Gyrokinetic Vlasov code including full three-dimensional geometry of experiments

International Nuclear Information System (INIS)

Nunami, Masanori; Watanabe, Tomohiko; Sugama, Hideo

2010-03-01

A new gyrokinetic Vlasov simulation code, GKV-X, is developed for investigating the turbulent transport in magnetic confinement devices with non-axisymmetric configurations. Effects of the magnetic surface shapes in a three-dimensional equilibrium obtained from the VMEC code are accurately incorporated. Linear simulations of the ion temperature gradient instabilities and the zonal flows in the Large Helical Device (LHD) configuration are carried out by the GKV-X code for a benchmark test against the GKV code. The frequency, the growth rate, and the mode structure of the ion temperature gradient instability are influenced by the VMEC geometrical data such as the metric tensor components of the Boozer coordinates for high poloidal wave numbers, while the difference between the zonal flow responses obtained by the GKV and GKV-X codes is found to be small in the core LHD region. (author)

TESLA: Large Signal Simulation Code for Klystrons

International Nuclear Information System (INIS)

Vlasov, Alexander N.; Cooke, Simon J.; Chernin, David P.; Antonsen, Thomas M. Jr.; Nguyen, Khanh T.; Levush, Baruch

2003-01-01

TESLA (Telegraphist's Equations Solution for Linear Beam Amplifiers) is a new code designed to simulate linear beam vacuum electronic devices with cavities, such as klystrons, extended interaction klystrons, twistrons, and coupled cavity amplifiers. The model includes a self-consistent, nonlinear solution of the three-dimensional electron equations of motion and the solution of time-dependent field equations. The model differs from the conventional Particle in Cell approach in that the field spectrum is assumed to consist of a carrier frequency and its harmonics with slowly varying envelopes. Also, fields in the external cavities are modeled with circuit like equations and couple to fields in the beam region through boundary conditions on the beam tunnel wall. The model in TESLA is an extension of the model used in gyrotron code MAGY. The TESLA formulation has been extended to be capable to treat the multiple beam case, in which each beam is transported inside its own tunnel. The beams interact with each other as they pass through the gaps in their common cavities. The interaction is treated by modification of the boundary conditions on the wall of each tunnel to include the effect of adjacent beams as well as the fields excited in each cavity. The extended version of TESLA for the multiple beam case, TESLA-MB, has been developed for single processor machines, and can run on UNIX machines and on PC computers with a large memory (above 2GB). The TESLA-MB algorithm is currently being modified to simulate multiple beam klystrons on multiprocessor machines using the MPI (Message Passing Interface) environment. The code TESLA has been verified by comparison with MAGIC for single and multiple beam cases. The TESLA code and the MAGIC code predict the same power within 1% for a simple two cavity klystron design while the computational time for TESLA is orders of magnitude less than for MAGIC 2D. In addition, recently TESLA was used to model the L-6048 klystron, code

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

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)

XFEL OSCILLATOR SIMULATION INCLUDING ANGLE-DEPENDENT CRYSTAL REFLECTIVITY

International Nuclear Information System (INIS)

Fawley, William; Lindberg, Ryan; Kim, K.-J.; Shvyd'ko, Yuri

2010-01-01

The oscillator package within the GINGER FEL simulation code has now been extended to include angle-dependent reflectivity properties of Bragg crystals. Previously, the package was modified to include frequencydependent reflectivity in order to model x-ray FEL oscillators from start-up from shot noise through to saturation. We present a summary of the algorithms used for modeling the crystal reflectivity and radiation propagation outside the undulator, discussing various numerical issues relevant to the domain of high Fresnel number and efficient Hankel transforms. We give some sample XFEL-O simulation results obtained with the angle-dependent reflectivity model, with particular attention directed to the longitudinal and transverse coherence of the radiation output.

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...

Numerical simulation on the screw pinch by 2-D MHD pinch code 'topics' including impurities and neutrals effects

International Nuclear Information System (INIS)

Nagata, A.; Ashida, H.; Okamoto, M.; Hirano, K.

1981-03-01

Two dimentional fluid simulation code ''TOPICS'' is developed for the STP-2, the shock heated screw pinch at Nagoya. It involves the effects of impurity ions and neutral atoms. In order to estimate the radiation losses, the impurity continuity equations with ionizations and recombinations are solved simultaneously with the plasma fluid equations. The results are compared with the coronal equilibrium model. It is found that the coronal equilibrium model underestimates the radiation losses from shock heated pinch plasmas in its initial dynamic phase. The present calculations including impurities and neutrals show the importance of the radiation losses from the plasma of the STP-2. Introducing the anomalous resistivity caused by the ion acoustic instability, the observed magnetic field penetration is explained fairly well. (author)

Nexus: A modular workflow management system for quantum simulation codes

Science.gov (United States)

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.

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

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)

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.

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.

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.

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

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

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)

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

Science.gov (United States)

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.

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)

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

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

Low-temperature plasma simulations with the LSP PIC code

Science.gov (United States)

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.

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)

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

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

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

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

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)

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.)

Simulations of Laboratory Astrophysics Experiments using the CRASH code

Science.gov (United States)

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.

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

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

Including collisions in gyrokinetic tokamak and stellarator simulations

International Nuclear Information System (INIS)

Kauffmann, Karla

2012-01-01

Particle and heat transport in fusion devices often exceed the neoclassical prediction. This anomalous transport is thought to be produced by turbulence caused by microinstabilities such as ion and electron-temperature-gradient (ITG/ETG) and trapped-electron-mode (TEM) instabilities, the latter ones known for being strongly influenced by collisions. Additionally, in stellarators, the neoclassical transport can be important in the core, and therefore investigation of the effects of collisions is an important field of study. Prior to this thesis, however, no gyrokinetic simulations retaining collisions had been performed in stellarator geometry. In this work, collisional effects were added to EUTERPE, a previously collisionless gyrokinetic code which utilizes the δf method. To simulate the collisions, a pitch-angle scattering operator was employed, and its implementation was carried out following the methods proposed in [Takizuka and Abe 1977, Vernay Master's thesis 2008]. To test this implementation, the evolution of the distribution function in a homogeneous plasma was first simulated, where Legendre polynomials constitute eigenfunctions of the collision operator. Also, the solution of the Spitzer problem was reproduced for a cylinder and a tokamak. Both these tests showed that collisions were correctly implemented and that the code is suited for more complex simulations. As a next step, the code was used to calculate the neoclassical radial particle flux by neglecting any turbulent fluctuations in the distribution function and the electric field. Particle fluxes in the neoclassical analytical regimes were simulated for tokamak and stellarator (LHD) configurations. In addition to the comparison with analytical fluxes, a successful benchmark with the DKES code was presented for the tokamak case, which further validates the code for neoclassical simulations. In the final part of the work, the effects of collisions were investigated for slab and toroidal ITGs and

Including collisions in gyrokinetic tokamak and stellarator simulations

Energy Technology Data Exchange (ETDEWEB)

Kauffmann, Karla

2012-04-10

Particle and heat transport in fusion devices often exceed the neoclassical prediction. This anomalous transport is thought to be produced by turbulence caused by microinstabilities such as ion and electron-temperature-gradient (ITG/ETG) and trapped-electron-mode (TEM) instabilities, the latter ones known for being strongly influenced by collisions. Additionally, in stellarators, the neoclassical transport can be important in the core, and therefore investigation of the effects of collisions is an important field of study. Prior to this thesis, however, no gyrokinetic simulations retaining collisions had been performed in stellarator geometry. In this work, collisional effects were added to EUTERPE, a previously collisionless gyrokinetic code which utilizes the {delta}f method. To simulate the collisions, a pitch-angle scattering operator was employed, and its implementation was carried out following the methods proposed in [Takizuka and Abe 1977, Vernay Master's thesis 2008]. To test this implementation, the evolution of the distribution function in a homogeneous plasma was first simulated, where Legendre polynomials constitute eigenfunctions of the collision operator. Also, the solution of the Spitzer problem was reproduced for a cylinder and a tokamak. Both these tests showed that collisions were correctly implemented and that the code is suited for more complex simulations. As a next step, the code was used to calculate the neoclassical radial particle flux by neglecting any turbulent fluctuations in the distribution function and the electric field. Particle fluxes in the neoclassical analytical regimes were simulated for tokamak and stellarator (LHD) configurations. In addition to the comparison with analytical fluxes, a successful benchmark with the DKES code was presented for the tokamak case, which further validates the code for neoclassical simulations. In the final part of the work, the effects of collisions were investigated for slab and toroidal

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.

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)

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)

MicroHH 1.0: a computational fluid dynamics code for direct numerical simulation and large-eddy simulation of atmospheric boundary layer flows

Science.gov (United States)

van Heerwaarden, Chiel C.; van Stratum, Bart J. H.; Heus, Thijs; Gibbs, Jeremy A.; Fedorovich, Evgeni; Mellado, Juan Pedro

2017-08-01

This paper describes MicroHH 1.0, a new and open-source (www.microhh.org) computational fluid dynamics code for the simulation of turbulent flows in the atmosphere. It is primarily made for direct numerical simulation but also supports large-eddy simulation (LES). The paper covers the description of the governing equations, their numerical implementation, and the parameterizations included in the code. Furthermore, the paper presents the validation of the dynamical core in the form of convergence and conservation tests, and comparison of simulations of channel flows and slope flows against well-established test cases. The full numerical model, including the associated parameterizations for LES, has been tested for a set of cases under stable and unstable conditions, under the Boussinesq and anelastic approximations, and with dry and moist convection under stationary and time-varying boundary conditions. The paper presents performance tests showing good scaling from 256 to 32 768 processes. The graphical processing unit (GPU)-enabled version of the code can reach a speedup of more than an order of magnitude for simulations that fit in the memory of a single GPU.

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.

SPACE CHARGE SIMULATION METHODS INCORPORATED IN SOME MULTI - PARTICLE TRACKING CODES AND THEIR RESULTS COMPARISON

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

SPACE CHARGE SIMULATION METHODS INCORPORATED IN SOME MULTI - PARTICLE TRACKING CODES AND THEIR RESULTS COMPARISON.

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.

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.

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.

ANNarchy: a code generation approach to neural simulations on parallel hardware

Science.gov (United States)

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

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.

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.

NRC model simulations in support of the hydrologic code intercomparison study (HYDROCOIN): Level 1-code verification

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

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)

CHOLLA: A NEW MASSIVELY PARALLEL HYDRODYNAMICS CODE FOR ASTROPHYSICAL SIMULATION

International Nuclear Information System (INIS)

Schneider, Evan E.; Robertson, Brant E.

2015-01-01

We present Computational Hydrodynamics On ParaLLel Architectures (Cholla ), a new three-dimensional hydrodynamics code that harnesses the power of graphics processing units (GPUs) to accelerate astrophysical simulations. Cholla models the Euler equations on a static mesh using state-of-the-art techniques, including the unsplit Corner Transport Upwind algorithm, a variety of exact and approximate Riemann solvers, and multiple spatial reconstruction techniques including the piecewise parabolic method (PPM). Using GPUs, Cholla evolves the fluid properties of thousands of cells simultaneously and can update over 10 million cells per GPU-second while using an exact Riemann solver and PPM reconstruction. Owing to the massively parallel architecture of GPUs and the design of the Cholla code, astrophysical simulations with physically interesting grid resolutions (≳256 3 ) can easily be computed on a single device. We use the Message Passing Interface library to extend calculations onto multiple devices and demonstrate nearly ideal scaling beyond 64 GPUs. A suite of test problems highlights the physical accuracy of our modeling and provides a useful comparison to other codes. We then use Cholla to simulate the interaction of a shock wave with a gas cloud in the interstellar medium, showing that the evolution of the cloud is highly dependent on its density structure. We reconcile the computed mixing time of a turbulent cloud with a realistic density distribution destroyed by a strong shock with the existing analytic theory for spherical cloud destruction by describing the system in terms of its median gas density

CHOLLA: A NEW MASSIVELY PARALLEL HYDRODYNAMICS CODE FOR ASTROPHYSICAL SIMULATION

Energy Technology Data Exchange (ETDEWEB)

Schneider, Evan E.; Robertson, Brant E. [Steward Observatory, University of Arizona, 933 North Cherry Avenue, Tucson, AZ 85721 (United States)

2015-04-15

We present Computational Hydrodynamics On ParaLLel Architectures (Cholla ), a new three-dimensional hydrodynamics code that harnesses the power of graphics processing units (GPUs) to accelerate astrophysical simulations. Cholla models the Euler equations on a static mesh using state-of-the-art techniques, including the unsplit Corner Transport Upwind algorithm, a variety of exact and approximate Riemann solvers, and multiple spatial reconstruction techniques including the piecewise parabolic method (PPM). Using GPUs, Cholla evolves the fluid properties of thousands of cells simultaneously and can update over 10 million cells per GPU-second while using an exact Riemann solver and PPM reconstruction. Owing to the massively parallel architecture of GPUs and the design of the Cholla code, astrophysical simulations with physically interesting grid resolutions (≳256{sup 3}) can easily be computed on a single device. We use the Message Passing Interface library to extend calculations onto multiple devices and demonstrate nearly ideal scaling beyond 64 GPUs. A suite of test problems highlights the physical accuracy of our modeling and provides a useful comparison to other codes. We then use Cholla to simulate the interaction of a shock wave with a gas cloud in the interstellar medium, showing that the evolution of the cloud is highly dependent on its density structure. We reconcile the computed mixing time of a turbulent cloud with a realistic density distribution destroyed by a strong shock with the existing analytic theory for spherical cloud destruction by describing the system in terms of its median gas density.

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

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.)

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.)

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

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.

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

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)

Computer Code for Nanostructure Simulation

Science.gov (United States)

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.

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)

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)

Parallelization of a Monte Carlo particle transport simulation code

Science.gov (United States)

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.

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

Development of code PRETOR for stellarator simulation

International Nuclear Information System (INIS)

Dies, J.; Fontanet, J.; Fontdecaba, J.M.; Castejon, F.; Alejandre, C.

1998-01-01

The Department de Fisica i Enginyeria Nuclear (DFEN) of the UPC has some experience in the development of the transport code PRETOR. This code has been validated with shots of DIII-D, JET and TFTR, it has also been used in the simulation of operational scenarios of ITER fast burnt termination. Recently, the association EURATOM-CIEMAT has started the operation of the TJ-II stellarator. Due to the need of validating the results given by others transport codes applied to stellarators and because all of them made some approximations, as a averaging magnitudes in each magnetic surface, it was thought suitable to adapt the PRETOR code to devices without axial symmetry, like stellarators, which is very suitable for the specific needs of the study of TJ-II. Several modifications are required in PRETOR; the main concerns to the models of: magnetic equilibrium, geometry and transport of energy and particles. In order to solve the complex magnetic equilibrium geometry the powerful numerical code VMEC has been used. This code gives the magnetic surface shape as a Fourier series in terms of the harmonics (m,n). Most of the geometric magnitudes are also obtained from the VMEC results file. The energy and particle transport models will be replaced by other phenomenological models that are better adapted to stellarator simulation. Using the proposed models, it is pretended to reproduce experimental data available from present stellarators, given especial attention to the TJ-II of the association EURATOM-CIEMAT. (Author)

Parallel and vector implementation of APROS simulator code

International Nuclear Information System (INIS)

Niemi, J.; Tommiska, J.

