Comparison of Radiation Transport Codes, HZETRN, HETC and FLUKA, Using the 1956 Webber SPE Spectrum
Heinbockel, John H.; Slaba, Tony C.; Blattnig, Steve R.; Tripathi, Ram K.; Townsend, Lawrence W.; Handler, Thomas; Gabriel, Tony A.; Pinsky, Lawrence S.; Reddell, Brandon; Clowdsley, Martha S.;
2009-01-01
Protection of astronauts and instrumentation from galactic cosmic rays (GCR) and solar particle events (SPE) in the harsh environment of space is of prime importance in the design of personal shielding, spacec raft, and mission planning. Early entry of radiation constraints into the design process enables optimal shielding strategies, but demands efficient and accurate tools that can be used by design engineers in every phase of an evolving space project. The radiation transport code , HZETRN, is an efficient tool for analyzing the shielding effectiveness of materials exposed to space radiation. In this paper, HZETRN is compared to the Monte Carlo codes HETC-HEDS and FLUKA, for a shield/target configuration comprised of a 20 g/sq cm Aluminum slab in front of a 30 g/cm^2 slab of water exposed to the February 1956 SPE, as mode led by the Webber spectrum. Neutron and proton fluence spectra, as well as dose and dose equivalent values, are compared at various depths in the water target. This study shows that there are many regions where HZETRN agrees with both HETC-HEDS and FLUKA for this shield/target configuration and the SPE environment. However, there are also regions where there are appreciable differences between the three computer c odes.
Pre-engineering Spaceflight Validation of Environmental Models and the 2005 HZETRN Simulation Code
Nealy, John E.; Cucinotta, Francis A.; Wilson, John W.; Badavi, Francis F.; Dachev, Ts. P.; Tomov, B. T.; Walker, Steven A.; DeAngelis, Giovanni; Blattnig, Steve R.; Atwell, William
2006-01-01
The HZETRN code has been identified by NASA for engineering design in the next phase of space exploration highlighting a return to the Moon in preparation for a Mars mission. In response, a new series of algorithms beginning with 2005 HZETRN, will be issued by correcting some prior limitations and improving control of propagated errors along with established code verification processes. Code validation processes will use new/improved low Earth orbit (LEO) environmental models with a recently improved International Space Station (ISS) shield model to validate computational models and procedures using measured data aboard ISS. These validated models will provide a basis for flight-testing the designs of future space vehicles and systems of the Constellation program in the LEO environment.
Comparison of Space Radiation Calculations from Deterministic and Monte Carlo Transport Codes
Adams, J. H.; Lin, Z. W.; Nasser, A. F.; Randeniya, S.; Tripathi, r. K.; Watts, J. W.; Yepes, P.
2010-01-01
The presentation outline includes motivation, radiation transport codes being considered, space radiation cases being considered, results for slab geometry, results from spherical geometry, and summary. ///////// main physics in radiation transport codes hzetrn uprop fluka geant4, slab geometry, spe, gcr,
NASA space radiation transport code development consortium
International Nuclear Information System (INIS)
Townsend, L. W.
2005-01-01
Recently, NASA established a consortium involving the Univ. of Tennessee (lead institution), the Univ. of Houston, Roanoke College and various government and national laboratories, to accelerate the development of a standard set of radiation transport computer codes for NASA human exploration applications. This effort involves further improvements of the Monte Carlo codes HETC and FLUKA and the deterministic code HZETRN, including developing nuclear reaction databases necessary to extend the Monte Carlo codes to carry out heavy ion transport, and extending HZETRN to three dimensions. The improved codes will be validated by comparing predictions with measured laboratory transport data, provided by an experimental measurements consortium, and measurements in the upper atmosphere on the balloon-borne Deep Space Test Bed (DSTB). In this paper, we present an overview of the consortium members and the current status and future plans of consortium efforts to meet the research goals and objectives of this extensive undertaking. (authors)
Gronoff, Guillaume; Norman, Ryan B.; Mertens, Christopher J.
2014-01-01
The ability to evaluate the cosmic ray environment at Mars is of interest for future manned exploration. To support exploration, tools must be developed to accurately access the radiation environment in both free space and on planetary surfaces. The primary tool NASA uses to quantify radiation exposure behind shielding materials is the space radiation transport code, HZETRN. In order to build confidence in HZETRN, code benchmarking against Monte Carlo radiation transport codes is often used. This work compares the dose calculations at Mars by HZETRN and the Geant4 application Planetocosmics. The dose at ground and the energy deposited in the atmosphere by galactic cosmic ray protons and alpha particles has been calculated for the Curiosity landing conditions. In addition, this work has considered Solar Energetic Particle events, allowing for the comparison of varying input radiation environments. The results for protons and alpha particles show very good agreement between HZETRN and Planetocosmics.
International Nuclear Information System (INIS)
Clancy, B.E.
1986-01-01
This chapter begins with a neutron transport equation which includes the one dimensional plane geometry problems, the one dimensional spherical geometry problems, and numerical solutions. The section on the ANISN code and its look-alikes covers problems which can be solved; eigenvalue problems; outer iteration loop; inner iteration loop; and finite difference solution procedures. The input and output data for ANISN is also discussed. Two dimensional problems such as the DOT code are given. Finally, an overview of the Monte-Carlo methods and codes are elaborated on
DEFF Research Database (Denmark)
Hansen, Jonas; Krigslund, Jeppe; Roetter, Daniel Enrique Lucani
2014-01-01
oblivious to the congestion control algorithms of the utilised transport layer protocol. Although our coding shim is indifferent towards the transport layer protocol, we focus on the performance of TCP when ran on top of our proposed coding mechanism due to its widespread use. The coding shim provides gains...
Electron transport code theoretical basis
International Nuclear Information System (INIS)
Dubi, A.; Horowitz, Y.S.
1978-04-01
This report mainly describes the physical and mathematical considerations involved in the treatment of the multiple collision processes. A brief description is given of the traditional methods used in electron transport via Monte Carlo, and a somewhat more detailed description, of the approach to be used in the presently developed code
In-facility transport code review
International Nuclear Information System (INIS)
Spore, J.W.; Boyack, B.E.; Bohl, W.R.
1996-07-01
The following computer codes were reviewed by the In-Facility Transport Working Group for application to the in-facility transport of radioactive aerosols, flammable gases, and/or toxic gases: (1) CONTAIN, (2) FIRAC, (3) GASFLOW, (4) KBERT, and (5) MELCOR. Based on the review criteria as described in this report and the versions of each code available at the time of the review, MELCOR is the best code for the analysis of in-facility transport when multidimensional effects are not significant. When multi-dimensional effects are significant, GASFLOW should be used
Reactive transport codes for subsurface environmental simulation
Steefel, C.I.; Appelo, C.A.J.; Arora, B.; Kalbacher, D.; Kolditz, O.; Lagneau, V.; Lichtner, P.C.; Mayer, K.U.; Meeussen, J.C.L.; Molins, S.; Moulton, D.; Shao, D.; Simunek, J.; Spycher, N.; Yabusaki, S.B.; Yeh, G.T.
2015-01-01
A general description of the mathematical and numerical formulations used in modern numerical reactive transport codes relevant for subsurface environmental simulations is presented. The formulations are followed by short descriptions of commonly used and available subsurface simulators that
Parallel processing Monte Carlo radiation transport codes
International Nuclear Information System (INIS)
McKinney, G.W.
1994-01-01
Issues related to distributed-memory multiprocessing as applied to Monte Carlo radiation transport are discussed. Measurements of communication overhead are presented for the radiation transport code MCNP which employs the communication software package PVM, and average efficiency curves are provided for a homogeneous virtual machine
Computer codes in particle transport physics
International Nuclear Information System (INIS)
Pesic, M.
2004-01-01
Simulation of transport and interaction of various particles in complex media and wide energy range (from 1 MeV up to 1 TeV) is very complicated problem that requires valid model of a real process in nature and appropriate solving tool - computer code and data library. A brief overview of computer codes based on Monte Carlo techniques for simulation of transport and interaction of hadrons and ions in wide energy range in three dimensional (3D) geometry is shown. Firstly, a short attention is paid to underline the approach to the solution of the problem - process in nature - by selection of the appropriate 3D model and corresponding tools - computer codes and cross sections data libraries. Process of data collection and evaluation from experimental measurements and theoretical approach to establishing reliable libraries of evaluated cross sections data is Ion g, difficult and not straightforward activity. For this reason, world reference data centers and specialized ones are acknowledged, together with the currently available, state of art evaluated nuclear data libraries, as the ENDF/B-VI, JEF, JENDL, CENDL, BROND, etc. Codes for experimental and theoretical data evaluations (e.g., SAMMY and GNASH) together with the codes for data processing (e.g., NJOY, PREPRO and GRUCON) are briefly described. Examples of data evaluation and data processing to generate computer usable data libraries are shown. Among numerous and various computer codes developed in transport physics of particles, the most general ones are described only: MCNPX, FLUKA and SHIELD. A short overview of basic application of these codes, physical models implemented with their limitations, energy ranges of particles and types of interactions, is given. General information about the codes covers also programming language, operation system, calculation speed and the code availability. An example of increasing computation speed of running MCNPX code using a MPI cluster compared to the code sequential option
FLUKA: A Multi-Particle Transport Code
Energy Technology Data Exchange (ETDEWEB)
Ferrari, A.; Sala, P.R.; /CERN /INFN, Milan; Fasso, A.; /SLAC; Ranft, J.; /Siegen U.
2005-12-14
This report describes the 2005 version of the Fluka particle transport code. The first part introduces the basic notions, describes the modular structure of the system, and contains an installation and beginner's guide. The second part complements this initial information with details about the various components of Fluka and how to use them. It concludes with a detailed history and bibliography.
