formulation. The successful simulation of the collisionless internal kink mode by new Gpic-MHD with the realistic values of the large-scale and high-beta, revealed the usefulness of the new algorithm. (author)
Leveraging Quick Response Code Technology to Facilitate Simulation-Based Leaderboard Competition.
Chang, Todd P; Doughty, Cara B; Mitchell, Diana; Rutledge, Chrystal; Auerbach, Marc A; Frisell, Karin; Jani, Priti; Kessler, David O; Wolfe, Heather; MacKinnon, Ralph J; Dewan, Maya; Pirie, Jonathan; Lemke, Daniel; Khattab, Mona; Tofil, Nancy; Nagamuthu, Chenthila; Walsh, Catharine M
2018-02-01
Leaderboards provide feedback on relative performance and a competitive atmosphere for both self-guided improvement and social comparison. Because simulation can provide substantial quantitative participant feedback, leaderboards can be used, not only locally but also in a multidepartment, multicenter fashion. Quick Response (QR) codes can be integrated to allow participants to access and upload data. We present the development, implementation, and initial evaluation of an online leaderboard employing principles of gamification using points, badges, and leaderboards designed to enhance competition among healthcare providers. This article details the fundamentals behind the development and implementation of a user-friendly, online, multinational leaderboard that employs principles of gamification to enhance competition and integrates a QR code system to promote both self-reporting of performance data and data integrity. An open-ended survey was administered to capture perceptions of leaderboard implementation. Conceptual step-by-step instructions detailing how to apply the QR code system to any leaderboard using simulated or real performance metrics are outlined using an illustrative example of a leaderboard that employed simulated cardiopulmonary resuscitation performance scores to compare participants across 17 hospitals in 4 countries for 16 months. The following three major descriptive categories that captured perceptions of leaderboard implementation emerged from initial evaluation data from 10 sites: (1) competition, (2) longevity, and (3) perceived deficits. A well-designed leaderboard should be user-friendly and encompass best practices in gamification principles while collecting and storing data for research analyses. Easy storage and export of data allow for longitudinal record keeping that can be leveraged both to track compliance and to enable social competition.
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
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.
Performance Analysis of an Astrophysical Simulation Code on the Intel Xeon Phi Architecture
Noormofidi, Vahid; Atlas, Susan R.; Duan, Huaiyu
2015-01-01
We have developed the astrophysical simulation code XFLAT to study neutrino oscillations in supernovae. XFLAT is designed to utilize multiple levels of parallelism through MPI, OpenMP, and SIMD instructions (vectorization). It can run on both CPU and Xeon Phi co-processors based on the Intel Many Integrated Core Architecture (MIC). We analyze the performance of XFLAT on configurations with CPU only, Xeon Phi only and both CPU and Xeon Phi. We also investigate the impact of I/O and the multi-n...
Simulations of the broad line region of NGC 5548 with CLOUDY code: Temperature determination
Directory of Open Access Journals (Sweden)
Ilić D.
2007-01-01
Full Text Available In this paper an analysis of the physical properties of the Broad Line Region (BLR of the active galaxy NGC 5548 is presented. Using the photoionization code CLOUDY and the measurements of Peterson et al. (2002, the physical conditions of the BLR are simulated and the BLR temperature is obtained. This temperature was compared to the temperature estimated with the Boltzmann-Plot (BP method (Popović et al. 2007. It was shown that the measured variability in the BLR temperature could be due to the change in the hydrogen density.
Design and construction of a graphical interface for automatic generation of simulation code GEANT4
International Nuclear Information System (INIS)
Driss, Mozher; Bouzaine Ismail
2007-01-01
This work is set in the context of the engineering studies final project; it is accomplished in the center of nuclear sciences and technologies in Sidi Thabet. This project is about conceiving and developing a system based on graphical user interface which allows an automatic codes generation for simulation under the GEANT4 engine. This system aims to facilitate the use of GEANT4 by scientific not necessary expert in this engine and to be used in different areas: research, industry and education. The implementation of this project uses Root library and several programming languages such as XML and XSL. (Author). 5 refs
Large Eddy Simulation of turbulent flows in compound channels with a finite element code
International Nuclear Information System (INIS)
Xavier, C.M.; Petry, A.P.; Moeller, S.V.
2011-01-01
This paper presents the numerical investigation of the developing flow in a compound channel formed by a rectangular main channel and a gap in one of the sidewalls. A three dimensional Large Eddy Simulation computational code with the classic Smagorinsky model is introduced, where the transient flow is modeled through the conservation equations of mass and momentum of a quasi-incompressible, isothermal continuous medium. Finite Element Method, Taylor-Galerkin scheme and linear hexahedrical elements are applied. Numerical results of velocity profile show the development of a shear layer in agreement with experimental results obtained with Pitot tube and hot wires. (author)
International Nuclear Information System (INIS)
Rojas C, E.L.; Varon T, C.F.; Pedraza N, R.
2007-01-01
The treatment of the breast cancer at early stages is of vital importance. For that, most of the investigations are dedicated to the early detection of the suffering and their treatment. As investigation consequence and clinical practice, in 2002 it was developed in U.S.A. an irradiation system of high dose rate known as Mammosite. In this work we carry out dose calculations for a simplified Mammosite system with the Monte Carlo Penelope simulation code and MCNPX, varying the concentration of the contrast material that it is used in the one. (Author)
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.)
A modular simulation code applied to pressurized water nuclear power plants
International Nuclear Information System (INIS)
Agnoux, D.
1992-01-01
Analysis of the overall operation of an installation requires taking into account all couplings between the various components and integrating all the automatic actions initiated by control and instrumentation. The tool used for this analysis must be a high performing simulation model, flexible enough to be able to be quickly adapted to varying configurations. In order to study the behaviour of PWR nuclear power stations during normal or incidental operating transients, EDF-SEPTEN has developed the ERABLE code (Etudes Reacteurs a Base LEGO), based on the LEGO software package. (author)
Parallel treatment of simulation particles in particle-in-cell codes on SUPRENUM
International Nuclear Information System (INIS)
Seldner, D.
1990-02-01
This report contains the program documentation and description of the program package 2D-PLAS, which has been developed at the Nuclear Research Center Karlsruhe in the Institute for Data Processing in Technology (IDT) under the auspices of the BMFT. 2D-PLAS is a parallel program version of the treatment of the simulation particles of the two-dimensional stationary particle-in-cell code BFCPIC which has been developed at the Nuclear Research Center Karlsruhe. This parallel version has been designed for the parallel computer SUPRENUM. (orig.) [de
MOCARS: a Monte Carlo code for determining the distribution and simulation limits
International Nuclear Information System (INIS)
Matthews, S.D.
1977-07-01
MOCARS is a computer program designed for the INEL CDC 76-173 operating system to determine the distribution and simulation limits for a function by Monte Carlo techniques. The code randomly samples data from any of the 12 user-specified distributions and then either evaluates the cut set system unavailability or a user-specified function with the sample data. After the data are ordered, the values at various quantities and associated confidence bounds are calculated for output. Also available for output on microfilm are the frequency and cumulative distribution histograms from the sample data. 29 figures, 4 tables