1990-01-01

In this paper the vector and parallel processing implementation of a general purpose simulator code is discussed. In this code the utilization of vector processing is straightforward. In addition to the loop level parallel processing, the functional decomposition and the domain decomposition have been considered. Results represented for a PWR-plant simulation illustrate the potential speed-up factors of the alternatives. It turns out that the loop level parallelism and the domain decomposition are the most promising alternative to employ the parallel processing. (author)

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)

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

Simulating Coupling Complexity in Space Plasmas: First Results from a new code

Science.gov (United States)

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

Partially coherent X-ray wavefront propagation simulations including grazing-incidence focusing optics.

Science.gov (United States)

Canestrari, Niccolo; Chubar, Oleg; Reininger, Ruben

2014-09-01

X-ray beamlines in modern synchrotron radiation sources make extensive use of grazing-incidence reflective optics, in particular Kirkpatrick-Baez elliptical mirror systems. These systems can focus the incoming X-rays down to nanometer-scale spot sizes while maintaining relatively large acceptance apertures and high flux in the focused radiation spots. In low-emittance storage rings and in free-electron lasers such systems are used with partially or even nearly fully coherent X-ray beams and often target diffraction-limited resolution. Therefore, their accurate simulation and modeling has to be performed within the framework of wave optics. Here the implementation and benchmarking of a wave-optics method for the simulation of grazing-incidence mirrors based on the local stationary-phase approximation or, in other words, the local propagation of the radiation electric field along geometrical rays, is described. The proposed method is CPU-efficient and fully compatible with the numerical methods of Fourier optics. It has been implemented in the Synchrotron Radiation Workshop (SRW) computer code and extensively tested against the geometrical ray-tracing code SHADOW. The test simulations have been performed for cases without and with diffraction at mirror apertures, including cases where the grazing-incidence mirrors can be hardly approximated by ideal lenses. Good agreement between the SRW and SHADOW simulation results is observed in the cases without diffraction. The differences between the simulation results obtained by the two codes in diffraction-dominated cases for illumination with fully or partially coherent radiation are analyzed and interpreted. The application of the new method for the simulation of wavefront propagation through a high-resolution X-ray microspectroscopy beamline at the National Synchrotron Light Source II (Brookhaven National Laboratory, USA) is demonstrated.

OpenQ∗D simulation code for QCD+QED

DEFF Research Database (Denmark)

Campos, Isabel; Fritzsch, Patrick; Hansen, Martin

2018-01-01

The openQ∗D code for the simulation of QCD+QED with C∗ boundary conditions is presented. This code is based on openQCD-1.6, from which it inherits the core features that ensure its efficiency: the locally-deflated SAP-preconditioned GCR solver, the twisted-mass frequency splitting of the fermion....... An alpha version of this code is publicly available and can be downloaded from http://rcstar.web.cern.ch/....

A methodology for the rigorous verification of plasma simulation codes

Science.gov (United States)

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.

DNA strand breaks induced by electrons simulated with nanodosimetry Monte Carlo simulation code: NASIC

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)

Comparisons of 'Identical' Simulations by the Eulerian Gyrokinetic Codes GS2 and GYRO

Science.gov (United States)

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).

Developing HYDMN code to include the transient of MNSR

International Nuclear Information System (INIS)

Al-Barhoum, M.

2000-11-01

A description of the programs added to HYDMN code (a code for thermal-hydraulic steady state of MNSR) to include the transient of the same MNSR is presented. The code asks the initial conditions for the power (in k W) and the cold initial core inlet temperature (in degrees centigrade). A time-dependent study of the coolant inlet and outlet temperature, its speed, pool and tank temperatures is done for MNSR in general and for the Syrian MNSR in particular. The study solves the differential equations taken from reference (1) by using some numerical methods found in reference (3). The code becomes this way independent of any external information source. (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

A multiscale numerical algorithm for heat transfer simulation between multidimensional CFD and monodimensional system codes

Science.gov (United States)

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

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)

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.

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.

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

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.

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

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)

Phase 1 Validation Testing and Simulation for the WEC-Sim Open Source Code

Science.gov (United States)

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

Simulation and interpretation codes for the JET ECE diagnostic. Part 1: physics of the codes' operation

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

Development of a multi-grid FDTD code for three-dimensional simulation of large microwave sintering experiments

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.

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.

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)

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)

MOCCA Code for Star Cluster Simulation: Comparison with Optical Observations using COCOA

OpenAIRE

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...

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)

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

Simulations of linear and Hamming codes using SageMath

Science.gov (United States)

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.

Simulations of Tokamak Edge Turbulence Including Self-Consistent Zonal Flows

Science.gov (United States)

Cohen, Bruce; Umansky, Maxim

2013-10-01

Progress on simulations of electromagnetic drift-resistive ballooning turbulence in the tokamak edge is summarized in this mini-conference talk. A more detailed report on this work is presented in a poster at this conference. This work extends our previous work to include self-consistent zonal flows and their effects. The previous work addressed the simulation of L-mode tokamak edge turbulence using the turbulence code BOUT. The calculations used realistic single-null geometry and plasma parameters of the DIII-D tokamak and produced fluctuation amplitudes, fluctuation spectra, and particle and thermal fluxes that compare favorably to experimental data. In the effect of sheared ExB poloidal rotation is included with an imposed static radial electric field fitted to experimental data. In the new work here we include the radial electric field self-consistently driven by the microturbulence, which contributes to the sheared ExB poloidal rotation (zonal flow generation). We present simulations with/without zonal flows for both cylindrical geometry, as in the UCLA Large Plasma Device, and for the DIII-D tokamak L-mode cases in to quantify the influence of self-consistent zonal flows on the microturbulence and the concomitant transport. This work was performed under the auspices of the US Department of Energy under contract DE-AC52-07NA27344 at the Lawrence Livermore National Laboratory.

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

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)

Simulation and verification studies of reactivity initiated accident by comparative approach of NK/TH coupling codes and RELAP5 code

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.

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)

Tristan code and its application

Science.gov (United States)

Nishikawa, K.-I.

Since TRISTAN: The 3-D Electromagnetic Particle Code was introduced in 1990, it has been used for many applications including the simulations of global solar windmagnetosphere interaction. The most essential ingridients of this code have been published in the ISSS-4 book. In this abstract we describe some of issues and an application of this code for the study of global solar wind-magnetosphere interaction including a substorm study. The basic code (tristan.f) for the global simulation and a local simulation of reconnection with a Harris model (issrec2.f) are available at http:/www.physics.rutger.edu/˜kenichi. For beginners the code (isssrc2.f) with simpler boundary conditions is suitable to start to run simulations. The future of global particle simulations for a global geospace general circulation (GGCM) model with predictive capability (for Space Weather Program) is discussed.

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

MOCCA code for star cluster simulation: comparison with optical observations using COCOA

Science.gov (United States)

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.

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.

Generation of initial geometries for the simulation of the physical system in the DualPHYsics code

International Nuclear Information System (INIS)

Segura Q, E.

2013-01-01

In the diverse research areas of the Instituto Nacional de Investigaciones Nucleares (ININ) are different activities related to science and technology, one of great interest is the study and treatment of the collection and storage of radioactive waste. Therefore at ININ the draft on the simulation of the pollutants diffusion in the soil through a porous medium (third stage) has this problem inherent aspects, hence a need for such a situation is to generate the initial geometry of the physical system For the realization of the simulation method is implemented smoothed particle hydrodynamics (SPH). This method runs in DualSPHysics code, which has great versatility and ability to simulate phenomena of any physical system where hydrodynamic aspects combine. In order to simulate a physical system DualSPHysics code, you need to preset the initial geometry of the system of interest, then this is included in the input file of the code. The simulation sets the initial geometry through regular geometric bodies positioned at different points in space. This was done through a programming language (Fortran, C + +, Java, etc..). This methodology will provide the basis to simulate more complex geometries future positions and form. (Author)

TASS code topical report. V.1 TASS code technical manual

International Nuclear Information System (INIS)

Sim, Suk K.; Chang, W. P.; Kim, K. D.; Kim, H. C.; Yoon, H. Y.

1997-02-01

TASS 1.0 code has been developed at KAERI for the initial and reload non-LOCA safety analysis for the operating PWRs as well as the PWRs under construction in Korea. TASS code will replace various vendor's non-LOCA safety analysis codes currently used for the Westinghouse and ABB-CE type PWRs in Korea. This can be achieved through TASS code input modifications specific to each reactor type. The TASS code can be run interactively through the keyboard operation. A simimodular configuration used in developing the TASS code enables the user easily implement new models. TASS code has been programmed using FORTRAN77 which makes it easy to install and port for different computer environments. The TASS code can be utilized for the steady state simulation as well as the non-LOCA transient simulations such as power excursions, reactor coolant pump trips, load rejections, loss of feedwater, steam line breaks, steam generator tube ruptures, rod withdrawal and drop, and anticipated transients without scram (ATWS). The malfunctions of the control systems, components, operator actions and the transients caused by the malfunctions can be easily simulated using the TASS code. This technical report describes the TASS 1.0 code models including reactor thermal hydraulic, reactor core and control models. This TASS code models including reactor thermal hydraulic, reactor core and control models. This TASS code technical manual has been prepared as a part of the TASS code manual which includes TASS code user's manual and TASS code validation report, and will be submitted to the regulatory body as a TASS code topical report for a licensing non-LOCA safety analysis for the Westinghouse and ABB-CE type PWRs operating and under construction in Korea. (author). 42 refs., 29 tabs., 32 figs

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

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

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.

Comparing TCV experimental VDE responses with DINA code simulations

Science.gov (United States)

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)

Calibration of Monte Carlo simulation code to low voltage electron beams through radiachromic dosimetry

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

Development of three-dimensional neoclassical transport simulation code with high performance Fortran on a vector-parallel computer

International Nuclear Information System (INIS)

Satake, Shinsuke; Okamoto, Masao; Nakajima, Noriyoshi; Takamaru, Hisanori

2005-11-01

A neoclassical transport simulation code (FORTEC-3D) applicable to three-dimensional configurations has been developed using High Performance Fortran (HPF). Adoption of computing techniques for parallelization and a hybrid simulation model to the δf Monte-Carlo method transport simulation, including non-local transport effects in three-dimensional configurations, makes it possible to simulate the dynamism of global, non-local transport phenomena with a self-consistent radial electric field within a reasonable computation time. In this paper, development of the transport code using HPF is reported. Optimization techniques in order to achieve both high vectorization and parallelization efficiency, adoption of a parallel random number generator, and also benchmark results, are shown. (author)

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)

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

Finite element methods in a simulation code for offshore wind turbines

Science.gov (United States)

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.).

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

COOL: A code for Dynamic Monte Carlo Simulation of molecular dynamics

Science.gov (United States)

Barletta, Paolo

2012-02-01

Cool is a program to simulate evaporative and sympathetic cooling for a mixture of two gases co-trapped in an harmonic potential. The collisions involved are assumed to be exclusively elastic, and losses are due to evaporation from the trap. Each particle is followed individually in its trajectory, consequently properties such as spatial densities or energy distributions can be readily evaluated. The code can be used sequentially, by employing one output as input for another run. The code can be easily generalised to describe more complicated processes, such as the inclusion of inelastic collisions, or the possible presence of more than two species in the trap. New version program summaryProgram title: COOL Catalogue identifier: AEHJ_v2_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEHJ_v2_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 1 097 733 No. of bytes in distributed program, including test data, etc.: 18 425 722 Distribution format: tar.gz Programming language: C++ Computer: Desktop Operating system: Linux RAM: 500 Mbytes Classification: 16.7, 23 Catalogue identifier of previous version: AEHJ_v1_0 Journal reference of previous version: Comput. Phys. Comm. 182 (2011) 388 Does the new version supersede the previous version?: Yes Nature of problem: Simulation of the sympathetic process occurring for two molecular gases co-trapped in a deep optical trap. Solution method: The Direct Simulation Monte Carlo method exploits the decoupling, over a short time period, of the inter-particle interaction from the trapping potential. The particle dynamics is thus exclusively driven by the external optical field. The rare inter-particle collisions are considered with an acceptance/rejection mechanism, that is, by comparing a random number to the collisional probability

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.

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)

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.

Benchmarking and scaling studies of pseudospectral code Tarang for turbulence simulations

KAUST Repository

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

KAUST Repository

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.

EGS code system: computer programs for the Monte Carlo simulation of electromagnetic cascade showers. Version 3

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

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.

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

Srna-Monte Carlo codes for proton transport simulation in combined and voxelized geometries

CERN Document Server

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...

Python Radiative Transfer Emission code (PyRaTE): non-LTE spectral lines simulations

Science.gov (United States)

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.

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

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)

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

Simulation of hydrogen deflagration experiment – Benchmark exercise with lumped-parameter codes

Energy Technology Data Exchange (ETDEWEB)

Kljenak, Ivo, E-mail: ivo.kljenak@ijs.si [Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Kuznetsov, Mikhail, E-mail: mike.kuznetsov@kit.edu [Karlsruhe Institute of Technology, Kaiserstraße 12, 76131 Karlsruhe (Germany); Kostka, Pal, E-mail: kostka@nubiki.hu [NUBIKI Nuclear Safety Research Institute, Konkoly-Thege Miklós út 29-33, 1121 Budapest (Hungary); Kubišova, Lubica, E-mail: lubica.kubisova@ujd.gov.sk [Nuclear Regulatory Authority of the Slovak Republic, Bajkalská 27, 82007 Bratislava (Slovakia); Maltsev, Mikhail, E-mail: maltsev_MB@aep.ru [JSC Atomenergoproekt, 1, st. Podolskykh Kursantov, Moscow (Russian Federation); Manzini, Giovanni, E-mail: giovanni.manzini@rse-web.it [Ricerca sul Sistema Energetico, Via Rubattino 54, 20134 Milano (Italy); Povilaitis, Mantas, E-mail: mantas.p@mail.lei.lt [Lithuania Energy Institute, Breslaujos g.3, 44403 Kaunas (Lithuania)

2015-03-15

Highlights: • Blind and open simulations of hydrogen combustion experiment in large-scale containment-like facility with different lumped-parameter codes. • Simulation of axial as well as radial flame propagation. • Confirmation of adequacy of lumped-parameter codes for safety analyses of actual nuclear power plants. - Abstract: An experiment on hydrogen deflagration (Upward Flame Propagation Experiment – UFPE) was proposed by the Jozef Stefan Institute (Slovenia) and performed in the HYKA A2 facility at the Karlsruhe Institute of Technology (Germany). The experimental results were used to organize a benchmark exercise for lumped-parameter codes. Six organizations (JSI, AEP, LEI, NUBIKI, RSE and UJD SR) participated in the benchmark exercise, using altogether four different computer codes: ANGAR, ASTEC, COCOSYS and ECART. Both blind and open simulations were performed. In general, all the codes provided satisfactory results of the pressure increase, whereas the results of the temperature show a wider dispersal. Concerning the flame axial and radial velocities, the results may be considered satisfactory, given the inherent simplification of the lumped-parameter description compared to the local instantaneous description.