Utilization of CAM, CAF, MAX, and FAX for space radiation analyses using HZETRN
Slaba, T. C.; Qualls, G. D.; Clowdsley, M. S.; Blattnig, S. R.; Walker, S. A.; Simonsen, L. C.
2010-04-01
To estimate astronaut health risk due to space radiation, one must have the ability to calculate various exposure-related quantities that are averaged over specific organs and tissue types. Such calculations require computational models of the ambient space radiation environment, particle transport, nuclear and atomic physics, and the human body. While significant efforts have been made to verify, validate, and quantify the uncertainties associated with many of these models and tools, relatively little work has focused on the uncertainties associated with the representation and utilization of the human phantoms. In this study, we first examine the anatomical properties of the Computerized Anatomical Man (CAM), Computerized Anatomical Female (CAF), Male Adult voXel (MAX), and Female Adult voXel (FAX) models by comparing the masses of various model tissues used to calculate effective dose to the reference values specified by the International Commission on Radiological Protection (ICRP). The MAX and FAX tissue masses are found to be in good agreement with the reference data, while major discrepancies are found between the CAM and CAF tissue masses and the reference data for almost all of the effective dose tissues. We next examine the distribution of target points used with the deterministic transport code HZETRN (High charge (Z) and Energy TRaNsport) to compute mass averaged exposure quantities. A numerical algorithm is presented and used to generate multiple point distributions of varying fidelity for many of the effective dose tissues identified in CAM, CAF, MAX, and FAX. The point distributions are used to compute mass averaged dose equivalent values under both a galactic cosmic ray (GCR) and solar particle event (SPE) environment impinging isotropically on three spherical aluminum shells with areal densities of 0.4 g/cm2, 2.0 g/cm2, and 10.0 g/cm2. The dose equivalent values are examined to identify a recommended set of target points for each of the tissues and
Morse Monte Carlo Radiation Transport Code System
Energy Technology Data Exchange (ETDEWEB)
Emmett, M.B.
1975-02-01
The report contains sections containing descriptions of the MORSE and PICTURE codes, input descriptions, sample problems, deviations of the physical equations and explanations of the various error messages. The MORSE code is a multipurpose neutron and gamma-ray transport Monte Carlo code. Time dependence for both shielding and criticality problems is provided. General three-dimensional geometry may be used with an albedo option available at any material surface. The PICTURE code provide aid in preparing correct input data for the combinatorial geometry package CG. It provides a printed view of arbitrary two-dimensional slices through the geometry. By inspecting these pictures one may determine if the geometry specified by the input cards is indeed the desired geometry. 23 refs. (WRF)
Colloid transport code-nuclear user's manual
International Nuclear Information System (INIS)
Jain, R.
1992-01-01
This report describes the CTCN computer code, designed to solve the equations of transient colloidal transport of radionuclides in porous and fractured media. This Fortran 77 package solves systems of coupled nonlinear differential equations with a wide range of boundary conditions. The package uses the Method of Lines technique with a special section which forms finite-difference discretizations in up to four spatial dimensions to automatically convert the system into a set of ordinary differential equations. The CTCN code then solves these equations using a robust, efficient ODE solver. Thus CTCN can be used to solve population balance equations along with the usual transport equations to model colloid transport processes or as a general problem solver to treat up to four-dimensional differential systems
The MC21 Monte Carlo Transport Code
International Nuclear Information System (INIS)
Sutton TM; Donovan TJ; Trumbull TH; Dobreff PS; Caro E; Griesheimer DP; Tyburski LJ; Carpenter DC; Joo H
2007-01-01
MC21 is a new Monte Carlo neutron and photon transport code currently under joint development at the Knolls Atomic Power Laboratory and the Bettis Atomic Power Laboratory. MC21 is the Monte Carlo transport kernel of the broader Common Monte Carlo Design Tool (CMCDT), which is also currently under development. The vision for CMCDT is to provide an automated, computer-aided modeling and post-processing environment integrated with a Monte Carlo solver that is optimized for reactor analysis. CMCDT represents a strategy to push the Monte Carlo method beyond its traditional role as a benchmarking tool or ''tool of last resort'' and into a dominant design role. This paper describes various aspects of the code, including the neutron physics and nuclear data treatments, the geometry representation, and the tally and depletion capabilities
Hydrogen recycle modeling in transport codes
International Nuclear Information System (INIS)
Howe, H.C.
1979-01-01
The hydrogen recycling models now used in Tokamak transport codes are reviewed and the method by which realistic recycling models are being added is discussed. Present models use arbitrary recycle coefficients and therefore do not model the actual recycling processes at the wall. A model for the hydrogen concentration in the wall serves two purposes: (1) it allows a better understanding of the density behavior in present gas puff, pellet, and neutral beam heating experiments; and (2) it allows one to extrapolate to long pulse devices such as EBT, ISX-C and reactors where the walls are observed or expected to saturate. Several wall models are presently being studied for inclusion in transport codes
Utilization of MAX and FAX human phantoms for space radiation exposure calculations using HZETRN
Qualls, Garry; Slaba, Tony; Clowdsley, Martha; Blattnig, Steve; Walker, Steven; Simonsen, Lisa
To estimate astronaut health risk due to space radiation, one must have the ability to calculate, for known radiation environments external to the body, particle spectra, LET spectra, dose, dose equivalent, or gray equivalent that are averaged over specific organs or tissue types. This may be accomplished using radiation transport software and computational human body tissue models. Historically, NASA scientists have used the HZETRN software to calculate radiation transport through both vehicle shielding materials and body tissue. The Computerized Anatomical Man (CAM) and the Computerized Anatomical Female (CAF) body models, combined with the CAMERA software, have been used for body tissue self-shielding calculations. The CAM and CAF, which were developed in 1973 and 1992, respectively, model the 50th percentile U.S. Air Force male and female and are constructed using individual quadric surfaces that combine to form thousands of solid regions that represent specific tissues and structures within the body. In order to transport an external radiation environment to a point within one of the body models using HZETRN, a directional distribution of the tissues surrounding that point is needed. The CAMERA software is used to "ray trace" the CAM and CAF models, providing the thickness of each tissue type traversed along each of a large number of rays originating at a dose point. More recently, R. Kramer of the Departmento de Energia Nuclear, Universidade Federal de Pernambuco in Brazil and his co-workers developed the Male Adult voXel (MAX) model and the Female Adult voXel (FAX). These voxel-based body models were developed using segmented Computed Tomography (CT) scans of adult cadavers, and the quantities and distributions of various body tissues have been adjusted to match those specified in the International Commission on Radiological Protection (ICRP) reference adult male and female. A new set of tools has been developed to facilitate space radiation exposure
Creation and utilization of a World Wide Web based space radiation effects code: SIREST
Singleterry, R. C. Jr; Wilson, J. W.; Shinn, J. L.; Tripathi, R. K.; Thibeault, S. A.; Noor, A. K.; Cucinotta, F. A.; Badavi, F. F.; Chang, C. K.; Qualls, G. D.;
2001-01-01
In order for humans and electronics to fully and safely operate in the space environment, codes like HZETRN (High Charge and Energy Transport) must be included in any designer's toolbox for design evaluation with respect to radiation damage. Currently, spacecraft designers do not have easy access to accurate radiation codes like HZETRN to evaluate their design for radiation effects on humans and electronics. Today, the World Wide Web is sophisticated enough to support the entire HZETRN code and all of the associated pre and post processing tools. This package is called SIREST (Space Ionizing Radiation Effects and Shielding Tools). There are many advantages to SIREST. The most important advantage is the instant update capability of the web. Another major advantage is the modularity that the web imposes on the code. Right now, the major disadvantage of SIREST will be its modularity inside the designer's system. This mostly comes from the fact that a consistent interface between the designer and the computer system to evaluate the design is incomplete. This, however, is to be solved in the Intelligent Synthesis Environment (ISE) program currently being funded by NASA.
Current status of high energy nucleon-meson transport code
Energy Technology Data Exchange (ETDEWEB)
Takada, Hiroshi; Sasa, Toshinobu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1998-03-01
Current status of design code of accelerator (NMTC/JAERI code), outline of physical model and evaluation of accuracy of code were reported. To evaluate the nuclear performance of accelerator and strong spallation neutron origin, the nuclear reaction between high energy proton and target nuclide and behaviors of various produced particles are necessary. The nuclear design of spallation neutron system used a calculation code system connected the high energy nucleon{center_dot}meson transport code and the neutron{center_dot}photon transport code. NMTC/JAERI is described by the particle evaporation process under consideration of competition reaction of intranuclear cascade and fission process. Particle transport calculation was carried out for proton, neutron, {pi}- and {mu}-meson. To verify and improve accuracy of high energy nucleon-meson transport code, data of spallation and spallation neutron fragment by the integral experiment were collected. (S.Y.)
RADTRAN: a computer code to analyze transportation of radioactive material
International Nuclear Information System (INIS)
Taylor, J.M.; Daniel, S.L.
1977-04-01
A computer code is presented which predicts the environmental impact of any specific scheme of radioactive material transportation. Results are presented in terms of annual latent cancer fatalities and annual early fatility probability resulting from exposure, during normal transportation or transport accidents. The code is developed in a generalized format to permit wide application including normal transportation analysis; consideration of alternatives; and detailed consideration of specific sectors of industry
Recent developments in the Los Alamos radiation transport code system
International Nuclear Information System (INIS)
Forster, R.A.; Parsons, K.