Simulation of hydrogen deflagration experiment – Benchmark exercise with lumped-parameter codes

International Nuclear Information System (INIS)

Kljenak, Ivo; Kuznetsov, Mikhail; Kostka, Pal; Kubišova, Lubica; Maltsev, Mikhail; Manzini, Giovanni; Povilaitis, Mantas

2015-01-01

Highlights: • Blind and open simulations of hydrogen combustion experiment in large-scale containment-like facility with different lumped-parameter codes. • Simulation of axial as well as radial flame propagation. • Confirmation of adequacy of lumped-parameter codes for safety analyses of actual nuclear power plants. - Abstract: An experiment on hydrogen deflagration (Upward Flame Propagation Experiment – UFPE) was proposed by the Jozef Stefan Institute (Slovenia) and performed in the HYKA A2 facility at the Karlsruhe Institute of Technology (Germany). The experimental results were used to organize a benchmark exercise for lumped-parameter codes. Six organizations (JSI, AEP, LEI, NUBIKI, RSE and UJD SR) participated in the benchmark exercise, using altogether four different computer codes: ANGAR, ASTEC, COCOSYS and ECART. Both blind and open simulations were performed. In general, all the codes provided satisfactory results of the pressure increase, whereas the results of the temperature show a wider dispersal. Concerning the flame axial and radial velocities, the results may be considered satisfactory, given the inherent simplification of the lumped-parameter description compared to the local instantaneous description

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

Coding considerations for standalone molecular dynamics simulations of atomistic structures

Science.gov (United States)

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.

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)

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

The importance of including dynamic soil-structure interaction into wind turbine simulation codes

DEFF Research Database (Denmark)

Damgaard, Mads; Andersen, Lars Vabbersgaard; Ibsen, Lars Bo

2014-01-01

A rigorous numerical model, describing a wind turbine structure and subsoil, may contain thousands of degrees of freedom, making the approach computationally inefficient for fast time domain analysis. In order to meet the requirements of real-time calculations, the dynamic impedance of the founda......A rigorous numerical model, describing a wind turbine structure and subsoil, may contain thousands of degrees of freedom, making the approach computationally inefficient for fast time domain analysis. In order to meet the requirements of real-time calculations, the dynamic impedance...... of the foundation from a rigorous analysis can be formulated into a so-called lumped-parameter model consisting of a few springs, dashpots and point masses which are easily implemented into aeroelastic codes. In this paper, the quality of consistent lumped-parameter models of rigid surface footings and mono piles...... is examined. The optimal order of the models is determined and implemented into the aeroelastic code HAWC2, where the dynamic response of a 5.0 MW wind turbine is evaluated. In contrast to the fore-aft vibrations, the inclusion of soil-structure interaction is shown to be critical for the side-side vibrations...

SCAR - Post-Accident Simulator SIPA with safety analysis code CATHARE-2 and PWR cold shutdown state simulation

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

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.

Generating performance portable geoscientific simulation code with Firedrake (Invited)

Science.gov (United States)

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

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

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

CERN Multimedia

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

openQ*D simulation code for QCD+QED

Science.gov (United States)

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 single-phase rod bundle flow. Comparison between CFD-code ESTET, PWR core code THYC and experimental results

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)

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)

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)

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

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)

Monte Carlo simulation code modernization

CERN Multimedia

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...

OECD/NRC BWR Turbine Trip Benchmark: Simulation by POLCA-T Code

International Nuclear Information System (INIS)

Panayotov, Dobromir

2004-01-01

Westinghouse transient code POLCA-T brings together the system thermal-hydraulics plant models and three-dimensional (3-D) neutron kinetics core models. Participation in the OECD/NRC BWR Turbine Trip (TT) Benchmark is a part of our efforts toward the code's validation. The paper describes the objectives for TT analyses and gives a brief overview of the developed plant system input deck and 3-D core model.The results of exercise 1, system model without netronics, are presented. Sensitivity studies performed cover the maximal time step, turbine stop valve position and mass flow, feedwater temperature, and steam bypass mass flow. Results of exercise 2, 3-D core neutronic and thermal-hydraulic model with boundary conditions, are also presented. Sensitivity studies include the core inlet temperature, cladding properties, and direct heating to core coolant and bypass.The entire plant model was validated in the framework of the benchmark's phase 3. Sensitivity studies include the effect of SCRAM initialization and carry-under. The results obtained - transient fission power and its initial axial distribution and steam dome, core exit, lower and upper plenum, main steam line, and turbine inlet pressures - showed good agreement with measured data. Thus, the POLCA-T code capabilities for correct simulation of pressurizing transients with very fast power were proved

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)

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.

An object oriented code for simulating supersymmetric Yang-Mills theories

Science.gov (United States)

Catterall, Simon; Joseph, Anosh

2012-06-01

We present SUSY_LATTICE - a C++ program that can be used to simulate certain classes of supersymmetric Yang-Mills (SYM) theories, including the well known N=4 SYM in four dimensions, on a flat Euclidean space-time lattice. Discretization of SYM theories is an old problem in lattice field theory. It has resisted solution until recently when new ideas drawn from orbifold constructions and topological field theories have been brought to bear on the question. The result has been the creation of a new class of lattice gauge theories in which the lattice action is invariant under one or more supersymmetries. The resultant theories are local, free of doublers and also possess exact gauge-invariance. In principle they form the basis for a truly non-perturbative definition of the continuum SYM theories. In the continuum limit they reproduce versions of the SYM theories formulated in terms of twisted fields, which on a flat space-time is just a change of the field variables. In this paper, we briefly review these ideas and then go on to provide the details of the C++ code. We sketch the design of the code, with particular emphasis being placed on SYM theories with N=(2,2) in two dimensions and N=4 in three and four dimensions, making one-to-one comparisons between the essential components of the SYM theories and their corresponding counterparts appearing in the simulation code. The code may be used to compute several quantities associated with the SYM theories such as the Polyakov loop, mean energy, and the width of the scalar eigenvalue distributions. Program summaryProgram title: SUSY_LATTICE Catalogue identifier: AELS_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AELS_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC license, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 9315 No. of bytes in distributed program

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

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

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 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.

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)

The statistical significance of error probability as determined from decoding simulations for long codes

Science.gov (United States)

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.

DGR, GGR; molecular dynamical codes for simulating radiation damages in diamond and graphite crystals

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)

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.

A fully-implicit Particle-In-Cell Monte Carlo Collision code for the simulation of inductively coupled plasmas

Science.gov (United States)

Mattei, S.; Nishida, K.; Onai, M.; Lettry, J.; Tran, M. Q.; Hatayama, A.

2017-12-01

We present a fully-implicit electromagnetic Particle-In-Cell Monte Carlo collision code, called NINJA, written for the simulation of inductively coupled plasmas. NINJA employs a kinetic enslaved Jacobian-Free Newton Krylov method to solve self-consistently the interaction between the electromagnetic field generated by the radio-frequency coil and the plasma response. The simulated plasma includes a kinetic description of charged and neutral species as well as the collision processes between them. The algorithm allows simulations with cell sizes much larger than the Debye length and time steps in excess of the Courant-Friedrichs-Lewy condition whilst preserving the conservation of the total energy. The code is applied to the simulation of the plasma discharge of the Linac4 H- ion source at CERN. Simulation results of plasma density, temperature and EEDF are discussed and compared with optical emission spectroscopy measurements. A systematic study of the energy conservation as a function of the numerical parameters is presented.

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.

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.

TRACK The New Beam Dynamics Code

CERN Document Server

Mustapha, Brahim; Ostroumov, Peter; Schnirman-Lessner, Eliane

2005-01-01

The new ray-tracing code TRACK was developed* to fulfill the special requirements of the RIA accelerator systems. The RIA lattice includes an ECR ion source, a LEBT containing a MHB and a RFQ followed by three SC linac sections separated by two stripping stations with appropriate magnetic transport systems. No available beam dynamics code meet all the necessary requirements for an end-to-end simulation of the RIA driver linac. The latest version of TRACK was used for end-to-end simulations of the RIA driver including errors and beam loss analysis.** In addition to the standard capabilities, the code includes the following new features: i) multiple charge states ii) realistic stripper model; ii) static and dynamic errors iii) automatic steering to correct for misalignments iv) detailed beam-loss analysis; v) parallel computing to perform large scale simulations. Although primarily developed for simulations of the RIA machine, TRACK is a general beam dynamics code. Currently it is being used for the design and ...

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

Code Samples Used for Complexity and Control

Science.gov (United States)

Ivancevic, Vladimir G.; Reid, Darryn J.

2015-11-01

The following sections are included: * MathematicaⓇ Code * Generic Chaotic Simulator * Vector Differential Operators * NLS Explorer * 2C++ Code * C++ Lambda Functions for Real Calculus * Accelerometer Data Processor * Simple Predictor-Corrector Integrator * Solving the BVP with the Shooting Method * Linear Hyperbolic PDE Solver * Linear Elliptic PDE Solver * Method of Lines for a Set of the NLS Equations * C# Code * Iterative Equation Solver * Simulated Annealing: A Function Minimum * Simple Nonlinear Dynamics * Nonlinear Pendulum Simulator * Lagrangian Dynamics Simulator * Complex-Valued Crowd Attractor Dynamics * Freeform Fortran Code * Lorenz Attractor Simulator * Complex Lorenz Attractor * Simple SGE Soliton * Complex Signal Presentation * Gaussian Wave Packet * Hermitian Matrices * Euclidean L2-Norm * Vector/Matrix Operations * Plain C-Code: Levenberg-Marquardt Optimizer * Free Basic Code: 2D Crowd Dynamics with 3000 Agents

QR code for medical information uses.

Science.gov (United States)

Fontelo, Paul; Liu, Fang; Ducut, Erick G

2008-11-06

We developed QR code online tools, simulated and tested QR code applications for medical information uses including scanning QR code labels, URLs and authentication. Our results show possible applications for QR code in medicine.

Single-phase multi-dimensional thermohydraulics direct numerical simulation code DINUS-3. Input data description

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)

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)

Code modernization and modularization of APEX and SWAT watershed simulation models

Science.gov (United States)

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...

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

Rn3D: A finite element code for simulating gas flow and radon transport in variably saturated, nonisothermal porous media

International Nuclear Information System (INIS)

Holford, D.J.

1994-01-01

This document is a user's manual for the Rn3D finite element code. Rn3D was developed to simulate gas flow and radon transport in variably saturated, nonisothermal porous media. The Rn3D model is applicable to a wide range of problems involving radon transport in soil because it can simulate either steady-state or transient flow and transport in one-, two- or three-dimensions (including radially symmetric two-dimensional problems). The porous materials may be heterogeneous and anisotropic. This manual describes all pertinent mathematics related to the governing, boundary, and constitutive equations of the model, as well as the development of the finite element equations used in the code. Instructions are given for constructing Rn3D input files and executing the code, as well as a description of all output files generated by the code. Five verification problems are given that test various aspects of code operation, complete with example input files, FORTRAN programs for the respective analytical solutions, and plots of model results. An example simulation is presented to illustrate the type of problem Rn3D is designed to solve. Finally, instructions are given on how to convert Rn3D to simulate systems other than radon, air, and water

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.

Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC).

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.

Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC)

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.

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.

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)

Development of Monte Carlo input code for proton, alpha and heavy ion microdosimetric trac structure simulations

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)

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)

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)

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)

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.)

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

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)

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.

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

Numerical simulators are playing an increasingly important role in advancing our fundamental understanding of hydrological systems. They are indispensable tools for managing groundwater resources, analyzing proposed and actual remediation activities at contaminated sites, optimizing recovery of oil, gas, and geothermal energy, evaluating subsurface structures and mining activities, designing monitoring systems, assessing the long-term impacts of chemical and nuclear waste disposal, and devising improved irrigation and drainage practices in agricultural areas, among many other applications. The complexity of subsurface hydrology in the vadose zone calls for sophisticated modeling codes capable of handling the strong nonlinearities involved, the interactions of coupled physical, chemical and biological processes, and the multiscale heterogeneities inherent in such systems. The papers in this special section of ''Vadose Zone Journal'' are illustrative of the enormous potential of such numerical simulators as applied to the vadose zone. The papers describe recent developments and applications of one particular set of codes, the TOUGH family of codes, as applied to nonisothermal flow and transport in heterogeneous porous and fractured media (http://www-esd.lbl.gov/TOUGH2). The contributions were selected from presentations given at the TOUGH Symposium 2003, which brought together developers and users of the TOUGH codes at the Lawrence Berkeley National Laboratory (LBNL) in Berkeley, California, for three days of information exchange in May 2003 (http://www-esd.lbl.gov/TOUGHsymposium). The papers presented at the symposium covered a wide range of topics, including geothermal reservoir engineering, fracture flow and vadose zone hydrology, nuclear waste disposal, mining engineering, reactive chemical transport, environmental remediation, and gas transport. This Special Section of ''Vadose Zone Journal'' contains revised and expanded versions of selected papers from the

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)

MMAPDNG: A new, fast code backed by a memory-mapped database for simulating delayed γ-ray emission with MCNPX package

Science.gov (United States)

Lou, Tak Pui; Ludewigt, Bernhard

2015-09-01

The simulation of the emission of beta-delayed gamma rays following nuclear fission and the calculation of time-dependent energy spectra is a computational challenge. The widely used radiation transport code MCNPX includes a delayed gamma-ray routine that is inefficient and not suitable for simulating complex problems. This paper describes the code "MMAPDNG" (Memory-Mapped Delayed Neutron and Gamma), an optimized delayed gamma module written in C, discusses usage and merits of the code, and presents results. The approach is based on storing required Fission Product Yield (FPY) data, decay data, and delayed particle data in a memory-mapped file. When compared to the original delayed gamma-ray code in MCNPX, memory utilization is reduced by two orders of magnitude and the ray sampling is sped up by three orders of magnitude. Other delayed particles such as neutrons and electrons can be implemented in future versions of MMAPDNG code using its existing framework.

Simulation of Injection Molding Process Including Mold Filling and Compound Curing

Directory of Open Access Journals (Sweden)

Mohamad Reza Erfanian

2012-12-01

Full Text Available The present work reports and discusses the results of a 3D simulation of the injection molding process of a rubber compound that includes the mold flling stage and  material curing, using the computer code is developed in “UDF” part of the Fluent 6.3 CAE software. The data obtained from a rheometer (MDR 2000 is used to characterize the rubber material in order to fnd the cure model parameters which exist in curing model. Because of non-newtonian behavior of rubber, in this work the non-newtonian model for viscosity was used and viscosity parameters were computed by mean of viscometry test by RPA. After calculation of the physical and curing properties, vulcanization process was simulated for a complex rubber article with non-uniform thickness by solving the continuity, momentum, energy and curing process equations. Predicted flling and curing time in a complex and 3D rubber part is compared with experimentally measured data which confrmed  the accuracy and applicability of the method.

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

Enhancing the ABAQUS Thermomechanics Code to Simulate Multidimensional Steady and Transient Fuel Rod Behavior

Energy Technology Data Exchange (ETDEWEB)

Williamson, R.L.; Knoll, D.A. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-3855 (United States)

2009-06-15

Important aspects of fuel rod behavior, for example pellet-clad interaction (PCI), fuel fracture, and non-axisymmetric cooling and oxide formation, are inherently 3-D. Current fuel rod simulation codes typically approximate such behavior using a quasi 2D (or 1.5D) approach and, often, separate codes must be used for steady and transient (or accident) conditions. Notable exceptions are the EPRI propriety code FALCON which is 2D and can be applied to steady or transient operation, and TOUTATIS which is 3D. Recent studies have indicated the need for multidimensional fuel rod simulation capability, particularly for accurate predictions of PCI. The Idaho National Laboratory (INL) is currently developing next-generation capability to model nuclear fuel performance. The goal is to develop a 2D/3D computer code (BISON) which solves the fully coupled thermomechanics equations, includes multi-physics constitutive behavior for both fuel and cladding materials, and is designed for efficient use on highly parallel computers. To provide guidance and a proto-typing environment for this effort, plus provide the INL with near-term fuel modeling capability, the commercially available ABAQUS thermomechanics software has been enhanced to include the fuel behavior phenomena necessary to afford a practical fuel performance simulation capability. This paper details the enhancements which have been implemented in ABAQUS to date, and provides results of a multi-pellet fuel problem which demonstrates the new capability. ABAQUS employs modern finite element methods to solve the nonlinear thermomechanics equations in 1, 2, or 3-D, using linear or quadratic elements. The temperature and displacement fields are solved in a fully-coupled fashion, using sophisticated iteration and time integration error control. The code includes robust contact algorithms, essential for computing multidimensional pellet-pellet or pellet-clad interaction. Extensive constitutive models are available, including

Two-fluid 2.5D code for simulations of small scale magnetic fields in the lower solar atmosphere

Science.gov (United States)

Piantschitsch, Isabell; Amerstorfer, Ute; Thalmann, Julia Katharina; Hanslmeier, Arnold; Lemmerer, Birgit

2015-08-01

Our aim is to investigate magnetic reconnection as a result of the time evolution of magnetic flux tubes in the solar chromosphere. A new numerical two-fluid code was developed, which will perform a 2.5D simulation of the dynamics from the upper convection zone up to the transition region. The code is based on the Total Variation Diminishing Lax-Friedrichs method and includes the effects of ion-neutral collisions, ionisation/recombination, thermal/resistive diffusivity as well as collisional/resistive heating. What is innovative about our newly developed code is the inclusion of a two-fluid model in combination with the use of analytically constructed vertically open magnetic flux tubes, which are used as initial conditions for our simulation. First magnetohydrodynamic (MHD) tests have already shown good agreement with known results of numerical MHD test problems like e.g. the Orszag-Tang vortex test, the Current Sheet test or the Spherical Blast Wave test. Furthermore, the single-fluid approach will also be applied to the initial conditions, in order to compare the different rates of magnetic reconnection in both codes, the two-fluid code and the single-fluid one.

N-body simulations for f(R) gravity using a self-adaptive particle-mesh code

International Nuclear Information System (INIS)

Zhao Gongbo; Koyama, Kazuya; Li Baojiu

2011-01-01

We perform high-resolution N-body simulations for f(R) gravity based on a self-adaptive particle-mesh code MLAPM. The chameleon mechanism that recovers general relativity on small scales is fully taken into account by self-consistently solving the nonlinear equation for the scalar field. We independently confirm the previous simulation results, including the matter power spectrum, halo mass function, and density profiles, obtained by Oyaizu et al.[Phys. Rev. D 78, 123524 (2008)] and Schmidt et al.[Phys. Rev. D 79, 083518 (2009)], and extend the resolution up to k∼20 h/Mpc for the measurement of the matter power spectrum. Based on our simulation results, we discuss how the chameleon mechanism affects the clustering of dark matter and halos on full nonlinear scales.

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)

Schnek: A C++ library for the development of parallel simulation codes on regular grids

Science.gov (United States)

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.

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.

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

A Comparison of Nuclear Power Plant Simulator with RELAP5/MOD3 code about Steam Generator Tube Rupture

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

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.

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)

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

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)

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

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.

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)

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)

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.

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)

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)

Feasibility Study of Coupling the CASMO-4/TABLES-3/SIMULATE-3 Code System to TRACE/PARCS

International Nuclear Information System (INIS)

Demaziere, Christophe; Staalek, Mathias

2004-12-01

This report investigates the feasibility of coupling the Studsvik Scandpower CASMO-4/TABLES-3/SIMULATE-3 codes to the US NRC TRACE/PARCS codes. The data required by TRACE/PARCS are actually the ones necessary to run its neutronic module PARCS. Such data are the macroscopic nuclear cross-sections, some microscopic nuclear cross-sections important for the Xenon and Samarium poisoning effects, the Assembly Discontinuity Factors, and the kinetic parameters. All these data can be retrieved from the Studsvik Scandpower codes. The data functionalization is explained in detail for both systems of codes and the possibility of coupling each of these codes to TRACE/PARCS is discussed. Due to confidentiality restrictions in the use of the CASMO-4 files and to an improper format of the TABLES-3 output files, it is demonstrated that TRACE/PARCS can only be coupled to SIMULATE-3. Specifically-dedicated SIMULATE-3 input decks allow easily editing the neutronic data at specific operating statepoints. Although the data functionalization is different between both systems of codes, such a procedure permits reconstructing a set of data directly compatible with PARCS

A fast and compact Fuel Rod Performance Simulator code for predictive, interpretive and educational purpose

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

Sensitivity Analysis and Uncertainty Quantification for the LAMMPS Molecular Dynamics Simulation Code

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.

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)

GeNN: a code generation framework for accelerated brain simulations

Science.gov (United States)

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/.

SPECT3D - A multi-dimensional collisional-radiative code for generating diagnostic signatures based on hydrodynamics and PIC simulation output

Science.gov (United States)

MacFarlane, J. J.; Golovkin, I. E.; Wang, P.; Woodruff, P. R.; Pereyra, N. A.

2007-05-01

SPECT3D is a multi-dimensional collisional-radiative code used to post-process the output from radiation-hydrodynamics (RH) and particle-in-cell (PIC) codes to generate diagnostic signatures (e.g. images, spectra) that can be compared directly with experimental measurements. This ability to post-process simulation code output plays a pivotal role in assessing the reliability of RH and PIC simulation codes and their physics models. SPECT3D has the capability to operate on plasmas in 1D, 2D, and 3D geometries. It computes a variety of diagnostic signatures that can be compared with experimental measurements, including: time-resolved and time-integrated spectra, space-resolved spectra and streaked spectra; filtered and monochromatic images; and X-ray diode signals. Simulated images and spectra can include the effects of backlighters, as well as the effects of instrumental broadening and time-gating. SPECT3D also includes a drilldown capability that shows where frequency-dependent radiation is emitted and absorbed as it propagates through the plasma towards the detector, thereby providing insights on where the radiation seen by a detector originates within the plasma. SPECT3D has the capability to model a variety of complex atomic and radiative processes that affect the radiation seen by imaging and spectral detectors in high energy density physics (HEDP) experiments. LTE (local thermodynamic equilibrium) or non-LTE atomic level populations can be computed for plasmas. Photoabsorption rates can be computed using either escape probability models or, for selected 1D and 2D geometries, multi-angle radiative transfer models. The effects of non-thermal (i.e. non-Maxwellian) electron distributions can also be included. To study the influence of energetic particles on spectra and images recorded in intense short-pulse laser experiments, the effects of both relativistic electrons and energetic proton beams can be simulated. SPECT3D is a user-friendly software package that runs

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

Simulation of gas mixing and transport in a multi-compartment geometry using the GOTHIC containment code and relatively coarse meshes

International Nuclear Information System (INIS)

Andreani, Michele; Paladino, Domenico

2010-01-01

The recently concluded OECD SETH project included twenty-four experiments on basic flows and gas transport and mixing driven by jets and plumes in two, large, connected vessels of the PANDA facility. The experiments featured injection of saturated or superheated steam, or a mixture of steam and helium in one vessel and venting from the same vessel or from the connected one. These tests have been especially designed for providing an extensive data base for the assessment of three-dimensional codes, including CFD codes. In particular, one of the goals of the analytical activities associated with the experiments was to evaluate the detail of the model (mesh) necessary for capturing the various phenomena. This work reports an overview of the results obtained for these experimental data using the advanced containment code GOTHIC and relatively coarse meshes, which are coarser than the ones typically used for the simulation with commercial CFD codes, but are still representative of the models which are currently affordable for a full containment analysis. In general, the phenomena were correctly represented in the simulations with GOTHIC, and the agreement of the results with the data was in most cases pretty good, in some cases excellent. Only for a few tests (or particular phenomena occurring in some tests) the simulations showed noticeable discrepancies with the experimental data, which could be referred to either an insufficiently detailed mesh or to lack of specialized models for local effects.

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.)

An evaluation of TRAC-PF1/MOD1 computer code performance during posttest simulations of Semiscale MOD-2C feedwater line break transients

International Nuclear Information System (INIS)

Hall, D.G.; Watkins, J.C.

1987-01-01

This report documents an evaluation of the TRAC-PF1/MOD1 reactor safety analysis computer code during computer simulations of feedwater line break transients. The experimental data base for the evaluation included the results of three bottom feedwater line break tests performed in the Semiscale Mod-2C test facility. The tests modeled 14.3% (S-FS-7), 50% (S-FS-11), and 100% (S-FS-6B) breaks. The test facility and the TRAC-PF1/MOD1 model used in the calculations are described. Evaluations of the accuracy of the calculations are presented in the form of comparisons of measured and calculated histories of selected parameters associated with the primary and secondary systems. In addition to evaluating the accuracy of the code calculations, the computational performance of the code during the simulations was assessed. A conclusion was reached that the code is capable of making feedwater line break transient calculations efficiently, but there is room for significant improvements in the simulations that were performed. Recommendations are made for follow-on investigations to determine how to improve future feedwater line break calculations and for code improvements to make the code easier to use

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)

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

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.

Simulations of inspiraling and merging double neutron stars using the Spectral Einstein Code

Science.gov (United States)

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.

EGS code system: computer programs for the Monte Carlo simulation of electromagnetic cascade showers. Version 3. [EGS, PEGS, TESTSR, in MORTRAN

Energy Technology Data Exchange (ETDEWEB)

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. (RWR)

RELAP5-3D Code Includes ATHENA Features and Models

International Nuclear Information System (INIS)

Riemke, Richard A.; Davis, Cliff B.; Schultz, Richard R.

2006-01-01

Version 2.3 of the RELAP5-3D computer program includes all features and models previously available only in the ATHENA version of the code. These include the addition of new working fluids (i.e., ammonia, blood, carbon dioxide, glycerol, helium, hydrogen, lead-bismuth, lithium, lithium-lead, nitrogen, potassium, sodium, and sodium-potassium) and a magnetohydrodynamic model that expands the capability of the code to model many more thermal-hydraulic systems. In addition to the new working fluids along with the standard working fluid water, one or more noncondensable gases (e.g., air, argon, carbon dioxide, carbon monoxide, helium, hydrogen, krypton, nitrogen, oxygen, SF 6 , xenon) can be specified as part of the vapor/gas phase of the working fluid. These noncondensable gases were in previous versions of RELAP5-3D. Recently four molten salts have been added as working fluids to RELAP5-3D Version 2.4, which has had limited release. These molten salts will be in RELAP5-3D Version 2.5, which will have a general release like RELAP5-3D Version 2.3. Applications that use these new features and models are discussed in this paper. (authors)

On-going activities in the European JASMIN project for the development and validation of ASTEC-Na SFR safety simulation code - 15072

International Nuclear Information System (INIS)

Girault, N.; Cloarec, L.; Herranz, L.; Bandini, G.; Perez-Martin, S.; Ammirabile, L.

2015-01-01

The 4-year JASMIN collaborative project (Joint Advanced Severe accidents Modelling and Integration for Na-cooled fast reactors), started in Dec.2011 in the frame of the 7. Framework Programme of the European Commission. It aims at developing a new European simulation code, ASTEC-Na, dealing with the primary phase of SFR core disruptive accidents. The development of a new code, based on a robust advanced simulation tool and able to encompass the in-vessel and in-containment phenomena occurring during a severe accident is indeed of utmost interest for advanced and innovative future SFRs for which an enhanced safety level will be required. This code, based on the ASTEC European code system developed by IRSN and GRS for severe accidents in water-cooled reactors, is progressively integrating and capitalizing the state-of-the-art knowledge of SFR accidents through physical model improvement or development of new ones. New models are assessed on in-pile (CABRI, SCARABEE etc...) and out-of pile experiments conducted during the 70's-80's and code-o-code benchmarking with current accident simulation tools for SFRs is also conducted. During the 2 and a half first years of the project, model specifications and developments were conducted and the validation test matrix was built. The first version of ASTEC-Na available in early 2014 already includes a thermal-hydraulics module able to simulate single and two-phase sodium flow conditions, a zero point neutronic model with simple definition of channel and axial dependences of reactivity feedbacks and models derived from SCANAIR IRSN code for simulating fuel pin thermo-mechanical behaviour and fission gas release/retention. Meanwhile, models have been developed in the source term area for in-containment particle generation and particle chemical transformation, but their implementation is still to be done. As a first validation step, the ASTEC-Na calculations were satisfactorily compared to thermal-hydraulics experimental

Experience with Wolsong-1 Phase-B pre-simulations using WIMS/DRAGON/RFSP-IST code suite

International Nuclear Information System (INIS)

Chung, D-H.; Kim, B-G.; Kim, S-M.; Suh, H-B.; Kim, H-S.; Kim, H-J.

2010-01-01

The Wolsong-1 Phase-B pre-simulations have been carried out with the exclusive use of the code suite WIMS/DRAGON/RFSP-IST in replacement of the previous PPV/MULTICELL/RFSP code system in preparation of tests to be conducted as scheduled in December 2010 after the refurbishment. A comprehensive simulation package has been undertaken starting from the approach to first criticality to the flux measurements and scan. In order to secure the validity of the results, the simulations are performed using both the Uniform and SCM fuel tables. An elaborating contribution has been invested into the work in view of the inexperience of using WIMS/SCM fuel tables as well as incremental cross sections generated by using DRAGON-IST. The overall assessment of simulation results indicates that the newly adopted WIMS/DRAGON/RFSP-IST code suite could be used in replacement of PPV/MULTICELL/RFSP for the verification against the Phase-B test results. (author)

N-body simulations for f(R) gravity using a self-adaptive particle-mesh code

Science.gov (United States)

Zhao, Gong-Bo; Li, Baojiu; Koyama, Kazuya

2011-02-01

We perform high-resolution N-body simulations for f(R) gravity based on a self-adaptive particle-mesh code MLAPM. The chameleon mechanism that recovers general relativity on small scales is fully taken into account by self-consistently solving the nonlinear equation for the scalar field. We independently confirm the previous simulation results, including the matter power spectrum, halo mass function, and density profiles, obtained by Oyaizu [Phys. Rev. DPRVDAQ1550-7998 78, 123524 (2008)10.1103/PhysRevD.78.123524] and Schmidt [Phys. Rev. DPRVDAQ1550-7998 79, 083518 (2009)10.1103/PhysRevD.79.083518], and extend the resolution up to k˜20h/Mpc for the measurement of the matter power spectrum. Based on our simulation results, we discuss how the chameleon mechanism affects the clustering of dark matter and halos on full nonlinear scales.

Particle-in-cell/accelerator code for space-charge dominated beam simulation

Energy Technology Data Exchange (ETDEWEB)

2012-05-08

Warp is a multidimensional discrete-particle beam simulation program designed to be applicable where the beam space-charge is non-negligible or dominant. It is being developed in a collaboration among LLNL, LBNL and the University of Maryland. It was originally designed and optimized for heave ion fusion accelerator physics studies, but has received use in a broader range of applications, including for example laser wakefield accelerators, e-cloud studies in high enery accelerators, particle traps and other areas. At present it incorporates 3-D, axisymmetric (r,z) planar (x-z) and transverse slice (x,y) descriptions, with both electrostatic and electro-magnetic fields, and a beam envelope model. The code is guilt atop the Python interpreter language.