1997-01-01
A brief progress report on updates to the Los Alamos Radiation Transport Code System (LARTCS) for solving criticality and fixed-source problems is provided. LARTCS integrates the Diffusion Accelerated Neutral Transport (DANT) discrete ordinates codes with the Monte Carlo N-Particle (MCNP) code. The LARCTS code is being developed with a graphical user interface for problem setup and analysis. Progress in the DANT system for criticality applications include a two-dimensional module which can be linked to a mesh-generation code and a faster iteration scheme. Updates to MCNP Version 4A allow statistical checks of calculated Monte Carlo results
Interfacial and Wall Transport Models for SPACE-CAP Code
Energy Technology Data Exchange (ETDEWEB)
Hong, Soon Joon; Choo, Yeon Joon; Han, Tae Young; Hwang, Su Hyun; Lee, Byung Chul [FNC Tech., Seoul (Korea, Republic of); Choi, Hoon; Ha, Sang Jun [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)
2009-10-15
The development project for the domestic design code was launched to be used for the safety and performance analysis of pressurized light water reactors. And CAP (Containment Analysis Package) code has been also developed for the containment safety and performance analysis side by side with SPACE. The CAP code treats three fields (gas, continuous liquid, and dispersed drop) for the assessment of containment specific phenomena, and is featured by its multidimensional assessment capabilities. Thermal hydraulics solver was already developed and now under testing of its stability and soundness. As a next step, interfacial and wall transport models was setup. In order to develop the best model and correlation package for the CAP code, various models currently used in major containment analysis codes, which are GOTHIC, CONTAIN2.0, and CONTEMPT-LT, have been reviewed. The origins of the selected models used in these codes have also been examined to find out if the models have not conflict with a proprietary right. In addition, a literature survey of the recent studies has been performed in order to incorporate the better models for the CAP code. The models and correlations of SPACE were also reviewed. CAP models and correlations are composed of interfacial heat/mass, and momentum transport models, and wall heat/mass, and momentum transport models. This paper discusses on those transport models in the CAP code.
Benchmarking NNWSI flow and transport codes: COVE 1 results
International Nuclear Information System (INIS)
Hayden, N.K.
1985-06-01
The code verification (COVE) activity of the Nevada Nuclear Waste Storage Investigations (NNWSI) Project is the first step in certification of flow and transport codes used for NNWSI performance assessments of a geologic repository for disposing of high-level radioactive wastes. The goals of the COVE activity are (1) to demonstrate and compare the numerical accuracy and sensitivity of certain codes, (2) to identify and resolve problems in running typical NNWSI performance assessment calculations, and (3) to evaluate computer requirements for running the codes. This report describes the work done for COVE 1, the first step in benchmarking some of the codes. Isothermal calculations for the COVE 1 benchmarking have been completed using the hydrologic flow codes SAGUARO, TRUST, and GWVIP; the radionuclide transport codes FEMTRAN and TRUMP; and the coupled flow and transport code TRACR3D. This report presents the results of three cases of the benchmarking problem solved for COVE 1, a comparison of the results, questions raised regarding sensitivities to modeling techniques, and conclusions drawn regarding the status and numerical sensitivities of the codes. 30 refs
Bounce-averaged Fokker-Planck code for stellarator transport
International Nuclear Information System (INIS)
Mynick, H.E.; Hitchon, W.N.G.
1985-07-01
A computer code for solving the bounce-averaged Fokker-Planck equation appropriate to stellarator transport has been developed, and its first applications made. The code is much faster than the bounce-averaged Monte-Carlo codes, which up to now have provided the most efficient numerical means for studying stellarator transport. Moreover, because the connection to analytic kinetic theory of the Fokker-Planck approach is more direct than for the Monte-Carlo approach, a comparison of theory and numerical experiment is now possible at a considerably more detailed level than previously
Recent advances in neutral particle transport methods and codes
International Nuclear Information System (INIS)
Azmy, Y.Y.
1996-01-01
An overview of ORNL's three-dimensional neutral particle transport code, TORT, is presented. Special features of the code that make it invaluable for large applications are summarized for the prospective user. Advanced capabilities currently under development and installation in the production release of TORT are discussed; they include: multitasking on Cray platforms running the UNICOS operating system; Adjacent cell Preconditioning acceleration scheme; and graphics codes for displaying computed quantities such as the flux. Further developments for TORT and its companion codes to enhance its present capabilities, as well as expand its range of applications are disucssed. Speculation on the next generation of neutron particle transport codes at ORNL, especially regarding unstructured grids and high order spatial approximations, are also mentioned
Acceleration of a Monte Carlo radiation transport code
International Nuclear Information System (INIS)
Hochstedler, R.D.; Smith, L.M.
1996-01-01
Execution time for the Integrated TIGER Series (ITS) Monte Carlo radiation transport code has been reduced by careful re-coding of computationally intensive subroutines. Three test cases for the TIGER (1-D slab geometry), CYLTRAN (2-D cylindrical geometry), and ACCEPT (3-D arbitrary geometry) codes were identified and used to benchmark and profile program execution. Based upon these results, sixteen top time-consuming subroutines were examined and nine of them modified to accelerate computations with equivalent numerical output to the original. The results obtained via this study indicate that speedup factors of 1.90 for the TIGER code, 1.67 for the CYLTRAN code, and 1.11 for the ACCEPT code are achievable. copyright 1996 American Institute of Physics
ARTEMIS: a 3D transport code for shielding calculations
Energy Technology Data Exchange (ETDEWEB)
Varin, E.; Samba, G. [Commissariat a l' Energie Atomique, Bruyeres-Le-Chatels (France); Roy, R. [Ecole Polytechnique de Montreal, Montreal, Quebec (Canada)
2002-07-01
In radiation transport problems, as shielding applications, the solution of the Boltzmann transport equation is usually obtained by the discrete ordinates deterministic method. An alternative methodology has been developed in three dimensions into the code ARTEMIS. A Spherical Harmonics expansion of the angular flux has been chosen to guaranty solutions free of ray-effects. A least squares approach is applied over the linear transport equation; this approach leads to well-defined symmetric positive definite systems which allows the use of finite element spatial discretization. This paper presents the basic derivation of the discrete equations and provides examples on the use of this technique to solve different transport problems. (author)
Multidimensional electron-photon transport with standard discrete ordinates codes
International Nuclear Information System (INIS)
Drumm, C.R.
1995-01-01
A method is described for generating electron cross sections that are compatible with standard discrete ordinates codes without modification. There are many advantages of using an established discrete ordinates solver, e.g. immediately available adjoint capability. Coupled electron-photon transport capability is needed for many applications, including the modeling of the response of electronics components to space and man-made radiation environments. The cross sections have been successfully used in the DORT, TWODANT and TORT discrete ordinates codes. The cross sections are shown to provide accurate and efficient solutions to certain multidimensional electronphoton transport problems
RADTRAN 5: A computer code for transportation risk analysis
International Nuclear Information System (INIS)
Neuhauser, K.S.; Kanipe, F.L.
1991-01-01
RADTRAN 5 is a computer code developed at Sandia National Laboratories (SNL) in Albuquerque, NM, to estimate radiological and nonradiological risks of radioactive materials transportation. RADTRAN 5 is written in ANSI Standard FORTRAN 77 and contains significant advances in the methodology for route-specific analysis first developed by SNL for RADTRAN 4 (Neuhauser and Kanipe, 1992). Like the previous RADTRAN codes, RADTRAN 5 contains two major modules for incident-free and accident risk amlysis, respectively. All commercially important transportation modes may be analyzed with RADTRAN 5: highway by combination truck; highway by light-duty vehicle; rail; barge; ocean-going ship; cargo air; and passenger air
Planning guide for validation of fission product transport codes
International Nuclear Information System (INIS)
Jensen, D.D.; Haire, M.J.; Baldassare, J.E.; Hanson, D.L.
1975-01-01
The program for validating fission product transport codes utilized in the design of the high-temperature gas-cooled reactor (HTGR) is described herein. The importance of fission product code verification is discussed as it relates to achieving a competitive reactor system that fully complies with federal regulations. A brief description of the RAD, PAD, and FIPER codes and their validation status is given. Individual validation tests are described in detail, including test conditions and measurements to be evaluated, and accompanying test schedules. Also included are validation schedules for each code inclusive through fiscal year 1978. Codes will be appropriately validated and utilized for fission product predictions for the Delmarva Final Safety Analysis Report (FSAR) due for release in early 1978. (U.S.)
DANTSYS: A diffusion accelerated neutral particle transport code system
Energy Technology Data Exchange (ETDEWEB)
Alcouffe, R.E.; Baker, R.S.; Brinkley, F.W.; Marr, D.R.; O`Dell, R.D.; Walters, W.F.
1995-06-01
The DANTSYS code package includes the following transport codes: ONEDANT, TWODANT, TWODANT/GQ, TWOHEX, and THREEDANT. The DANTSYS code package is a modular computer program package designed to solve the time-independent, multigroup discrete ordinates form of the boltzmann transport equation in several different geometries. The modular construction of the package separates the input processing, the transport equation solving, and the post processing (or edit) functions into distinct code modules: the Input Module, one or more Solver Modules, and the Edit Module, respectively. The Input and Edit Modules are very general in nature and are common to all the Solver Modules. The ONEDANT Solver Module contains a one-dimensional (slab, cylinder, and sphere), time-independent transport equation solver using the standard diamond-differencing method for space/angle discretization. Also included in the package are solver Modules named TWODANT, TWODANT/GQ, THREEDANT, and TWOHEX. The TWODANT Solver Module solves the time-independent two-dimensional transport equation using the diamond-differencing method for space/angle discretization. The authors have also introduced an adaptive weighted diamond differencing (AWDD) method for the spatial and angular discretization into TWODANT as an option. The TWOHEX Solver Module solves the time-independent two-dimensional transport equation on an equilateral triangle spatial mesh. The THREEDANT Solver Module solves the time independent, three-dimensional transport equation for XYZ and RZ{Theta} symmetries using both diamond differencing with set-to-zero fixup and the AWDD method. The TWODANT/GQ Solver Module solves the 2-D transport equation in XY and RZ symmetries using a spatial mesh of arbitrary quadrilaterals. The spatial differencing method is based upon the diamond differencing method with set-to-zero fixup with changes to accommodate the generalized spatial meshing.