Status of reactor core design code system in COSINE code package

Energy Technology Data Exchange (ETDEWEB)

Chen, Y.; Yu, H.; Liu, Z., E-mail: yuhui@snptc.com.cn [State Nuclear Power Software Development Center, SNPTC, National Energy Key Laboratory of Nuclear Power Software (NEKLS), Beijiing (China)

2014-07-01

For self-reliance, COre and System INtegrated Engine for design and analysis (COSINE) code package is under development in China. In this paper, recent development status of the reactor core design code system (including the lattice physics code and the core simulator) is presented. The well-established theoretical models have been implemented. The preliminary verification results are illustrated. And some special efforts, such as updated theory models and direct data access application, are also made to achieve better software product. (author)

Status of reactor core design code system in COSINE code package

International Nuclear Information System (INIS)

Chen, Y.; Yu, H.; Liu, Z.

2014-01-01

For self-reliance, COre and System INtegrated Engine for design and analysis (COSINE) code package is under development in China. In this paper, recent development status of the reactor core design code system (including the lattice physics code and the core simulator) is presented. The well-established theoretical models have been implemented. The preliminary verification results are illustrated. And some special efforts, such as updated theory models and direct data access application, are also made to achieve better software product. (author)

Coupling the MCNP Monte Carlo code and the FISPACT activation code with automatic visualization of the results of simulations

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.)

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.)

UNSAT-H, an unsaturated soil water flow code for use at the Hanford site: code documentation

International Nuclear Information System (INIS)

Fayer, M.J.; Gee, G.W.

1985-10-01

The unsaturated soil moisture flow code, UNSAT-H, which was developed at Pacific Northwest Laboratory for assessing water movement at waste sites on the Hanford site, is documented in this report. This code is used in simulating the water dynamics of arid sites under consideration for waste disposal. The results of an example simulation of constant infiltration show excellent agreement with an analytical solution and another numerical solution, thus providing some verification of the UNSAT-H code. Areas of the code are identified for future work and include runoff, snowmelt, long-term climate and plant models, and parameter measurement. 29 refs., 7 figs., 2 tabs

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

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)

The Analysis and the Performance Simulation of the Capacity of Bit-interleaved Coded Modulation System

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.

Development of a computational framework on fluid-solid mixture flow simulations for the COMPASS code

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)

Radiation transport simulation in gamma irradiator systems using E G S 4 Monte Carlo code and dose mapping calculations based on point kernel technique

International Nuclear Information System (INIS)

Raisali, G.R.

1992-01-01

A series of computer codes based on point kernel technique and also Monte Carlo method have been developed. These codes perform radiation transport calculations for irradiator systems having cartesian, cylindrical and mixed geometries. The monte Carlo calculations, the computer code 'EGS4' has been applied to a radiation processing type problem. This code has been acompanied by a specific user code. The set of codes developed include: GCELLS, DOSMAPM, DOSMAPC2 which simulate the radiation transport in gamma irradiator systems having cylinderical, cartesian, and mixed geometries, respectively. The program 'DOSMAP3' based on point kernel technique, has been also developed for dose rate mapping calculations in carrier type gamma irradiators. Another computer program 'CYLDETM' as a user code for EGS4 has been also developed to simulate dose variations near the interface of heterogeneous media in gamma irradiator systems. In addition a system of computer codes 'PRODMIX' has been developed which calculates the absorbed dose in the products with different densities. validation studies of the calculated results versus experimental dosimetry has been performed and good agreement has been obtained

The European source-term evaluation code ASTEC: status and applications, including CANDU plant applications

International Nuclear Information System (INIS)

Van Dorsselaere, J.P.; Giordano, P.; Kissane, M.P.; Montanelli, T.; Schwinges, B.; Ganju, S.; Dickson, L.

2004-01-01

Research on light-water reactor severe accidents (SA) is still required in a limited number of areas in order to confirm accident-management plans. Thus, 49 European organizations have linked their SA research in a durable way through SARNET (Severe Accident Research and management NETwork), part of the European 6th Framework Programme. One goal of SARNET is to consolidate the integral code ASTEC (Accident Source Term Evaluation Code, developed by IRSN and GRS) as the European reference tool for safety studies; SARNET efforts include extending the application scope to reactor types other than PWR (including VVER) such as BWR and CANDU. ASTEC is used in IRSN's Probabilistic Safety Analysis level 2 of 900 MWe French PWRs. An earlier version of ASTEC's SOPHAEROS module, including improvements by AECL, is being validated as the Canadian Industry Standard Toolset code for FP-transport analysis in the CANDU Heat Transport System. Work with ASTEC has also been performed by Bhabha Atomic Research Centre, Mumbai, on IPHWR containment thermal hydraulics. (author)

Packing simulation code to calculate distribution function of hard spheres by Monte Carlo method : MCRDF

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)

Enhancing the ABAQUS thermomechanics code to simulate multipellet steady and transient LWR fuel rod behavior

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.

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.

A Novel Technique for Running the NASA Legacy Code LAPIN Synchronously With Simulations Developed Using Simulink

Science.gov (United States)

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.

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

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)

Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

Science.gov (United States)

Williamson, R. L.; Capps, N. A.; Liu, W.; Rashid, Y. R.; Wirth, B. D.

2016-11-01

Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial ( R- Z) or plane radial-circumferential ( R- θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. In comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.

Simulation of lower hybrid current drive in enhanced reversed shear plasmas in the tokamak fusion test reactor using the lower hybrid simulation code

International Nuclear Information System (INIS)

Kaita, R.; Bernabei, S.; Budny, R.

1996-01-01

The Enhanced Reversed Shear (ERS) mode has already shown great potential for improving the performance of the Tokamak Fusion Test Reactor (TFTR) and other devices. Sustaining the ERS, however, remains an outstanding problem. Lower hybrid (LH) current drive is a possible method for modifying the current profile and controlling its time evolution. To predict its effectiveness in TFTR, the Lower Hybrid Simulation Code (LSC) model is used in the TRANSP code and the Tokamak Simulation Code (TSC). Among the results from the simulations are the following. (1) Single-pass absorption is expected in TFTR ERS plasmas. The simulations show that the LH current follows isotherms of the electron temperature. The ability to control the location of the minimum in the q profile (q min ) has been demonstrated by varying the phase velocity of the launched LH waves and observing the change in the damping location. (2) LH current drive can been used to sustain the q min location. The tendency of qmin to drift inward, as the inductive current diffuses during the formation phase of the reversed shear discharge, is prevented by the LH current driven at a fixed radial location. If this results in an expanded plasma volume with improved confinement as high power neutral beam injection is applied, the high bootstrap currents induced during this phase can then maintain the larger qmin radius. (3) There should be no LH wave damping on energetic beam particles. The values of perpendicular index of refraction in the calculations never exceed about 20, while ions at TFR injection energies are resonant with waves having values closer to 100. Other issues being addressed in the study include the LH current drive efficiency in the presence of high bootstrap currents, and the effect of fast electron diffusion on LH current localization

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)

From the direct numerical simulation to system codes-perspective for the multi-scale analysis of LWR thermal hydraulics

International Nuclear Information System (INIS)

Bestion, D.

2010-01-01

A multi-scale analysis of water-cooled reactor thermal hydraulics can be used to take advantage of increased computer power and improved simulation tools, including Direct Numerical Simulation (DNS), Computational Fluid Dynamics (CFD) (in both open and porous mediums), and system thermalhydraulic codes. This paper presents a general strategy for this procedure for various thermalhydraulic scales. A short state of the art is given for each scale, and the role of the scale in the overall multi-scale analysis process is defined. System thermalhydraulic codes will remain a privileged tool for many investigations related to safety. CFD in porous medium is already being frequently used for core thermal hydraulics, either in 3D modules of system codes or in component codes. CFD in open medium allows zooming on some reactor components in specific situations, and may be coupled to the system and component scales. Various modeling approaches exist in the domain from DNS to CFD which may be used to improve the understanding of flow processes, and as a basis for developing more physically based models for macroscopic tools. A few examples are given to illustrate the multi-scale approach. Perspectives for the future are drawn from the present state of the art and directions for future research and development are given

Leveraging Quick Response Code Technology to Facilitate Simulation-Based Leaderboard Competition.

Science.gov (United States)

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.

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

The MCUCN simulation code for ultracold neutron physics

Science.gov (United States)

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.

Advanced methodology to simulate boiling water reactor transient using coupled thermal-hydraulic/neutron-kinetic codes

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

Development of simulation code for MOX dissolution using silver-mediated electrochemical method (Contract research)

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)

Multi-dimensional free-electron laser simulation codes: a comparison study

CERN Document Server

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

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

ATES/heat pump simulations performed with ATESSS code

Science.gov (United States)

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.

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)

DIONISIO 2.0: new version of the code for simulating the behavior of a power fuel rod under irradiation

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

AX-GADGET: a new code for cosmological simulations of Fuzzy Dark Matter and Axion models

Science.gov (United States)

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.

Evaluation of SPACE code for simulation of inadvertent opening of spray valve in Shin Kori unit 1

International Nuclear Information System (INIS)

Kim, Seyun; Youn, Bumsoo

2013-01-01

SPACE code is expected to be applied to the safety analysis for LOCA (Loss of Coolant Accident) and Non-LOCA scenarios. SPACE code solves two-fluid, three-field governing equations and programmed with C++ computer language using object-oriented concepts. To evaluate the analysis capability for the transient phenomena in the actual nuclear power plant, an inadvertent opening of spray valve in startup test phase of Shin Kori unit 1 was simulated with SPACE code. To evaluate the analysis capability for the transient phenomena in the actual nuclear power plant, an inadvertent opening of spray valve in startup test phase of Shin Kori unit 1 was simulated with SPACE code

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

Development of dynamic simulation code for fuel cycle of fusion reactor. 1. Single pulse operation simulation

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)

TERRA: a computer code for simulating the transport of environmentally released radionuclides through agriculture

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

TERRA: a computer code for simulating the transport of environmentally released radionuclides through agriculture

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.

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)

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.

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)

A program code generator for multiphysics biological simulation using markup languages.

Science.gov (United States)

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.

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

Simulating the performance of a distance-3 surface code in a linear ion trap

Science.gov (United States)

Trout, Colin J.; Li, Muyuan; Gutiérrez, Mauricio; Wu, Yukai; Wang, Sheng-Tao; Duan, Luming; Brown, Kenneth R.

2018-04-01

We explore the feasibility of implementing a small surface code with 9 data qubits and 8 ancilla qubits, commonly referred to as surface-17, using a linear chain of 171Yb+ ions. Two-qubit gates can be performed between any two ions in the chain with gate time increasing linearly with ion distance. Measurement of the ion state by fluorescence requires that the ancilla qubits be physically separated from the data qubits to avoid errors on the data due to scattered photons. We minimize the time required to measure one round of stabilizers by optimizing the mapping of the two-dimensional surface code to the linear chain of ions. We develop a physically motivated Pauli error model that allows for fast simulation and captures the key sources of noise in an ion trap quantum computer including gate imperfections and ion heating. Our simulations showed a consistent requirement of a two-qubit gate fidelity of ≥99.9% for the logical memory to have a better fidelity than physical two-qubit operations. Finally, we perform an analysis of the error subsets from the importance sampling method used to bound the logical error rates to gain insight into which error sources are particularly detrimental to error correction.

Status of the development of a fully integrated code system for the simulation of high temperature reactor cores

Energy Technology Data Exchange (ETDEWEB)

Kasselmann, Stefan, E-mail: s.kasselmann@fz-juelich.de [Institute of Energy and Climate Research, Nuclear Waste Management and Reactor Safety (IEK-6), Forschungszentrum Jülich GmbH, 52425 Jülich (Germany); Druska, Claudia [Institute of Energy and Climate Research, Nuclear Waste Management and Reactor Safety (IEK-6), Forschungszentrum Jülich GmbH, 52425 Jülich (Germany); Herber, Stefan [Institute of Energy and Climate Research, Nuclear Waste Management and Reactor Safety (IEK-6), Forschungszentrum Jülich GmbH, 52425 Jülich (Germany); Lehrstuhl für Reaktorsicherheit und -technik, RWTH Aachen, 52062 Aachen (Germany); Jühe, Stephan [Lehrstuhl für Reaktorsicherheit und -technik, RWTH Aachen, 52062 Aachen (Germany); Keller, Florian; Lambertz, Daniela; Li, Jingjing; Scholthaus, Sarah; Shi, Dunfu [Institute of Energy and Climate Research, Nuclear Waste Management and Reactor Safety (IEK-6), Forschungszentrum Jülich GmbH, 52425 Jülich (Germany); Xhonneux, Andre; Allelein, Hans-Josef [Institute of Energy and Climate Research, Nuclear Waste Management and Reactor Safety (IEK-6), Forschungszentrum Jülich GmbH, 52425 Jülich (Germany); Lehrstuhl für Reaktorsicherheit und -technik, RWTH Aachen, 52062 Aachen (Germany)

2014-05-01

The HTR code package (HCP) is a new code system, which couples a variety of stand-alone codes for the simulation of different aspects of HTR. HCP will allow the steady-state and transient operating conditions of a 3D reactor core to be simulated including new features such as spatially resolved fission product release calculations or production and transport of graphite dust. For this code the latest programming techniques and standards are applied. As a first step an object-oriented data model was developed which features a high level of readability because it is based on problem-specific data types like Nuclide, Reaction, ReactionHandler, CrossSectionSet, etc. Those classes help to encapsulate and therefore hide specific implementations, which are not relevant with respect to physics. HCP will make use of one consistent data library for which an automatic generation tool was developed. The new data library consists of decay information, cross sections, fission yields, scattering matrices etc. for all available nuclides (e.g. ENDF/B-VII.1). The data can be stored in different formats such as binary, ASCII or XML. The new burn up code TNT (Topological Nuclide Transmutation) applies graph theory to represent nuclide chains and to minimize the calculation effort when solving the burn up equations. New features are the use of energy-dependent fission yields or the calculation of thermal power for decay, fission and capture reactions. With STACY (source term analysis code system) the fission product release for steady state as well as accident scenarios can be simulated for each fuel batch. For a full-core release calculation several thousand fuel elements are tracked while passing through the core. This models the stochastic behavior of a pebble bed in a realistic manner. In this paper we report on the current status of the HCP and present first results, which prove the applicability of the selected approach.

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)

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)

Neutrons Flux Distributions of the Pu-Be Source and its Simulation by the MCNP-4B Code

Science.gov (United States)

Faghihi, F.; Mehdizadeh, S.; Hadad, K.

Neutron Fluence rate of a low intense Pu-Be source is measured by Neutron Activation Analysis (NAA) of 197Au foils. Also, the neutron fluence rate distribution versus energy is calculated using the MCNP-4B code based on ENDF/B-V library. Theoretical simulation as well as our experimental performance are a new experience for Iranians to make reliability with the code for further researches. In our theoretical investigation, an isotropic Pu-Be source with cylindrical volume distribution is simulated and relative neutron fluence rate versus energy is calculated using MCNP-4B code. Variation of the fast and also thermal neutrons fluence rate, which are measured by NAA method and MCNP code, are compared.