DANTSYS: A diffusion accelerated neutral particle transport code system
International Nuclear Information System (INIS)
Alcouffe, R.E.; Baker, R.S.; Brinkley, F.W.; Marr, D.R.; O'Dell, R.D.; Walters, W.F.
1995-06-01
The DANTSYS code package includes the following transport codes: ONEDANT, TWODANT, TWODANT/GQ, TWOHEX, and THREEDANT. The DANTSYS code package is a modular computer program package designed to solve the time-independent, multigroup discrete ordinates form of the boltzmann transport equation in several different geometries. The modular construction of the package separates the input processing, the transport equation solving, and the post processing (or edit) functions into distinct code modules: the Input Module, one or more Solver Modules, and the Edit Module, respectively. The Input and Edit Modules are very general in nature and are common to all the Solver Modules. The ONEDANT Solver Module contains a one-dimensional (slab, cylinder, and sphere), time-independent transport equation solver using the standard diamond-differencing method for space/angle discretization. Also included in the package are solver Modules named TWODANT, TWODANT/GQ, THREEDANT, and TWOHEX. The TWODANT Solver Module solves the time-independent two-dimensional transport equation using the diamond-differencing method for space/angle discretization. The authors have also introduced an adaptive weighted diamond differencing (AWDD) method for the spatial and angular discretization into TWODANT as an option. The TWOHEX Solver Module solves the time-independent two-dimensional transport equation on an equilateral triangle spatial mesh. The THREEDANT Solver Module solves the time independent, three-dimensional transport equation for XYZ and RZΘ symmetries using both diamond differencing with set-to-zero fixup and the AWDD method. The TWODANT/GQ Solver Module solves the 2-D transport equation in XY and RZ symmetries using a spatial mesh of arbitrary quadrilaterals. The spatial differencing method is based upon the diamond differencing method with set-to-zero fixup with changes to accommodate the generalized spatial meshing
User's manual for the Oak Ridge Tokamak Transport Code
International Nuclear Information System (INIS)
Munro, J.K.; Hogan, J.T.; Howe, H.C.; Arnurius, D.E.
1977-02-01
A one-dimensional tokamak transport code is described which simulates a plasma discharge using a fluid model which includes power balances for electrons and ions, conservation of mass, and Maxwell's equations. The modular structure of the code allows a user to add models of various physical processes which can modify the discharge behavior. Such physical processes treated in the version of the code described here include effects of plasma transport, neutral gas transport, impurity diffusion, and neutral beam injection. Each process can be modeled by a parameterized analytic formula or at least one detailed numerical calculation. The program logic of each module is presented, followed by detailed descriptions of each subroutine used by the module. The physics underlying the models is only briefly summarized. The transport code was written in IBM FORTRAN-IV and implemented on IBM 360/370 series computers at the Oak Ridge National Laboratory and on the CDC 7600 computers of the Magnetic Fusion Energy (MFE) Computing Center of the Lawrence Livermore Laboratory. A listing of the current reference version is provided on accompanying microfiche
Simple one-dimensional transport code for magnetized target fusion
International Nuclear Information System (INIS)
Stefano Migluiolo
1999-01-01
A one-dimensional (in space) time-dependent simulation code is development to study the transport of energy and particles in a field reversed configuration (FRC) plasma that is undergoing radial contraction. This contraction is due to an imploding metallic liner, which is treated through a boundary condition
Transport code and nuclear data in intermediate energy region
Energy Technology Data Exchange (ETDEWEB)
Hasegawa, Akira; Odama, Naomitsu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Maekawa, F.; Ueki, K.; Kosaka, K.; Oyama, Y.
1998-11-01
We briefly reviewed the problems of intermediate energy nuclear data file and transport codes in connection with processing of the data. This is a summary of our group in the task force on JENDL High Energy File Integral Evaluation (JHEFIE). In this article we stress the necessity of the production of intermediate evaluated nuclear data file up to 3 GeV for the application of accelerator driven transmutation (ADT) system. And also we state the necessity of having our own transport code system to calculate the radiation fields using these evaluated files from the strategic points of view to keep our development of the ADT technology completely free from other conditions outside of our own such as imported codes and data with poor maintenance or unknown accuracy. (author)
Development of BERMUDA: a radiation transport code system, 1
International Nuclear Information System (INIS)
Suzuki, Tomoo; Hasegawa, Akira; Tanaka, Shun-ichi; Nakashima, Hiroshi
1992-05-01
A radiation transport code system BERMUDA has been developed for one-, two- and three-dimensional geometries. The time-independent transport equation is numerically solved using a direct integration method in a multigroup model, to obtain spatial, angular and energy distributions of neutron, gamma rays or adjoint neutron flux. As to group constants, a library with an any structure of energy groups is capable to be produced from a data base JSSTDL, or by a processing code PROF-GROUCH-G/B, selecting objective nuclear data through a retrieval system EDFSRS. Validity of the present code system has been tested by analyzing the shielding benchmark experiments. The test has shown that accurate results are obtainable with this system especially in deep penetration calculation. Described are the devised calculation method and the results of validity tests. Input data specification, job control languages and output data are also described as a user's manual for the following four neutron transport codes: BERMUDA-1DN : sphere, slab(S 20 ), BERMUDA-2DN : cylinder (S 8 ), BERMUDA-2DN-S16 : cylinder (S 16 ), and BERMUDA-3DN : rectangular parallelpiped (S 8 ). (J.P.N.)
The Initial Atmospheric Transport (IAT) Code: Description and Validation
Energy Technology Data Exchange (ETDEWEB)
Morrow, Charles W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bartel, Timothy James [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
2015-10-01
The Initial Atmospheric Transport (IAT) computer code was developed at Sandia National Laboratories as part of their nuclear launch accident consequences analysis suite of computer codes. The purpose of IAT is to predict the initial puff/plume rise resulting from either a solid rocket propellant or liquid rocket fuel fire. The code generates initial conditions for subsequent atmospheric transport calculations. The Initial Atmospheric Transfer (IAT) code has been compared to two data sets which are appropriate to the design space of space launch accident analyses. The primary model uncertainties are the entrainment coefficients for the extended Taylor model. The Titan 34D accident (1986) was used to calibrate these entrainment settings for a prototypic liquid propellant accident while the recent Johns Hopkins University Applied Physics Laboratory (JHU/APL, or simply APL) large propellant block tests (2012) were used to calibrate the entrainment settings for prototypic solid propellant accidents. North American Meteorology (NAM )formatted weather data profiles are used by IAT to determine the local buoyancy force balance. The IAT comparisons for the APL solid propellant tests illustrate the sensitivity of the plume elevation to the weather profiles; that is, the weather profile is a dominant factor in determining the plume elevation. The IAT code performed remarkably well and is considered validated for neutral weather conditions.
Regional Atmospheric Transport Code for Hanford Emission Tracking (RATCHET)
International Nuclear Information System (INIS)
Ramsdell, J.V. Jr.; Simonen, C.A.; Burk, K.W.
1994-02-01
The purpose of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate radiation doses that individuals may have received from operations at the Hanford Site since 1944. This report deals specifically with the atmospheric transport model, Regional Atmospheric Transport Code for Hanford Emission Tracking (RATCHET). RATCHET is a major rework of the MESOILT2 model used in the first phase of the HEDR Project; only the bookkeeping framework escaped major changes. Changes to the code include (1) significant changes in the representation of atmospheric processes and (2) incorporation of Monte Carlo methods for representing uncertainty in input data, model parameters, and coefficients. To a large extent, the revisions to the model are based on recommendations of a peer working group that met in March 1991. Technical bases for other portions of the atmospheric transport model are addressed in two other documents. This report has three major sections: a description of the model, a user's guide, and a programmer's guide. These sections discuss RATCHET from three different perspectives. The first provides a technical description of the code with emphasis on details such as the representation of the model domain, the data required by the model, and the equations used to make the model calculations. The technical description is followed by a user's guide to the model with emphasis on running the code. The user's guide contains information about the model input and output. The third section is a programmer's guide to the code. It discusses the hardware and software required to run the code. The programmer's guide also discusses program structure and each of the program elements
3D unstructured-mesh radiation transport codes
International Nuclear Information System (INIS)
Morel, J.
1997-01-01
Three unstructured-mesh radiation transport codes are currently being developed at Los Alamos National Laboratory. The first code is ATTILA, which uses an unstructured tetrahedral mesh in conjunction with standard Sn (discrete-ordinates) angular discretization, standard multigroup energy discretization, and linear-discontinuous spatial differencing. ATTILA solves the standard first-order form of the transport equation using source iteration in conjunction with diffusion-synthetic acceleration of the within-group source iterations. DANTE is designed to run primarily on workstations. The second code is DANTE, which uses a hybrid finite-element mesh consisting of arbitrary combinations of hexahedra, wedges, pyramids, and tetrahedra. DANTE solves several second-order self-adjoint forms of the transport equation including the even-parity equation, the odd-parity equation, and a new equation called the self-adjoint angular flux equation. DANTE also offers three angular discretization options: $S n$ (discrete-ordinates), $P n$ (spherical harmonics), and $SP n$ (simplified spherical harmonics). DANTE is designed to run primarily on massively parallel message-passing machines, such as the ASCI-Blue machines at LANL and LLNL. The third code is PERICLES, which uses the same hybrid finite-element mesh as DANTE, but solves the standard first-order form of the transport equation rather than a second-order self-adjoint form. DANTE uses a standard $S n$ discretization in angle in conjunction with trilinear-discontinuous spatial differencing, and diffusion-synthetic acceleration of the within-group source iterations. PERICLES was initially designed to run on workstations, but a version for massively parallel message-passing machines will be built. The three codes will be described in detail and computational results will be presented
Development of general-purpose particle and heavy ion transport monte carlo code
International Nuclear Information System (INIS)
Iwase, Hiroshi; Nakamura, Takashi; Niita, Koji
2002-01-01
The high-energy particle transport code NMTC/JAM, which has been developed at JAERI, was improved for the high-energy heavy ion transport calculation by incorporating the JQMD code, the SPAR code and the Shen formula. The new NMTC/JAM named PHITS (Particle and Heavy-Ion Transport code System) is the first general-purpose heavy ion transport Monte Carlo code over the incident energies from several MeV/nucleon to several GeV/nucleon. (author)
Parallelization of a Monte Carlo particle transport simulation code
Hadjidoukas, P.; Bousis, C.; Emfietzoglou, D.
2010-05-01
We have developed a high performance version of the Monte Carlo particle transport simulation code MC4. The original application code, developed in Visual Basic for Applications (VBA) for Microsoft Excel, was first rewritten in the C programming language for improving code portability. Several pseudo-random number generators have been also integrated and studied. The new MC4 version was then parallelized for shared and distributed-memory multiprocessor systems using the Message Passing Interface. Two parallel pseudo-random number generator libraries (SPRNG and DCMT) have been seamlessly integrated. The performance speedup of parallel MC4 has been studied on a variety of parallel computing architectures including an Intel Xeon server with 4 dual-core processors, a Sun cluster consisting of 16 nodes of 2 dual-core AMD Opteron processors and a 200 dual-processor HP cluster. For large problem size, which is limited only by the physical memory of the multiprocessor server, the speedup results are almost linear on all systems. We have validated the parallel implementation against the serial VBA and C implementations using the same random number generator. Our experimental results on the transport and energy loss of electrons in a water medium show that the serial and parallel codes are equivalent in accuracy. The present improvements allow for studying of higher particle energies with the use of more accurate physical models, and improve statistics as more particles tracks can be simulated in low response time.
The RADionuclide Transport, Removal, and Dose (RADTRAD) code
International Nuclear Information System (INIS)
Miller, L.A.; Chanin, D.I.; Lee, J.
1993-01-01
The RADionuclide Transport, Removal, And Dose (RADTRAD) code is designed for US Nuclear Regulatory Commission (USNRC) use to calculate the radiological consequences to the offsite population and to control room operators following a design-basis accident at Light Water Reactor (LWR) power plants. This code utilizes updated reactor accident source terms published in draft NUREG-1465, ''Accident Source Terms for Light-Water Nuclear Power Plants.'' The code will track the transport of radionuclides as they are released from the reactor pressure vessel, travel through the primary containment and other buildings, and are released to the environment. As the radioactive material is transported through the primary containment and other buildings, credit for several removal mechanisms may be taken including sprays, suppression pools, overlying pools, filters, and natural deposition. Simple models are available for these different removal mechanisms that use, as input, information about the conditions in the plant and predict either a removal coefficient (λ) or decontamination factor. The user may elect to use these models or input a single value for a removal coefficient or decontamination factor
High energy particle transport code NMTC/JAM
International Nuclear Information System (INIS)
Niita, Koji; Meigo, Shin-ichiro; Takada, Hiroshi; Ikeda, Yujiro
2001-03-01
We have developed a high energy particle transport code NMTC/JAM, which is an upgraded version of NMTC/JAERI97. The applicable energy range of NMTC/JAM is extended in principle up to 200 GeV for nucleons and mesons by introducing the high energy nuclear reaction code JAM for the intra-nuclear cascade part. For the evaporation and fission process, we have also implemented a new model, GEM, by which the light nucleus production from the excited residual nucleus can be described. According to the extension of the applicable energy, we have upgraded the nucleon-nucleus non-elastic, elastic and differential elastic cross section data by employing new systematics. In addition, the particle transport in a magnetic field has been implemented for the beam transport calculations. In this upgrade, some new tally functions are added and the format of input of data has been improved very much in a user friendly manner. Due to the implementation of these new calculation functions and utilities, consequently, NMTC/JAM enables us to carry out reliable neutronics study of a large scale target system with complex geometry more accurately and easily than before. This report serves as a user manual of the code. (author)
TOPIC: a debugging code for torus geometry input data of Monte Carlo transport code
International Nuclear Information System (INIS)
Iida, Hiromasa; Kawasaki, Hiromitsu.
1979-06-01
TOPIC has been developed for debugging geometry input data of the Monte Carlo transport code. the code has the following features: (1) It debugs the geometry input data of not only MORSE-GG but also MORSE-I capable of treating torus geometry. (2) Its calculation results are shown in figures drawn by Plotter or COM, and the regions not defined or doubly defined are easily detected. (3) It finds a multitude of input data errors in a single run. (4) The input data required in this code are few, so that it is readily usable in a time sharing system of FACOM 230-60/75 computer. Example TOPIC calculations in design study of tokamak fusion reactors (JXFR, INTOR-J) are presented. (author)
Overview of Particle and Heavy Ion Transport Code System PHITS
Sato, Tatsuhiko; Niita, Koji; Matsuda, Norihiro; Hashimoto, Shintaro; Iwamoto, Yosuke; Furuta, Takuya; Noda, Shusaku; Ogawa, Tatsuhiko; Iwase, Hiroshi; Nakashima, Hiroshi; Fukahori, Tokio; Okumura, Keisuke; Kai, Tetsuya; Chiba, Satoshi; Sihver, Lembit
2014-06-01
A general purpose Monte Carlo Particle and Heavy Ion Transport code System, PHITS, is being developed through the collaboration of several institutes in Japan and Europe. The Japan Atomic Energy Agency is responsible for managing the entire project. PHITS can deal with the transport of nearly all particles, including neutrons, protons, heavy ions, photons, and electrons, over wide energy ranges using various nuclear reaction models and data libraries. It is written in Fortran language and can be executed on almost all computers. All components of PHITS such as its source, executable and data-library files are assembled in one package and then distributed to many countries via the Research organization for Information Science and Technology, the Data Bank of the Organization for Economic Co-operation and Development's Nuclear Energy Agency, and the Radiation Safety Information Computational Center. More than 1,000 researchers have been registered as PHITS users, and they apply the code to various research and development fields such as nuclear technology, accelerator design, medical physics, and cosmic-ray research. This paper briefly summarizes the physics models implemented in PHITS, and introduces some important functions useful for specific applications, such as an event generator mode and beam transport functions.
Progress on the Data Server for the National Transport Code
Luetkemeyer, K. G.; Bateman, G.; Cary, J. R.; Fredian, T.; Greenwood, D.; Jong, R.; Wiley, J.
1999-11-01
The data server of the NTCC Demonstration Project provides a universal network interface to interpolated or raw data needed by transport codes. Data from a variety of sources is made available. CORBA is used for the networking interface. The second generation data server is now being developed. The new data server will make available data from the ITER profile database and data from TRANSP trees of MDS Plus data systems. (The MDS Plus data is retrieved via socket network calls and passed through the CORBA interface.) The use of Object Oriented Programming techniques permits data from multiple sources to be treated polymorphically, so that minimal coding is required to return the data from these multiple sources through the CORBA interface or to interpolate the data from any of the sources. The data server further makes use of exceptions to facilitate and generalize the handling of error conditions. The exception hierarchy and principles behind its design will be discussed.
RADTRAN 5 - A computer code for transportation risk analysis
International Nuclear Information System (INIS)
Neuhauser, K.S.; Kanipe, F.L.
1993-01-01
The RADTRAN 5 computer code has been developed to estimate radiological and nonradiological risks of radioactive materials transportation. RADTRAN 5 is written in ANSI standard FORTRAN 77; the code contains significant advances in the methodology first pioneered with the LINK option of RADTRAN 4. A major application of the LINK methodology is route-specific analysis. Another application is comparisons of attributes along the same route segments. Nonradiological risk factors have been incorporated to allow users to estimate nonradiological fatalities and injuries that might occur during the transportation event(s) being analyzed. These fatalities include prompt accidental fatalities from mechanical causes. Values of these risk factors for the United States have been made available in the code as optional defaults. Several new health effects models have been published in the wake of the Hiroshima-Nagasaki dosimetry reassessment, and this has emphasized the need for flexibility in the RADTRAN approach to health-effects calculations. Therefore, the basic set of health-effects conversion equations in RADTRAN have been made user-definable. All parameter values can be changed by the user, but a complete set of default values are available for both the new International Commission on Radiation Protection model (ICRP Publication 60) and the recent model of the U.S. National Research Council's Committee on the Biological Effects of Radiation (BEIR V). The meteorological input data tables have been modified to permit optional entry of maximum downwind distances for each dose isopleth. The expected dose to an individual in each isodose area is also calculated and printed automatically. Examples are given that illustrate the power and flexibility of the RADTRAN 5 computer code. (J.P.N.)
Computer codes in nuclear safety, radiation transport and dosimetry
International Nuclear Information System (INIS)
Bordy, J.M.; Kodeli, I.; Menard, St.; Bouchet, J.L.; Renard, F.; Martin, E.; Blazy, L.; Voros, S.; Bochud, F.; Laedermann, J.P.; Beaugelin, K.; Makovicka, L.; Quiot, A.; Vermeersch, F.; Roche, H.; Perrin, M.C.; Laye, F.; Bardies, M.; Struelens, L.; Vanhavere, F.; Gschwind, R.; Fernandez, F.; Quesne, B.; Fritsch, P.; Lamart, St.; Crovisier, Ph.; Leservot, A.; Antoni, R.; Huet, Ch.; Thiam, Ch.; Donadille, L.; Monfort, M.; Diop, Ch.; Ricard, M.