Simulations with the DIONISIO 1.0 Code of Experiments Proposed for the Fumex III Exercise. Final Report

International Nuclear Information System (INIS)

Soba, Alejandro; Denis, Alicia

2013-01-01

This presentation is intended to summarize the results obtained with DIONISIO within the frame of the FUMEX III exercise. From that list we selected those cases that fall within the scope of our code and excluded examples oriented to loss of coolant accidents, like FIO 131, Mt4 and Mt6, or to reactivity initiated accidents, like NSRR, or to oxidation and creep behavior of the cladding under variable stress conditions, like E110 or IFA 585. The models contained in DIONISIO permit the prediction of the temperature distribution in the pellet, cladding and gap, the elastic and plastic stress and strain, creep, swelling and densification, release of fission gases, cesium and iodine to the internal free volume of the rod, gas mixing, pressure increase, irradiation growth of the Zircaloy cladding, development of an oxide layer and hydrogen uptake on the waterside surface, restructuring and grain growth in the pellet. The effects of an internal or external corrosive atmosphere (SCC) as well as the possibility of pellet-cladding interaction (PCI) and crack propagation are also considered. Among the several models included in the code, we consider that two merit some more attention: the thermal description of the gap, since it is determinant of the quality of the simulation of pellet-cladding interaction, and the model of mechanical contact between them designed to predict the failure conditions of the cladding. A summary of both is given below. The participation of our code in the FUMEX III exercise prompted the development of the tools necessary to simulate the fuel behavior in the high burnup range. On the one hand, a subroutine was included in DIONISIO designed to give the radial distribution of the Pu and U isotopes that play a role in the absorption and capture of neutrons in the fuel. As a result, the local distribution of burnup can be obtained with the code. On that basis, the dependence of the thermal conductivity with burnup was included. The code is two

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

Parallel Finite Element Particle-In-Cell Code for Simulations of Space-charge Dominated Beam-Cavity Interactions

International Nuclear Information System (INIS)

Candel, A.; Kabel, A.; Ko, K.; Lee, L.; Li, Z.; Limborg, C.; Ng, C.; Prudencio, E.; Schussman, G.; Uplenchwar, R.

2007-01-01

Over the past years, SLAC's Advanced Computations Department (ACD) has developed the parallel finite element (FE) particle-in-cell code Pic3P (Pic2P) for simulations of beam-cavity interactions dominated by space-charge effects. As opposed to standard space-charge dominated beam transport codes, which are based on the electrostatic approximation, Pic3P (Pic2P) includes space-charge, retardation and boundary effects as it self-consistently solves the complete set of Maxwell-Lorentz equations using higher-order FE methods on conformal meshes. Use of efficient, large-scale parallel processing allows for the modeling of photoinjectors with unprecedented accuracy, aiding the design and operation of the next-generation of accelerator facilities. Applications to the Linac Coherent Light Source (LCLS) RF gun are presented

OECD/NEZ Main Steam Line Break Benchmark Problem Exercise I Simulation Using the SPACE Code with the Point Kinetics Model

International Nuclear Information System (INIS)

Kim, Yohan; Kim, Seyun; Ha, Sangjun

2014-01-01

The Safety and Performance Analysis Code for Nuclear Power Plants (SPACE) has been developed in recent years by the Korea Nuclear Hydro and Nuclear Power Co. (KHNP) through collaborative works with other Korean nuclear industries. The SPACE is a best-estimated two-phase three-field thermal-hydraulic analysis code to analyze the safety and performance of pressurized water reactors (PWRs). The SPACE code has sufficient features to replace outdated vendor supplied codes and to be used for the safety analysis of operating PWRs and the design of advanced reactors. As a result of the second phase of the development, the 2.14 version of the code was released through the successive various V and V works. The topical reports on the code and related safety analysis methodologies have been prepared for license works. In this study, the OECD/NEA Main Steam Line Break (MSLB) Benchmark Problem Exercise I was simulated as a V and V work. The results were compared with those of the participants in the benchmark project. The OECD/NEA MSLB Benchmark Problem Exercise I was simulated using the SPACE code. The results were compared with those of the participants in the benchmark project. Through the simulation, it was concluded that the SPACE code can effectively simulate PWR MSLB accidents

RELAP5/MOD3 code manual: Code structure, system models, and solution methods. Volume 1

International Nuclear Information System (INIS)

1995-08-01

The RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents, and operational transients, such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling, approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. RELAP5/MOD3 code documentation is divided into seven volumes: Volume I provides modeling theory and associated numerical schemes

Simulation of the fault transitory of the feedwater controller in a Boiling water reactor with the Ramona-3B code

International Nuclear Information System (INIS)

Hernandez M, J.L.; Ortiz V, J.

2005-01-01

The obtained results when carrying out the simulation of the fault transitory of the feedwater controller (FCAA) with the Ramona-3B code, happened in the Unit 2 of the Laguna Verde power plant (CNLV), in September of the year 2000 are presented. The transitory originates as consequence of the controller's fault of speed of a turbo pump of feedwater. The work includes a short description of the event, the suppositions considered for the simulation and the obtained results. Also, a discussion of the impact of the transitory event is presented on aspects of reactor safety. Although the carried out simulation is limited by the capacities of the code and for the lack of available information, it was found that even in a conservative situation, the power was incremented only in 12% above the nominal value, while that the thermal limit determined by the minimum reason of the critical power, MCPR, always stayed above the limit values of operation and safety. (Author)

SimProp: a simulation code for ultra high energy cosmic ray propagation

International Nuclear Information System (INIS)

Aloisio, R.; Grillo, A.F.; Boncioli, D.; Petrera, S.; Salamida, F.

2012-01-01

A new Monte Carlo simulation code for the propagation of Ultra High Energy Cosmic Rays is presented. The results of this simulation scheme are tested by comparison with results of another Monte Carlo computation as well as with the results obtained by directly solving the kinetic equation for the propagation of Ultra High Energy Cosmic Rays. A short comparison with the latest flux published by the Pierre Auger collaboration is also presented

Development of a coupled code system based on system transient code, RETRAN, and 3-D neutronics code, MASTER

International Nuclear Information System (INIS)

Kim, K. D.; Jung, J. J.; Lee, S. W.; Cho, B. O.; Ji, S. K.; Kim, Y. H.; Seong, C. K.

2002-01-01

A coupled code system of RETRAN/MASTER has been developed for best-estimate simulations of interactions between reactor core neutron kinetics and plant thermal-hydraulics by incorporation of a 3-D reactor core kinetics analysis code, MASTER into system transient code, RETRAN. The soundness of the consolidated code system is confirmed by simulating the MSLB benchmark problem developed to verify the performance of a coupled kinetics and system transient codes by OECD/NEA

Analysis of quantum error-correcting codes: Symplectic lattice codes and toric codes

Science.gov (United States)

Harrington, James William

Quantum information theory is concerned with identifying how quantum mechanical resources (such as entangled quantum states) can be utilized for a number of information processing tasks, including data storage, computation, communication, and cryptography. Efficient quantum algorithms and protocols have been developed for performing some tasks (e.g. , factoring large numbers, securely communicating over a public channel, and simulating quantum mechanical systems) that appear to be very difficult with just classical resources. In addition to identifying the separation between classical and quantum computational power, much of the theoretical focus in this field over the last decade has been concerned with finding novel ways of encoding quantum information that are robust against errors, which is an important step toward building practical quantum information processing devices. In this thesis I present some results on the quantum error-correcting properties of oscillator codes (also described as symplectic lattice codes) and toric codes. Any harmonic oscillator system (such as a mode of light) can be encoded with quantum information via symplectic lattice codes that are robust against shifts in the system's continuous quantum variables. I show the existence of lattice codes whose achievable rates match the one-shot coherent information over the Gaussian quantum channel. Also, I construct a family of symplectic self-dual lattices and search for optimal encodings of quantum information distributed between several oscillators. Toric codes provide encodings of quantum information into two-dimensional spin lattices that are robust against local clusters of errors and which require only local quantum operations for error correction. Numerical simulations of this system under various error models provide a calculation of the accuracy threshold for quantum memory using toric codes, which can be related to phase transitions in certain condensed matter models. I also present

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)

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)

A New Code SORD for Simulation of Polarized Light Scattering in the Earth Atmosphere

Science.gov (United States)

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/.

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)

Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC): gap analysis for high fidelity and performance assessment code development

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

Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC) : gap analysis for high fidelity and performance assessment code development.

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

Icarus: A 2-D Direct Simulation Monte Carlo (DSMC) Code for Multi-Processor Computers

International Nuclear Information System (INIS)

BARTEL, TIMOTHY J.; PLIMPTON, STEVEN J.; GALLIS, MICHAIL A.

2001-01-01

Icarus is a 2D Direct Simulation Monte Carlo (DSMC) code which has been optimized for the parallel computing environment. The code is based on the DSMC method of Bird[11.1] and models from free-molecular to continuum flowfields in either cartesian (x, y) or axisymmetric (z, r) coordinates. Computational particles, representing a given number of molecules or atoms, are tracked as they have collisions with other particles or surfaces. Multiple species, internal energy modes (rotation and vibration), chemistry, and ion transport are modeled. A new trace species methodology for collisions and chemistry is used to obtain statistics for small species concentrations. Gas phase chemistry is modeled using steric factors derived from Arrhenius reaction rates or in a manner similar to continuum modeling. Surface chemistry is modeled with surface reaction probabilities; an optional site density, energy dependent, coverage model is included. Electrons are modeled by either a local charge neutrality assumption or as discrete simulational particles. Ion chemistry is modeled with electron impact chemistry rates and charge exchange reactions. Coulomb collision cross-sections are used instead of Variable Hard Sphere values for ion-ion interactions. The electro-static fields can either be: externally input, a Langmuir-Tonks model or from a Green's Function (Boundary Element) based Poison Solver. Icarus has been used for subsonic to hypersonic, chemically reacting, and plasma flows. The Icarus software package includes the grid generation, parallel processor decomposition, post-processing, and restart software. The commercial graphics package, Tecplot, is used for graphics display. All of the software packages are written in standard Fortran

GAMERA - The New Magnetospheric Code

Science.gov (United States)

Lyon, J.; Sorathia, K.; Zhang, B.; Merkin, V. G.; Wiltberger, M. J.; Daldorff, L. K. S.

2017-12-01

The Lyon-Fedder-Mobarry (LFM) code has been a main-line magnetospheric simulation code for 30 years. The code base, designed in the age of memory to memory vector ma- chines,is still in wide use for science production but needs upgrading to ensure the long term sustainability. In this presentation, we will discuss our recent efforts to update and improve that code base and also highlight some recent results. The new project GAM- ERA, Grid Agnostic MHD for Extended Research Applications, has kept the original design characteristics of the LFM and made significant improvements. The original de- sign included high order numerical differencing with very aggressive limiting, the ability to use arbitrary, but logically rectangular, grids, and maintenance of div B = 0 through the use of the Yee grid. Significant improvements include high-order upwinding and a non-clipping limiter. One other improvement with wider applicability is an im- proved averaging technique for the singularities in polar and spherical grids. The new code adopts a hybrid structure - multi-threaded OpenMP with an overarching MPI layer for large scale and coupled applications. The MPI layer uses a combination of standard MPI and the Global Array Toolkit from PNL to provide a lightweight mechanism for coupling codes together concurrently. The single processor code is highly efficient and can run magnetospheric simulations at the default CCMC resolution faster than real time on a MacBook pro. We have run the new code through the Athena suite of tests, and the results compare favorably with the codes available to the astrophysics community. LFM/GAMERA has been applied to many different situations ranging from the inner and outer heliosphere and magnetospheres of Venus, the Earth, Jupiter and Saturn. We present example results the Earth's magnetosphere including a coupled ring current (RCM), the magnetospheres of Jupiter and Saturn, and the inner heliosphere.

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

Monte Carlo FLUKA code simulation for study of {sup 68}Ga production by direct proton-induced reaction

Energy Technology Data Exchange (ETDEWEB)

Mokhtari Oranj, Leila; Kakavand, Tayeb [Physics Faculty, Zanjan University, P.O. Box 451-313, Zanjan (Iran, Islamic Republic of); Sadeghi, Mahdi, E-mail: msadeghi@nrcam.org [Agricultural, Medical and Industrial Research School, Nuclear Science and Technology Research Institute, P.O. Box 31485-498, Karaj (Iran, Islamic Republic of); Aboudzadeh Rovias, Mohammadreza [Agricultural, Medical and Industrial Research School, Nuclear Science and Technology Research Institute, P.O. Box 31485-498, Karaj (Iran, Islamic Republic of)

2012-06-11

{sup 68}Ga is an important radionuclide for positron emission tomography. {sup 68}Ga can be produced by the {sup 68}Zn(p,n){sup 68}Ga reaction in a common biomedical cyclotrons. To facilitate optimization of target design and study activation of materials, Monte Carlo code can be used to simulate the irradiation of the target materials with charged hadrons. In this paper, FLUKA code simulation was employed to prototype a Zn target for the production of {sup 68}Ga by proton irradiation. Furthermore, the experimental data were compared with the estimated values for the thick target yield produced in the irradiation time according to FLUKA code. In conclusion, FLUKA code can be used for estimation of the production yield.

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

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)

LDPC Codes with Minimum Distance Proportional to Block Size

Science.gov (United States)

Divsalar, Dariush; Jones, Christopher; Dolinar, Samuel; Thorpe, Jeremy

2009-01-01

Low-density parity-check (LDPC) codes characterized by minimum Hamming distances proportional to block sizes have been demonstrated. Like the codes mentioned in the immediately preceding article, the present codes are error-correcting codes suitable for use in a variety of wireless data-communication systems that include noisy channels. The previously mentioned codes have low decoding thresholds and reasonably low error floors. However, the minimum Hamming distances of those codes do not grow linearly with code-block sizes. Codes that have this minimum-distance property exhibit very low error floors. Examples of such codes include regular LDPC codes with variable degrees of at least 3. Unfortunately, the decoding thresholds of regular LDPC codes are high. Hence, there is a need for LDPC codes characterized by both low decoding thresholds and, in order to obtain acceptably low error floors, minimum Hamming distances that are proportional to code-block sizes. The present codes were developed to satisfy this need. The minimum Hamming distances of the present codes have been shown, through consideration of ensemble-average weight enumerators, to be proportional to code block sizes. As in the cases of irregular ensembles, the properties of these codes are sensitive to the proportion of degree-2 variable nodes. A code having too few such nodes tends to have an iterative decoding threshold that is far from the capacity threshold. A code having too many such nodes tends not to exhibit a minimum distance that is proportional to block size. Results of computational simulations have shown that the decoding thresholds of codes of the present type are lower than those of regular LDPC codes. Included in the simulations were a few examples from a family of codes characterized by rates ranging from low to high and by thresholds that adhere closely to their respective channel capacity thresholds; the simulation results from these examples showed that the codes in question have low

The gravitational interaction between N-body (star clusters) and hydrodynamic (ISM) codes in disk galaxy simulations

International Nuclear Information System (INIS)

Schroeder, M.C.; Comins, N.F.