2006-01-01
The purpose of this conference was to describe the present state of computer codes dedicated to radiation transport or radiation source assessment or dosimetry. The presentations have been parted into 2 sessions: 1) methodology and 2) uses in industrial or medical or research domains. It appears that 2 different calculation strategies are prevailing, both are based on preliminary Monte-Carlo calculations with data storage. First, quick simulations made from a database of particle histories built though a previous Monte-Carlo simulation and secondly, a neuronal approach involving a learning platform generated through a previous Monte-Carlo simulation. This document gathers the slides of the presentations
Multidimensional electron-photon transport with standard discrete ordinates codes
International Nuclear Information System (INIS)
Drumm, C.R.
1997-01-01
A method is described for generating electron cross sections that are comparable with standard discrete ordinates codes without modification. There are many advantages of using an established discrete ordinates solver, e.g. immediately available adjoint capability. Coupled electron-photon transport capability is needed for many applications, including the modeling of the response of electronics components to space and man-made radiation environments. The cross sections have been successfully used in the DORT, TWODANT and TORT discrete ordinates codes. The cross sections are shown to provide accurate and efficient solutions to certain multidimensional electron-photon transport problems. The key to the method is a simultaneous solution of the continuous-slowing-down (CSD) portion and elastic-scattering portion of the scattering source by the Goudsmit-Saunderson theory. The resulting multigroup-Legendre cross sections are much smaller than the true scattering cross sections that they represent. Under certain conditions, the cross sections are guaranteed positive and converge with a low-order Legendre expansion
Multidimensional electron-photon transport with standard discrete ordinates codes
International Nuclear Information System (INIS)
Drumm, C.R.
1997-01-01
A method is described for generating electron cross sections that are compatible with standard discrete ordinates codes without modification. There are many advantages to using an established discrete ordinates solver, e.g., immediately available adjoint capability. Coupled electron-photon transport capability is needed for many applications, including the modeling of the response of electronics components to space and synthetic radiation environments. The cross sections have been successfully used in the DORT, TWODANT, and TORT discrete ordinates codes. The cross sections are shown to provide accurate and efficient solutions to certain multidimensional electron-photon transport problems. The key to the method is a simultaneous solution of the continuous-slowing-down and elastic-scattering portions of the scattering source by the Goudsmit-Saunderson theory. The resulting multigroup-Legendre cross sections are much smaller than the true scattering cross sections that they represent. Under certain conditions, the cross sections are guaranteed positive and converge with a low-order Legendre expansion
Physics models in the toroidal transport code PROCTR
Energy Technology Data Exchange (ETDEWEB)
Howe, H.C.
1990-08-01
The physics models that are contained in the toroidal transport code PROCTR are described in detail. Time- and space-dependent models are included for the plasma hydrogenic-ion, helium, and impurity densities, the electron and ion temperatures, the toroidal rotation velocity, and the toroidal current profile. Time- and depth-dependent models for the trapped and mobile hydrogenic particle concentrations in the wall and a time-dependent point model for the number of particles in the limiter are also included. Time-dependent models for neutral particle transport, neutral beam deposition and thermalization, fusion heating, impurity radiation, pellet injection, and the radial electric potential are included and recalculated periodically as the time-dependent models evolve. The plasma solution is obtained either in simple flux coordinates, where the radial shift of each elliptical, toroidal flux surface is included to maintain an approximate pressure equilibrium, or in general three-dimensional torsatron coordinates represented by series of helical harmonics. The detailed coupling of the plasma, scrape-off layer, limiter, and wall models through the neutral transport model makes PROCTR especially suited for modeling of recycling and particle control in toroidal plasmas. The model may also be used in a steady-state profile analysis mode for studying energy and particle balances starting with measured plasma profiles.
PRESTO low-level waste transport and risk assessment code
International Nuclear Information System (INIS)
Little, C.A.; Fields, D.E.; McDowell-Boyer, L.M.; Emerson, C.J.
1981-01-01
PRESTO (Prediction of Radiation Effects from Shallow Trench Operations) is a computer code developed under US Environmental Protection Agency (EPA) funding to evaluate possible health effects from shallow land burial trenches. The model is intended to be generic and to assess radionuclide transport, ensuing exposure, and health impact to a static local population for a 1000-y period following the end of burial operations. Human exposure scenarios considered by the model include normal releases (including leaching and operational spillage), human intrusion, and site farming or reclamation. Pathways and processes of transit from the trench to an individual or population inlude: groundwater transport, overland flow, erosion, surface water dilution, resuspension, atmospheric transport, deposition, inhalation, and ingestion of contaminated beef, milk, crops, and water. Both population doses and individual doses are calculated as well as doses to the intruder and farmer. Cumulative health effects in terms of deaths from cancer are calculated for the population over the thousand-year period using a life-table approach. Data bases are being developed for three extant shallow land burial sites: Barnwell, South Carolina; Beatty, Nevada; and West Valley, New York
New features of the mercury Monte Carlo particle transport code
International Nuclear Information System (INIS)
Procassini, Richard; Brantley, Patrick; Dawson, Shawn
2010-01-01
Several new capabilities have been added to the Mercury Monte Carlo transport code over the past four years. The most important algorithmic enhancement is a general, extensible infrastructure to support source, tally and variance reduction actions. For each action, the user defines a phase space, as well as any number of responses that are applied to a specified event. Tallies are accumulated into a correlated, multi-dimensional. Cartesian-product result phase space. Our approach employs a common user interface to specify the data sets and distributions that define the phase, response and result for each action. Modifications to the particle trackers include the use of facet halos (instead of extrapolative fuzz) for robust tracking, and material interface reconstruction for use in shape overlaid meshes. Support for expected-value criticality eigenvalue calculations has also been implemented. Computer science enhancements include an in-line Python interface for user customization of problem setup and output. (author)
Final Report for National Transport Code Collaboration PTRANSP
International Nuclear Information System (INIS)
Kritz, Arnold H.
2012-01-01
PTRANSP, which is the predictive version of the TRANSP code, was developed in a collaborative effort involving the Princeton Plasma Physics Laboratory, General Atomics Corporation, Lawrence Livermore National Laboratory, and Lehigh University. The PTRANSP/TRANSP suite of codes is the premier integrated tokamak modeling software in the United States. A production service for PTRANSP/TRANSP simulations is maintained at the Princeton Plasma Physics Laboratory; the server has a simple command line client interface and is subscribed to by about 100 researchers from tokamak projects in the US, Europe, and Asia. This service produced nearly 13000 PTRANSP/TRANSP simulations in the four year period FY 2005 through FY 2008. Major archives of TRANSP results are maintained at PPPL, MIT, General Atomics, and JET. Recent utilization, counting experimental analysis simulations as well as predictive simulations, more than doubled from slightly over 2000 simulations per year in FY 2005 and FY 2006 to over 4300 simulations per year in FY 2007 and FY 2008. PTRANSP predictive simulations applied to ITER increased eight fold from 30 simulations per year in FY 2005 and FY 2006 to 240 simulations per year in FY 2007 and FY 2008, accounting for more than half of combined PTRANSP/TRANSP service CPU resource utilization in FY 2008. PTRANSP studies focused on ITER played a key role in journal articles. Examples of validation studies carried out for momentum transport in PTRANSP simulations were presented at the 2008 IAEA conference. The increase in number of PTRANSP simulations has continued (more than 7000 TRANSP/PTRANSP simulations in 2010) and results of PTRANSP simulations appear in conference proceedings, for example the 2010 IAEA conference, and in peer reviewed papers. PTRANSP provides a bridge to the Fusion Simulation Program (FSP) and to the future of integrated modeling. Through years of widespread usage, each of the many parts of the PTRANSP suite of codes has been thoroughly
KAMCCO, a reactor physics Monte Carlo neutron transport code
International Nuclear Information System (INIS)
Arnecke, G.; Borgwaldt, H.; Brandl, V.; Lalovic, M.
1976-06-01
KAMCCO is a 3-dimensional reactor Monte Carlo code for fast neutron physics problems. Two options are available for the solution of 1) the inhomogeneous time-dependent neutron transport equation (census time scheme), and 2) the homogeneous static neutron transport equation (generation cycle scheme). The user defines the desired output, e.g. estimates of reaction rates or neutron flux integrated over specified volumes in phase space and time intervals. Such primary quantities can be arbitrarily combined, also ratios of these quantities can be estimated with their errors. The Monte Carlo techniques are mostly analogue (exceptions: Importance sampling for collision processes, ELP/MELP, Russian roulette and splitting). Estimates are obtained from the collision and track length estimators. Elastic scattering takes into account first order anisotropy in the center of mass system. Inelastic scattering is processed via the evaporation model or via the excitation of discrete levels. For the calculation of cross sections, the energy is treated as a continuous variable. They are computed by a) linear interpolation, b) from optionally Doppler broadened single level Breit-Wigner resonances or c) from probability tables (in the region of statistically distributed resonances). (orig.) [de
RADTRAN II: revised computer code to analyze transportation of radioactive material
International Nuclear Information System (INIS)
Taylor, J.M.; Daniel, S.L.
1982-10-01
A revised and updated version of the RADTRAN computer code is presented. This code has the capability to predict the radiological impacts associated with specific schemes of radioactive material shipments and mode specific transport variables
ITS - The integrated TIGER series of coupled electron/photon Monte Carlo transport codes
International Nuclear Information System (INIS)
Halbleib, J.A.; Mehlhorn, T.A.