1986-01-01

During the past twenty years, three approaches to numerical simulations of the evolution of galaxies have been developed. The first approach, N-body programs, models the motion of clusters of stars as point particles which interact via their gravitational potentials to determine the system dynamics. Some N-body codes model molecular clouds as colliding, inelastic particles. The second approach, hydrodynamic models of galactic dynamics, simulates the activity of the interstellar medium as a compressible gas. These models presently do not include stars, the effect of gravitational fields, or allow for stellar evolution and exchange of mass or angular momentum between stars and the interstellar medium. The third approach, stochastic star formation simulations of disk galaxies, allows for the interaction between stars and interstellar gas, but does not allow the star particles to move under the influence of gravity

TRANP - a computer code for digital simulation of steady - state and transient behavior of a pressurizer water reactor primary circuit

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

FENICIA: a generic plasma simulation code using a flux-independent field-aligned coordinate approach

International Nuclear Information System (INIS)

Hariri, Farah

2013-01-01

The primary thrust of this work is the development and implementation of a new approach to the problem of field-aligned coordinates in magnetized plasma turbulence simulations called the FCI approach (Flux-Coordinate Independent). The method exploits the elongated nature of micro-instability driven turbulence which typically has perpendicular scales on the order of a few ion gyro-radii, and parallel scales on the order of the machine size. Mathematically speaking, it relies on local transformations that align a suitable coordinate to the magnetic field to allow efficient computation of the parallel derivative. However, it does not rely on flux coordinates, which permits discretizing any given field on a regular grid in the natural coordinates such as (x, y, z) in the cylindrical limit. The new method has a number of advantages over methods constructed starting from flux coordinates, allowing for more flexible coding in a variety of situations including X-point configurations. In light of these findings, a plasma simulation code FENICIA has been developed based on the FCI approach with the ability to tackle a wide class of physical models. The code has been verified on several 3D test models. The accuracy of the approach is tested in particular with respect to the question of spurious radial transport. Tests on 3D models of the drift wave propagation and of the Ion Temperature Gradient (ITG) instability in cylindrical geometry in the linear regime demonstrate again the high quality of the numerical method. Finally, the FCI approach is shown to be able to deal with an X-point configuration such as one with a magnetic island with good convergence and conservation properties. (author) [fr

A simple method for simulation of coherent synchrotron radiation in a tracking code

International Nuclear Information System (INIS)

Borland, M.

2000-01-01

Coherent synchrotron radiation (CSR) is of great interest to those designing accelerators as drivers for free-electron lasers (FELs). Although experimental evidence is incomplete, CSR is predicted to have potentially severe effects on the emittance of high-brightness electron beams. The performance of an FEL depends critically on the emittance, current, and energy spread of the beam. Attempts to increase the current through magnetic bunch compression can lead to increased emittance and energy spread due to CSR in the dipoles of such a compressor. The code elegant was used for design and simulation of the bunch compressor for the Low-Energy Undulator Test Line (LEUTL) FEL at the Advanced Photon Source (APS). In order to facilitate this design, a fast algorithm was developed based on the 1-D formalism of Saldin and coworkers. In addition, a plausible method of including CSR effects in drift spaces following the chicane magnets was developed and implemented. The algorithm is fast enough to permit running hundreds of tolerance simulations including CSR for 50 thousand particles. This article describes the details of the implementation and shows results for the APS bunch compressor

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.

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)

Monte Carlo code for neutron radiography

International Nuclear Information System (INIS)

Milczarek, Jacek J.; Trzcinski, Andrzej; El-Ghany El Abd, Abd; Czachor, Andrzej

2005-01-01

The concise Monte Carlo code, MSX, for simulation of neutron radiography images of non-uniform objects is presented. The possibility of modeling the images of objects with continuous spatial distribution of specific isotopes is included. The code can be used for assessment of the scattered neutron component in neutron radiograms

Monte Carlo code for neutron radiography

Energy Technology Data Exchange (ETDEWEB)

Milczarek, Jacek J. [Institute of Atomic Energy, Swierk, 05-400 Otwock (Poland)]. E-mail: jjmilcz@cyf.gov.pl; Trzcinski, Andrzej [Institute for Nuclear Studies, Swierk, 05-400 Otwock (Poland); El-Ghany El Abd, Abd [Institute of Atomic Energy, Swierk, 05-400 Otwock (Poland); Nuclear Research Center, PC 13759, Cairo (Egypt); Czachor, Andrzej [Institute of Atomic Energy, Swierk, 05-400 Otwock (Poland)

2005-04-21

The concise Monte Carlo code, MSX, for simulation of neutron radiography images of non-uniform objects is presented. The possibility of modeling the images of objects with continuous spatial distribution of specific isotopes is included. The code can be used for assessment of the scattered neutron component in neutron radiograms.

Coded diffraction system in X-ray crystallography using a boolean phase coded aperture approximation

Science.gov (United States)

Pinilla, Samuel; Poveda, Juan; Arguello, Henry

2018-03-01

Phase retrieval is a problem present in many applications such as optics, astronomical imaging, computational biology and X-ray crystallography. Recent work has shown that the phase can be better recovered when the acquisition architecture includes a coded aperture, which modulates the signal before diffraction, such that the underlying signal is recovered from coded diffraction patterns. Moreover, this type of modulation effect, before the diffraction operation, can be obtained using a phase coded aperture, just after the sample under study. However, a practical implementation of a phase coded aperture in an X-ray application is not feasible, because it is computationally modeled as a matrix with complex entries which requires changing the phase of the diffracted beams. In fact, changing the phase implies finding a material that allows to deviate the direction of an X-ray beam, which can considerably increase the implementation costs. Hence, this paper describes a low cost coded X-ray diffraction system based on block-unblock coded apertures that enables phase reconstruction. The proposed system approximates the phase coded aperture with a block-unblock coded aperture by using the detour-phase method. Moreover, the SAXS/WAXS X-ray crystallography software was used to simulate the diffraction patterns of a real crystal structure called Rhombic Dodecahedron. Additionally, several simulations were carried out to analyze the performance of block-unblock approximations in recovering the phase, using the simulated diffraction patterns. Furthermore, the quality of the reconstructions was measured in terms of the Peak Signal to Noise Ratio (PSNR). Results show that the performance of the block-unblock phase coded apertures approximation decreases at most 12.5% compared with the phase coded apertures. Moreover, the quality of the reconstructions using the boolean approximations is up to 2.5 dB of PSNR less with respect to the phase coded aperture reconstructions.

SERPENT Monte Carlo reactor physics code

International Nuclear Information System (INIS)

Leppaenen, J.

2010-01-01

SERPENT is a three-dimensional continuous-energy Monte Carlo reactor physics burnup calculation code, developed at VTT Technical Research Centre of Finland since 2004. The code is specialized in lattice physics applications, but the universe-based geometry description allows transport simulation to be carried out in complicated three-dimensional geometries as well. The suggested applications of SERPENT include generation of homogenized multi-group constants for deterministic reactor simulator calculations, fuel cycle studies involving detailed assembly-level burnup calculations, validation of deterministic lattice transport codes, research reactor applications, educational purposes and demonstration of reactor physics phenomena. The Serpent code has been publicly distributed by the OECD/NEA Data Bank since May 2009 and RSICC in the U. S. since March 2010. The code is being used in some 35 organizations in 20 countries around the world. This paper presents an overview of the methods and capabilities of the Serpent code, with examples in the modelling of WWER-440 reactor physics. (Author)

A computer code simulating multistage chemical exchange column under wide range of operating conditions

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)

Simulation of nonlinear propagation of biomedical ultrasound using PZFlex and the KZK Texas code

Science.gov (United States)

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

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.

Atucha II NPP full scope simulator modelling with the thermal hydraulic code TRAC{sub R}T

Energy Technology Data Exchange (ETDEWEB)

Alonso, Pablo Rey; Ruiz, Jose Antonio; Rivero, Norberto, E-mail: prey@tecnatom.e, E-mail: jaruiz@tecnatom.e, E-mail: nrivero@tecnatom.e [Tecnatom S.A., Madrid (Spain)

2011-07-01

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{sub 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{sub 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{sub 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)

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)

Steady-State Gyrokinetics Transport Code (SSGKT), A Scientific Application Partnership with the Framework Application for Core-Edge Transport Simulations, Final Report

Energy Technology Data Exchange (ETDEWEB)

Fahey, Mark R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Candy, Jeff [General Atomics, San Diego, CA (United States)

2013-11-07

This project initiated the development of TGYRO - a steady-state Gyrokinetic transport code (SSGKT) that integrates micro-scale GYRO turbulence simulations into a framework for practical multi-scale simulation of conventional tokamaks as well as future reactors. Using a lightweight master transport code, multiple independent (each massively parallel) gyrokinetic simulations are coordinated. The capability to evolve profiles using the TGLF model was also added to TGYRO and represents a more typical use-case for TGYRO. The goal of the project was to develop a steady-state Gyrokinetic transport code (SSGKT) that integrates micro-scale gyrokinetic turbulence simulations into a framework for practical multi-scale simulation of a burning plasma core ? the International Thermonuclear Experimental Reactor (ITER) in particular. This multi-scale simulation capability will be used to predict the performance (the fusion energy gain, Q) given the H-mode pedestal temperature and density. At present, projections of this type rely on transport models like GLF23, which are based on rather approximate fits to the results of linear and nonlinear simulations. Our goal is to make these performance projections with precise nonlinear gyrokinetic simulations. The method of approach is to use a lightweight master transport code to coordinate multiple independent (each massively parallel) gyrokinetic simulations using the GYRO code. This project targets the practical multi-scale simulation of a reactor core plasma in order to predict the core temperature and density profiles given the H-mode pedestal temperature and density. A master transport code will provide feedback to O(16) independent gyrokinetic simulations (each massively parallel). A successful feedback scheme offers a novel approach to predictive modeling of an important national and international problem. Success in this area of fusion simulations will allow US scientists to direct the research path of ITER over the next two

Development of a severe accident training simulator using a MELCOR code

International Nuclear Information System (INIS)

Kim, Ko Ryu; Jeong, Kwang Sub; Ha, Jae Joo; Jung, Won Dae

2002-03-01

Nuclear power plants' severe accidents are, despite of their rareness, very important in safety aspects, because of their huge damages when occurred. For the appropriate execution of severe accident strategy, more information for decision-making are required because of the uncertainties included in severe accidents. Earlier NRC raised concerns over severe accident training in the report NUREC/CR-477, and accordingly, developing effective training tools for severe accident were emphasized. In fact the training tools were requested from industrial area, nevertheless, few training tools were developed due to the uncertainties in severe accidents, lacks of analysis computer codes and technical limitations. SATS, the severe accident training simulator, is developed as a multi-purpose tools for severe accident training. SATS uses the calculation results of MELCOR, an integral severe accident analysis code, and with the help of SL-GMS graphic tools, provides dynamic displays of severe accident phenomena on the terminal of IBM PC. It aimed to have two main features: one is to provide graphic displays to represent severe accident phenomena and the other is to process and simulate severe accident strategy given by plant operators and TSC staffs. Severe accident strategies are basically composed of series of operations of available pumps, valves and other equipments. Whenever executing strategies with SATS, the trainee should follow the HyperKAMG, the on line version of the recently developed severe accident guidance (KAMG). Severe accident strategies are closely related to accidents scenarios. TLOFW and LOCA , two representative severe accident scenarios of Uljin 3,4, are developed as a built-in scenarios of SATS. Although SATS has some minor problems at this time, we expect SATS will be a good severe accident training tool after the appropriate addition of accident scenarios. Moreover, we also expect SATS will be a good advisory tool for the severe accident research

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

Development of integrated SOL/Divertor code and simulation study of the JT-60U/JT-60SA tokamaks

International Nuclear Information System (INIS)

Kawashima, H.; Shimizu, K.; Takizuka, T.

2007-01-01

To predict the particle and heat controllability in the divertor of tokamak reactors such as ITER and to optimize the divertor design, comprehensive simulations by integrated modelling with taking in various physical processes are indispensable. For the design study of ITER divertor, the modelling codes such as B2, UEDGE and EDGE2D have been developed, and their results have contributed to the evolution of the divertor concept. In Japan Atomic Energy Agency (JAEA), SOL/divertor codes have also been developed for the interpretation and the prediction on behaviours of plasmas, neutrals and impurities in the SOL/divertor regions. The code development is originally carried out since physics models can be verified quickly and flexibly under the circumstance of close collaboration with JT-60 team. Figure 1 shows our code system, which consists of the 2 dimensional fluid code SOLDOR, the neutral Monte Carlo (MC) code NEUT2D, and the impurity MC code IMPMC. The particle simulation code PARASOL has also been developed in order to establish the physics modelling used in fluid simulations. Integration of SOLDOR, NEUT2D and IMPMC, called the '' SONIC '' code, is being carried out to simulate self-consistently the SOL/divertor plasmas in present tokamaks and in future devices. Combination of the SOLDOR and NEUT2D was completed, which has the features such as 1) high-resolution oscillation-free scheme in solving fluid equations, 2) neutral transport calculation under the fine meshes, 3) success in reduction of MC noise, 4) optimization on the massive parallel computer, etc. The simulation reproduces the X-point MARFE in the JT-60U experiment. It is found that the chemically sputtered carbon at the dome causes the radiation peaking near the X-point. The performance of divertor pumping in JT-60U is evaluated from the particle balances. We also present the divertor designing of JT-60SA, which is the modification program of JT-60U to establish high beta steady-state operation. To

An information theoretic approach to use high-fidelity codes to calibrate low-fidelity codes

Energy Technology Data Exchange (ETDEWEB)

Lewis, Allison, E-mail: lewis.allison10@gmail.com [Department of Mathematics, North Carolina State University, Raleigh, NC 27695 (United States); Smith, Ralph [Department of Mathematics, North Carolina State University, Raleigh, NC 27695 (United States); Williams, Brian [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Figueroa, Victor [Sandia National Laboratories, Albuquerque, NM 87185 (United States)

2016-11-01

For many simulation models, it can be prohibitively expensive or physically infeasible to obtain a complete set of experimental data to calibrate model parameters. In such cases, one can alternatively employ validated higher-fidelity codes to generate simulated data, which can be used to calibrate the lower-fidelity code. In this paper, we employ an information-theoretic framework to determine the reduction in parameter uncertainty that is obtained by evaluating the high-fidelity code at a specific set of design conditions. These conditions are chosen sequentially, based on the amount of information that they contribute to the low-fidelity model parameters. The goal is to employ Bayesian experimental design techniques to minimize the number of high-fidelity code evaluations required to accurately calibrate the low-fidelity model. We illustrate the performance of this framework using heat and diffusion examples, a 1-D kinetic neutron diffusion equation, and a particle transport model, and include initial results from the integration of the high-fidelity thermal-hydraulics code Hydra-TH with a low-fidelity exponential model for the friction correlation factor.

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

MCNP6.1 simulations for low-energy atomic relaxation: Code-to-code comparison with GATEv7.2, PENELOPE2014, and EGSnrc

Science.gov (United States)

Jung, Seongmoon; Sung, Wonmo; Lee, Jaegi; Ye, Sung-Joon

2018-01-01

Emerging radiological applications of gold nanoparticles demand low-energy electron/photon transport calculations including details of an atomic relaxation process. Recently, MCNP® version 6.1 (MCNP6.1) has been released with extended cross-sections for low-energy electron/photon, subshell photoelectric cross-sections, and more detailed atomic relaxation data than the previous versions. With this new feature, the atomic relaxation process of MCNP6.1 has not been fully tested yet with its new physics library (eprdata12) that is based on the Evaluated Atomic Data Library (EADL). In this study, MCNP6.1 was compared with GATEv7.2, PENELOPE2014, and EGSnrc that have been often used to simulate low-energy atomic relaxation processes. The simulations were performed to acquire both photon and electron spectra produced by interactions of 15 keV electrons or photons with a 10-nm-thick gold nano-slab. The photon-induced fluorescence X-rays from MCNP6.1 fairly agreed with those from GATEv7.2 and PENELOPE2014, while the electron-induced fluorescence X-rays of the four codes showed more or less discrepancies. A coincidence was observed in the photon-induced Auger electrons simulated by MCNP6.1 and GATEv7.2. A recent release of MCNP6.1 with eprdata12 can be used to simulate the photon-induced atomic relaxation.

Coding Instructions, Worksheets, and Keypunch Sheets for M.E.T.R.O.-APEX Simulation.

Science.gov (United States)

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…

Joint design of QC-LDPC codes for coded cooperation system with joint iterative decoding

Science.gov (United States)

Zhang, Shunwai; Yang, Fengfan; Tang, Lei; Ejaz, Saqib; Luo, Lin; Maharaj, B. T.

2016-03-01

In this paper, we investigate joint design of quasi-cyclic low-density-parity-check (QC-LDPC) codes for coded cooperation system with joint iterative decoding in the destination. First, QC-LDPC codes based on the base matrix and exponent matrix are introduced, and then we describe two types of girth-4 cycles in QC-LDPC codes employed by the source and relay. In the equivalent parity-check matrix corresponding to the jointly designed QC-LDPC codes employed by the source and relay, all girth-4 cycles including both type I and type II are cancelled. Theoretical analysis and numerical simulations show that the jointly designed QC-LDPC coded cooperation well combines cooperation gain and channel coding gain, and outperforms the coded non-cooperation under the same conditions. Furthermore, the bit error rate performance of the coded cooperation employing jointly designed QC-LDPC codes is better than those of random LDPC codes and separately designed QC-LDPC codes over AWGN channels.

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.)

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)

Light-water-reactor coupled neutronic and thermal-hydraulic codes

International Nuclear Information System (INIS)

Diamond, D.J.

1982-01-01

An overview is presented of computer codes that model light water reactor cores with coupled neutronics and thermal-hydraulics. This includes codes for transient analysis and codes for steady state analysis which include fuel depletion and fission product buildup. Applications in nuclear design, reactor operations and safety analysis are given and the major codes in use in the USA are identified. The neutronic and thermal-hydraulic methodologies and other code features are outlined for three steady state codes (PDQ7, NODE-P/B and SIMULATE) and four dynamic codes (BNL-TWIGL, MEKIN, RAMONA-3B, RETRAN-02). Speculation as to future trends with such codes is also presented

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.)

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.

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

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

WEC3: Wave Energy Converter Code Comparison Project: Preprint

Energy Technology Data Exchange (ETDEWEB)

Combourieu, Adrien; Lawson, Michael; Babarit, Aurelien; Ruehl, Kelley; Roy, Andre; Costello, Ronan; Laporte Weywada, Pauline; Bailey, Helen

2017-01-01

This paper describes the recently launched Wave Energy Converter Code Comparison (WEC3) project and present preliminary results from this effort. The objectives of WEC3 are to verify and validate numerical modelling tools that have been developed specifically to simulate wave energy conversion devices and to inform the upcoming IEA OES Annex VI Ocean Energy Modelling Verification and Validation project. WEC3 is divided into two phases. Phase 1 consists of a code-to-code verification and Phase II entails code-to-experiment validation. WEC3 focuses on mid-fidelity codes that simulate WECs using time-domain multibody dynamics methods to model device motions and hydrodynamic coefficients to model hydrodynamic forces. Consequently, high-fidelity numerical modelling tools, such as Navier-Stokes computational fluid dynamics simulation, and simple frequency domain modelling tools were not included in the WEC3 project.

V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

Energy Technology Data Exchange (ETDEWEB)

Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

2010-07-15

V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

International Nuclear Information System (INIS)

Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

2010-07-01

V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

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.

Depletion methodology in the 3-D whole core transport code DeCART

Energy Technology Data Exchange (ETDEWEB)

Kim, Kang Seog; Cho, Jin Young; Zee, Sung Quun

2005-02-01

Three dimensional whole-core transport code DeCART has been developed to include a characteristics of the numerical reactor to replace partly the experiment. This code adopts the deterministic method in simulating the neutron behavior with the least assumption and approximation. This neutronic code is also coupled with the thermal hydraulic code CFD and the thermo mechanical code to simulate the combined effects. Depletion module has been implemented in DeCART code to predict the depleted composition in the fuel. The exponential matrix method of ORIGEN-2 has been used for the depletion calculation. The library of including decay constants, yield matrix and others has been used and greatly simplified for the calculation efficiency. This report summarizes the theoretical backgrounds and includes the verification of the depletion module in DeCART by performing the benchmark calculations.

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)

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)

Tornado missile simulation and design methodology. Volume 1: simulation methodology, design applications, and TORMIS computer code. Final report

International Nuclear Information System (INIS)

Twisdale, L.A.; Dunn, W.L.

1981-08-01

A probabilistic methodology has been developed to predict the probabilities of tornado-propelled missiles impacting and damaging nuclear power plant structures. Mathematical models of each event in the tornado missile hazard have been developed and sequenced to form an integrated, time-history simulation methodology. The models are data based where feasible. The data include documented records of tornado occurrence, field observations of missile transport, results of wind tunnel experiments, and missile impact tests. Probabilistic Monte Carlo techniques are used to estimate the risk probabilities. The methodology has been encoded in the TORMIS computer code to facilitate numerical analysis and plant-specific tornado missile probability assessments. Sensitivity analyses have been performed on both the individual models and the integrated methodology, and risk has been assessed for a hypothetical nuclear power plant design case study

Development of a NSSS T/H Module for the YGN 1/2 NPP Simulator Using a Best-Estimate Code, RETRAN

International Nuclear Information System (INIS)

Seo, I. Y.; Lee, Y. K.; Jeun, G. D.; Suh, J. S.

2005-01-01

KEPRI(Korea Electric Power Research Institute) developed a realistic nuclear steam supply system thermal-hydraulic module, named ARTS code, based on the best-estimate code RETRAN for the improvement of the KNPEC(Korea Nuclear Plant Education Center) unit 2 full-scope simulator. In this work, we make a nuclear steam supply system thermal-hydraulic module for the YGN 1/2 nuclear power plant simulator using a practical application of a experience of ARTS code development. The ARTS code was developed based on RETRAN, which is a best estimate code developed by EPRI(Electric Power Research Institute) for various transient analyses of NPP(Nuclear Power Plants). Robustness and the real time calculation capability have been improved by simplifications, removing of discontinuities of the physical correlations of the RETRAN code and some other modifications. And its scope for the simulation has been extended by supplementation of new calculation modules such as a dedicated pressurizer relief tank model and a backup model. The supplement is developed so that users cannot recognize the model change from the main ARTS module

Diagonal Eigenvalue Unity (DEU) code for spectral amplitude coding-optical code division multiple access

Science.gov (United States)

Ahmed, Hassan Yousif; Nisar, K. S.

2013-08-01

Code with ideal in-phase cross correlation (CC) and practical code length to support high number of users are required in spectral amplitude coding-optical code division multiple access (SAC-OCDMA) systems. SAC systems are getting more attractive in the field of OCDMA because of its ability to eliminate the influence of multiple access interference (MAI) and also suppress the effect of phase induced intensity noise (PIIN). In this paper, we have proposed new Diagonal Eigenvalue Unity (DEU) code families with ideal in-phase CC based on Jordan block matrix with simple algebraic ways. Four sets of DEU code families based on the code weight W and number of users N for the combination (even, even), (even, odd), (odd, odd) and (odd, even) are constructed. This combination gives DEU code more flexibility in selection of code weight and number of users. These features made this code a compelling candidate for future optical communication systems. Numerical results show that the proposed DEU system outperforms reported codes. In addition, simulation results taken from a commercial optical systems simulator, Virtual Photonic Instrument (VPI™) shown that, using point to multipoint transmission in passive optical network (PON), DEU has better performance and could support long span with high data rate.

A development and an application of Mixset-X computer code for simulating the Purex solvent extraction system

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 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

Simulation of the containment spray system test PACOS PX2.2 with the integral code ASTEC and the containment code system COCOSYS

International Nuclear Information System (INIS)

Risken, Tobias; Koch, Marco K.

2011-01-01

The reactor safety research contains the analysis of postulated accidents in nuclear power plants (npp). These accidents may involve a loss of coolant from the nuclear plant's reactor coolant system, during which heat and pressure within the containment are increased. To handle these atmospheric conditions, containment spray systems are installed in various light water reactors (LWR) worldwide as a part of the accident management system. For the improvement and the safety ensurance in npp operation and accident management, numeric simulations of postulated accident scenarios are performed. The presented calculations regard the predictability of the containment spray system's effect with the integral code ASTEC and the containment code system COCOSYS, performed at Ruhr-Universitaet Bochum. Therefore the test PACOS Px2.2 is simulated, in which water is sprayed in the stratified containment atmosphere of the BMC (Battelle Modell-Containment). (orig.)

Kinetic models of gene expression including non-coding RNAs

Energy Technology Data Exchange (ETDEWEB)

Zhdanov, Vladimir P., E-mail: zhdanov@catalysis.r

2011-03-15

In cells, genes are transcribed into mRNAs, and the latter are translated into proteins. Due to the feedbacks between these processes, the kinetics of gene expression may be complex even in the simplest genetic networks. The corresponding models have already been reviewed in the literature. A new avenue in this field is related to the recognition that the conventional scenario of gene expression is fully applicable only to prokaryotes whose genomes consist of tightly packed protein-coding sequences. In eukaryotic cells, in contrast, such sequences are relatively rare, and the rest of the genome includes numerous transcript units representing non-coding RNAs (ncRNAs). During the past decade, it has become clear that such RNAs play a crucial role in gene expression and accordingly influence a multitude of cellular processes both in the normal state and during diseases. The numerous biological functions of ncRNAs are based primarily on their abilities to silence genes via pairing with a target mRNA and subsequently preventing its translation or facilitating degradation of the mRNA-ncRNA complex. Many other abilities of ncRNAs have been discovered as well. Our review is focused on the available kinetic models describing the mRNA, ncRNA and protein interplay. In particular, we systematically present the simplest models without kinetic feedbacks, models containing feedbacks and predicting bistability and oscillations in simple genetic networks, and models describing the effect of ncRNAs on complex genetic networks. Mathematically, the presentation is based primarily on temporal mean-field kinetic equations. The stochastic and spatio-temporal effects are also briefly discussed.

Numerical analysis of reflood simulation based on a mechanistic, best-estimate, approach by KREWET code

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

Numerical analysis for reflood simulation based on a mechanistic, best-estimate, approach by KREWET code

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

Geodynamo and mantle convection simulations on the Earth Simulator using the Yin-Yang grid

International Nuclear Information System (INIS)

Kageyama, Akira; Yoshida, Masaki

2005-01-01

We have developed finite difference codes based on the Yin-Yang grid for the geodynamo simulation and the mantle convection simulation. The Yin-Yang grid is a kind of spherical overset grid that is composed of two identical component grids. The intrinsic simplicity of the mesh configuration of the Yin-Yang grid enables us to develop highly optimized simulation codes on massively parallel supercomputers. The Yin-Yang geodynamo code has achieved 15.2 Tflops with 4096 processors on the Earth Simulator. This represents 46% of the theoretical peak performance. The Yin-Yang mantle code has enabled us to carry out mantle convection simulations in realistic regimes with a Rayleigh number of 10 7 including strongly temperature dependent viscosity with spatial contrast up to 10 6

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

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)

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

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.

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

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

Study and full simulation of ten different gases on sealed Multi-Wire Proportional Counter (MWPC) by using Garfield and Maxwell codes

International Nuclear Information System (INIS)

Shohani, M. Ebrahimi; Golgoun, S.M.; Aminipour, M.; Shabani, A.; Mazoochi, A.R.; Akbari, R. Maghsoudi; Mohammadzadeh, M.; Davarpanah, M.R.; Sardari, D.; Sadeghi, M.; Mofrad, F. Babapour; Jafari, A.

2016-01-01

In this research gas sealed Multi-Wire Proportional Counter (MWPC) including blades between anode wires and beta particles of "9"0Sr with 196 keV mean energy were considered. Ten different gases such as Noble gases mixtures with methane and several other pure gases were studied. In this type of detector, by using Garfield and Maxwell codes and for each of the gases, variation of different parameters such as first Townsend, electron attachment coefficients with variable electric field and their effects on pulse height or collected charge and in turn on Signal to Noise Ratio (SNR) were studied. Also the effect of anode voltage and its diameter and the pressure of gas on the pulse height were studied. Results show that Garfield and Maxwell codes can be used to study and improve the design of other gaseous detectors. - Highlights: • Simulation of different gases that is applicable for various gas detectors. • Two simulation codes were used and analyzed their results for beta particle. • Different detector parameters were studied (SNR, first Townsend, electron attachment coefficients, anode voltage and etc.). • The effect of blade in the detector were assessed. • The codes are useful for design and improvement of detector.

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.

Comparison benchmark between tokamak simulation code and TokSys for Chinese Fusion Engineering Test Reactor vertical displacement control design

International Nuclear Information System (INIS)

Qiu Qing-Lai; Xiao Bing-Jia; Guo Yong; Liu Lei; Wang Yue-Hang

2017-01-01

Vertical displacement event (VDE) is a big challenge to the existing tokamak equipment and that being designed. As a Chinese next-step tokamak, the Chinese Fusion Engineering Test Reactor (CFETR) has to pay attention to the VDE study with full-fledged numerical codes during its conceptual design. The tokamak simulation code (TSC) is a free boundary time-dependent axisymmetric tokamak simulation code developed in PPPL, which advances the MHD equations describing the evolution of the plasma in a rectangular domain. The electromagnetic interactions between the surrounding conductor circuits and the plasma are solved self-consistently. The TokSys code is a generic modeling and simulation environment developed in GA. Its RZIP model treats the plasma as a fixed spatial distribution of currents which couple with the surrounding conductors through circuit equations. Both codes have been individually used for the VDE study on many tokamak devices, such as JT-60U, EAST, NSTX, DIII-D, and ITER. Considering the model differences, benchmark work is needed to answer whether they reproduce each other’s results correctly. In this paper, the TSC and TokSys codes are used for analyzing the CFETR vertical instability passive and active controls design simultaneously. It is shown that with the same inputs, the results from these two codes conform with each other. (paper)

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

Numerical simulation of L.E.L. in Compton regime. Part II, GONDOLE, a three-dimensional code

International Nuclear Information System (INIS)

Deck, D.

1992-07-01

In the first part of this report, the BIWI two-dimensional numerical simulation code of L.E.L. in Compton regime has been described; the question was to simulate L.E.L. experiments in 'optical mode', that is to say for wavelengths of the order of one micron. The axisymmetric cylindrical geometry (r,z) of the BIWI code is adapted to these experiments. However, the increasingly frequent use of L.E.L. in the regime of microwaves requires the presence of a waveguide within the inverter, which breaks the cylindrical symmetry and forces us to adopt another geometry. On the other hand, the desire to take into account fields of inverters having a gradient in the direction transverse to the direction of propagation of the beam, and thus allowing various focalizations (quadrupole, parabolic, etc.), leads to work in Cartesian geometry. For these reasons (and for others that will appear later), the GONDOLE code has been written and is described in this note. The Gondole code is three-dimensional (x, y, z) and allows to simulate a large variety of L.E.L experiences. Then, all the inverter fields that the GONDOLE code takes into account are introduced. These fields are responsible for the existence of a current J(vector) perpendicular to the Z axis of propagation, and source of radiation. The dynamics of the electrons is then deduced, which derives directly from these fields, and it is shown to which equations of propagation of the laser wave each different J(vector) is coupling [fr

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.

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

Electromagnetic direct implicit PIC simulation

International Nuclear Information System (INIS)

Langdon, A.B.

1983-01-01

Interesting modelling of intense electron flow has been done with implicit particle-in-cell simulation codes. In this report, the direct implicit PIC simulation approach is applied to simulations that include full electromagnetic fields. The resulting algorithm offers advantages relative to moment implicit electromagnetic algorithms and may help in our quest for robust and simpler implicit codes

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].

Development of a space radiation Monte Carlo computer simulation based on the FLUKA and ROOT codes

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