1985-01-01
The TIGER series of time-independent coupled electron/photon Monte Carlo transport codes is a group of multimaterial, multidimensional codes designed to provide a state-of-the-art description of the production and transport of the electron/photon cascade. The codes follow both electrons and photons from 1.0 GeV down to 1.0 keV, and the user has the option of combining the collisional transport with transport in macroscopic electric and magnetic fields of arbitrary spatial dependence. Source particles can be either electrons or photons. The most important output data are (a) charge and energy deposition profiles, (b) integral and differential escape coefficients for both electrons and photons, (c) differential electron and photon flux, and (d) pulse-height distributions for selected regions of the problem geometry. The base codes of the series differ from one another primarily in their dimensionality and geometric modeling. They include (a) a one-dimensional multilayer code, (b) a code that describes the transport in two-dimensional axisymmetric cylindrical material geometries with a fully three-dimensional description of particle trajectories, and (c) a general three-dimensional transport code which employs a combinatorial geometry scheme. These base codes were designed primarily for describing radiation transport for those situations in which the detailed atomic structure of the transport medium is not important. For some applications, it is desirable to have a more detailed model of the low energy transport. The system includes three additional codes that contain a more elaborate ionization/relaxation model than the base codes. Finally, the system includes two codes that combine the collisional transport of the multidimensional base codes with transport in macroscopic electric and magnetic fields of arbitrary spatial dependence
Parallel implementation of the Monte Carlo transport code EGS4 on the hypercube
International Nuclear Information System (INIS)
Kirk, B.L.; Azmy, Y.Y.; Gabriel, T.A.; Fu, C.Y.
1991-01-01
Monte Carlo transport codes are commonly used in the study of particle interactions. The CALOR89 code system is a combination of several Monte Carlo transport and analysis programs. In order to produce good results, a typical Monte Carlo run will have to produce many particle histories. On a single processor computer, the transport calculation can take a huge amount of time. However, if the transport of particles were divided among several processors in a multiprocessor machine, the time can be drastically reduced
Transport calculations with the BALDUR code. Pt. 1
International Nuclear Information System (INIS)
Lackner, K.; Wunderlich, R.
1979-12-01
1-d transport calculations with the BALDUR-code are described for predicting the performance of ZEPHYR under D-T operation. Results presented in this report refer to the impurity-free case, and ion and electron heat conduction losses described by CHIsub(i) = neoclassical and CHIsub(e) = 6.25 x 10 17 /nsub(e) (cgs-units). A simple refuelling scenario taking account of the density limit for the ohmic heating phase, the contribution of neutral injection to the refuelling rate and the need for an approximately balanced D-T mixture at the instance of ignition is adopted. The heating scenario assumes a neutral injection beam with 160 keV particle energy in the main component, with a duration of 1.1 sec. Major radius compression by a factor of 1.5 starts 1 sec after the onset of neutral injection and lasts 100 msec. For this standard scenario the performance is studied in different density regimes and for different neutral injection powers. Under the above assumption ignition is predicted for total neutral injection powers < approx. 16 MW (9.6 MW in the main energy component) and average total β-values < 2.8%. Results including impurities, alternative scaling laws, and deviations from the standard scenario will be presented in another report. (orig.) 891 GG/orig. 892 HIS
Preparing diagnostic data for the SNAP transport code
International Nuclear Information System (INIS)
Murphy, J.A.; Scott, S.D.; Towner, H.H.
1992-01-01
This paper describes the program SNAPIN which is used to prepare data for transport analysis with the SNAP code. The data input to SNAP includes diagnostic profiles [n e (R), T e (R), T i (R), v φ (R), Z eff (R), P rad (R)] and measurements such as total plasma current, R major , beam power, gas puff rate, etc. SNAPIN reads in the necessary TFTR data, allows editing of that data, including graphical editing of profile data and the selection of physics models. SNAPIN allows comparison of profile data from all diagnostics that measure a quantity, for example, electron temperature profiles from Thomson scattering and electron cyclotron emission (ECE). A powerful user interface is important to help the user prepare input data sets quickly and consistently, because hundreds of variables must be specified for each analysis. SNAPIN facilitates this by a careful organization of menus, display of all scalar data and switch settings within the menus, the graphical editing and comparison of profiles, and step-by-step checking for consistent physics controls [J. Murphy, S. Scott, and H. Towner, The SNAP User's Guide, Technical Report PPPL-TM-393, Princeton Plasma Physics Laboratory (1992)
The KFA-Version of the high-energy transport code HETC and the generalized evaluation code SIMPEL
International Nuclear Information System (INIS)
Cloth, P.; Filges, D.; Sterzenbach, G.; Armstrong, T.W.; Colborn, B.L.
1983-03-01
This document describes the updates that have been made to the high-energy transport code HETC for use in the German spallation-neutron source project SNQ. Performance and purpose of the subsidiary code SIMPEL that has been written for general analysis of the HETC output are also described. In addition means of coupling to low energy transport programs, such as the Monte-Carlo code MORSE is provided. As complete input descriptions for HETC and SIMPEL are given together with a sample problem, this document can serve as a user's manual for these two codes. The document is also an answer to the demand that has been issued by a greater community of HETC users on the ICANS-IV meeting, Oct 20-24 1980, Tsukuba-gun, Japan for a complete description of at least one single version of HETC among the many different versions that exist. (orig.)
Regional Atmospheric Transport Code for Hanford Emission Tracking, Version 2(RATCHET2)
Energy Technology Data Exchange (ETDEWEB)
Ramsdell, James V.; Rishel, Jeremy P.
2006-07-01
This manual describes the atmospheric model and computer code for the Atmospheric Transport Module within SAC. The Atmospheric Transport Module, called RATCHET2, calculates the time-integrated air concentration and surface deposition of airborne contaminants to the soil. The RATCHET2 code is an adaptation of the Regional Atmospheric Transport Code for Hanford Emissions Tracking (RATCHET). The original RATCHET code was developed to perform the atmospheric transport for the Hanford Environmental Dose Reconstruction Project. Fundamentally, the two sets of codes are identical; no capabilities have been deleted from the original version of RATCHET. Most modifications are generally limited to revision of the run-specification file to streamline the simulation process for SAC.
Path Toward a Unifid Geometry for Radiation Transport
Lee, Kerry; Barzilla, Janet; Davis, Andrew; Zachmann
2014-01-01
The Direct Accelerated Geometry for Radiation Analysis and Design (DAGRAD) element of the RadWorks Project under Advanced Exploration Systems (AES) within the Space Technology Mission Directorate (STMD) of NASA will enable new designs and concepts of operation for radiation risk assessment, mitigation and protection. This element is designed to produce a solution that will allow NASA to calculate the transport of space radiation through complex computer-aided design (CAD) models using the state-of-the-art analytic and Monte Carlo radiation transport codes. Due to the inherent hazard of astronaut and spacecraft exposure to ionizing radiation in low-Earth orbit (LEO) or in deep space, risk analyses must be performed for all crew vehicles and habitats. Incorporating these analyses into the design process can minimize the mass needed solely for radiation protection. Transport of the radiation fields as they pass through shielding and body materials can be simulated using Monte Carlo techniques or described by the Boltzmann equation, which is obtained by balancing changes in particle fluxes as they traverse a small volume of material with the gains and losses caused by atomic and nuclear collisions. Deterministic codes that solve the Boltzmann transport equation, such as HZETRN [high charge and energy transport code developed by NASA Langley Research Center (LaRC)], are generally computationally faster than Monte Carlo codes such as FLUKA, GEANT4, MCNP(X) or PHITS; however, they are currently limited to transport in one dimension, which poorly represents the secondary light ion and neutron radiation fields. NASA currently uses HZETRN space radiation transport software, both because it is computationally efficient and because proven methods have been developed for using this software to analyze complex geometries. Although Monte Carlo codes describe the relevant physics in a fully three-dimensional manner, their computational costs have thus far prevented their
International Nuclear Information System (INIS)
VOOGD, J.A.
1999-01-01
An analysis of three software proposals is performed to recommend a computer code for immobilized low activity waste flow and transport modeling. The document uses criteria restablished in HNF-1839, ''Computer Code Selection Criteria for Flow and Transport Codes to be Used in Undisturbed Vadose Zone Calculation for TWRS Environmental Analyses'' as the basis for this analysis
Development of TIGER code for radionuclide transport in a geochemically evolving region
International Nuclear Information System (INIS)
Mihara, Morihiro; Ooi, Takao
2004-01-01
In a transuranic (TRU) waste geological disposal facility, using cementitious materials is being considered. Cementitious materials will gradually dissolve in groundwater over the long-term. In the performance assessment report of a TRU waste repository in Japan already published, the most conservative radionuclide migration parameter set was selected considering the evolving cementitious material. Therefore, a tool to perform the calculation of radionuclide transport considering long-term geochemically evolving cementitious materials, named the TIGER code, Transport In Geochemically Evolving Region was developed to calculate a more realistic performance assessment. It can calculate radionuclide transport in engineered and natural barrier systems. In this report, mathematical equations of this code are described and validated with analytical solutions and results of other codes for radionuclide transport. The more realistic calculation of radionuclide transport for a TRU waste geological disposal system using the TIGER code could be performed. (author)
Development of a transient three-dimensional neutron transport code with feedback
Energy Technology Data Exchange (ETDEWEB)
Waddell, M.W. Jr.
1994-07-19
A new code is being developed at the Y-12 Plant for solving the time-dependent, three-dimensional Boltzmann transport model with feedback. The new code, PADK, uses the quasi-static method in its adiabatic form and is to be utilized to analyze hypothetical criticality accidents. A description of the code along with preliminary results without feedback are presented in this paper. The code is applied to 2 standard benchmark problems and the results are compared to another method. Also, the code is used to model the GODIVA reactor. Further work needed to be completed is described.
Development of a transient three-dimensional neutron transport code with feedback
Energy Technology Data Exchange (ETDEWEB)
Waddell, M.W. Jr.
1994-12-31
A new code is being developed at the Y-12 plant for solving the time-dependent, three-dimensional Boltzmann transport model with feedback. The new code, PADK, uses the quasi-static method in its adiabatic form and is to be utilized to analyze hypothetical criticality accidents. A description of the code along with preliminary results without feedback are presented in this paper. The code is applied to two standard benchmark problems, and the results are compared to another method. Also, the code is used to model the GODIVA reactor. Further work needed to be completed is described.
Accelerating execution of the integrated TIGER series Monte Carlo radiation transport codes
International Nuclear Information System (INIS)
Smith, L.M.; Hochstedler, R.D.
1997-01-01
Execution of the integrated TIGER series (ITS) of coupled electron/photon Monte Carlo radiation transport codes has been accelerated by modifying the FORTRAN source code for more efficient computation. Each member code of ITS was benchmarked and profiled with a specific test case that directed the acceleration effort toward the most computationally intensive subroutines. Techniques for accelerating these subroutines included replacing linear search algorithms with binary versions, replacing the pseudo-random number generator, reducing program memory allocation, and proofing the input files for geometrical redundancies. All techniques produced identical or statistically similar results to the original code. Final benchmark timing of the accelerated code resulted in speed-up factors of 2.00 for TIGER (the one-dimensional slab geometry code), 1.74 for CYLTRAN (the two-dimensional cylindrical geometry code), and 1.90 for ACCEPT (the arbitrary three-dimensional geometry code)
Accelerating execution of the integrated TIGER series Monte Carlo radiation transport codes
Smith, L. M.; Hochstedler, R. D.
1997-02-01
Execution of the integrated TIGER series (ITS) of coupled electron/photon Monte Carlo radiation transport codes has been accelerated by modifying the FORTRAN source code for more efficient computation. Each member code of ITS was benchmarked and profiled with a specific test case that directed the acceleration effort toward the most computationally intensive subroutines. Techniques for accelerating these subroutines included replacing linear search algorithms with binary versions, replacing the pseudo-random number generator, reducing program memory allocation, and proofing the input files for geometrical redundancies. All techniques produced identical or statistically similar results to the original code. Final benchmark timing of the accelerated code resulted in speed-up factors of 2.00 for TIGER (the one-dimensional slab geometry code), 1.74 for CYLTRAN (the two-dimensional cylindrical geometry code), and 1.90 for ACCEPT (the arbitrary three-dimensional geometry code).
Modification of PRETOR Code to Be Applied to Transport Simulation in Stellarators
Energy Technology Data Exchange (ETDEWEB)
Fontanet, J.; Castejon, F.; Dies, J.; Fontdecaba, J.; Alejaldre, C.
2001-07-01
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.
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
Use of the Apollo-II multigroup transport code for criticality calculations
International Nuclear Information System (INIS)
Coste, M.; Mathonniere, G.; Sanchez, R.; Stankovski, Z.; Van der Gucht, C.; Zmijarevic, I.
1992-01-01
APPOLO-II is a new-generation multigroup transport code for assembly calculation. The code has been designed to be used as a tool for reactor design as well as for the analysis and interpretation of small nuclear facilities. As the first step in a criticality calculation, the collision probability module of the APPOLO-II code can be used to generate cell or assembly homogenized reaction-rate preserving cross sections that account for self-shielding effects as well as for the fine-energy within cell flux spectral variations. These cross section data can then be used either directly within the APPOLO-II code in a direct discrete ordinate multigroup transport calculation of a small nuclear facility or, more generally, be formatted by a post-processing module to be used by the multigroup diffusion code CRONOS-II or by the multigroup Monte Carlo code TRIMARAN
A Monte Carlo Code for Relativistic Radiation Transport Around Kerr Black Holes
Schnittman, Jeremy David; Krolik, Julian H.
2013-01-01
We present a new code for radiation transport around Kerr black holes, including arbitrary emission and absorption mechanisms, as well as electron scattering and polarization. The code is particularly useful for analyzing accretion flows made up of optically thick disks and optically thin coronae. We give a detailed description of the methods employed in the code and also present results from a number of numerical tests to assess its accuracy and convergence.
A predictive transport modeling code for ICRF-heated tokamaks
International Nuclear Information System (INIS)
Phillips, C.K.; Hwang, D.Q.
1992-02-01
In this report, a detailed description of the physic included in the WHIST/RAZE package as well as a few illustrative examples of the capabilities of the package will be presented. An in depth analysis of ICRF heating experiments using WHIST/RAZE will be discussed in a forthcoming report. A general overview of philosophy behind the structure of the WHIST/RAZE package, a summary of the features of the WHIST code, and a description of the interface to the RAZE subroutines are presented in section 2 of this report. Details of the physics contained in the RAZE code are examined in section 3. Sample results from the package follow in section 4, with concluding remarks and a discussion of possible improvements to the package discussed in section 5
Environmental, Transient, Three-Dimensional, Hydrothermal, Mass Transport Code - FLESCOT
Energy Technology Data Exchange (ETDEWEB)
Onishi, Yasuo [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bao, Jie [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Glass, Kevin A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Eyler, L. L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Okumura, Masahiko [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)
2015-03-28
The purpose of the project was to modify and apply the transient, three-dimensional FLESCOT code to be able to effectively simulate cesium behavior in Fukushima lakes/dam reservoirs, river mouths, and coastal areas. The ultimate objective of the FLESCOT simulation is to predict future changes of cesium accumulation in Fukushima area reservoirs and costal water. These evaluation results will assist ongoing and future environmental remediation activities and policies in a systematic and comprehensive manner.
HETFIS: High-Energy Nucleon-Meson Transport Code with Fission
Energy Technology Data Exchange (ETDEWEB)
Barish, J.; Gabriel, T.A.; Alsmiller, F.S.; Alsmiller, R.G. Jr.
1981-07-01
A model that includes fission for predicting particle production spectra from medium-energy nucleon and pion collisions with nuclei (Z greater than or equal to 91) has been incorporated into the nucleon-meson transport code, HETC. This report is primarily concerned with the programming aspects of HETFIS (High-Energy Nucleon-Meson Transport Code with Fission). A description of the program data and instructions for operating the code are given. HETFIS is written in FORTRAN IV for the IBM computers and is readily adaptable to other systems.
Resolution of the neutron transport equation by massively parallel computer in the Cronos code
International Nuclear Information System (INIS)
Zardini, D.M.
1996-01-01
The feasibility of neutron transport problems parallel resolution by CRONOS code's SN module is here studied. In this report we give the first data about the parallel resolution by angular variable decomposition of the transport equation. Problems about parallel resolution by spatial variable decomposition and memory stage limits are also explained here. (author)
Low-discrepancy point sets in transport codes
Energy Technology Data Exchange (ETDEWEB)
Warnock, T.T.
1985-01-01
A drawback to Monte Carlo methods of computation is its rate of convergence. There are methods of sampling that have a better error estimate than those using random numbers. This paper gives the result of some preliminary experiments with these sampling methods on two neutron transport problems.
Multiple-canister flow and transport code in 2-dimensional space. MCFT2D: user's manual
International Nuclear Information System (INIS)
Lim, Doo-Hyun
2006-03-01
A two-dimensional numerical code, MCFT2D (Multiple-Canister Flow and Transport code in 2-Dimensional space), has been developed for groundwater flow and radionuclide transport analyses in a water-saturated high-level radioactive waste (HLW) repository with multiple canisters. A multiple-canister configuration and a non-uniform flow field of the host rock are incorporated in the MCFT2D code. Effects of heterogeneous flow field of the host rock on migration of nuclides can be investigated using MCFT2D. The MCFT2D enables to take into account the various degrees of the dependency of canister configuration for nuclide migration in a water-saturated HLW repository, while the dependency was assumed to be either independent or perfectly dependent in previous studies. This report presents features of the MCFT2D code, numerical simulation using MCFT2D code, and graphical representation of the numerical results. (author)
Code of Practice for the safe transport of radioactive substances 1990
International Nuclear Information System (INIS)
1990-01-01
This Federal Code revises an earlier Code on the same subject issued in 1982 and was formulated under the Environment Protection (Nuclear Codes) Act 1978. The purpose of the Code is to establish uniform safety standards, applicable throughout the Commonwealth of Australia, to provide for the protection of persons and the environment, against any dangers associated with the transport of radioactive substances. The Code uses as a basis the IAEA Regulations for the Safe Transport of Radioactive Materials. This new edition takes into account the 1985 Edition of the Regulations incorporating the 1988 Supplement and provides, furthermore, that radiation protection standards will also be subject to recommendations of the Australian National Health and Medical Research Council [fr
Energy Technology Data Exchange (ETDEWEB)
Pruess, Karsten
2003-08-08
Numerical simulation has become a widely practiced andaccepted technique for studying flow and transport processes in thevadose zone and other subsurface flow systems. This article discusses asuite of codes, developed primarily at Lawrence Berkeley NationalLaboratory (LBNL), with the capability to model multiphase flows withphase change. We summarize history and goals in the development of theTOUGH codes, and present the governing equations for multiphase,multicomponent flow. Special emphasis is given to space discretization bymeans of integral finite differences (IFD). Issues of code implementationand architecture are addressed, as well as code applications,maintenance, and future developments.
Application of neutron/gamma transport codes for the design of explosive detection systems
International Nuclear Information System (INIS)
Elias, E.; Shayer, Z.
1994-01-01
Applications of neutron and gamma transport codes to the design of nuclear techniques for detecting concealed explosives material are discussed. The methodology of integrating radiation transport computations in the development, optimization and analysis phases of these new technologies is discussed. Transport and Monte Carlo codes are used for proof of concepts, guide the system integration, reduce the extend of experimental program and provide insight into the physical problem involved. The paper concentrates on detection techniques based on thermal and fast neutron interactions in the interrogated object. (authors). 6 refs., 1 tab., 5 figs
Plasmator. A numerical code for simulation of plasma transport in Tokamaks
International Nuclear Information System (INIS)
Guasp, J.
1979-01-01
Plasmator is a flexible monodimensional numerical code for plasma transport in Tokamaks of circular cross-section, it allows neutral particle transport and impurity effects. The code leaves a total freedom in the analytical form of transport coefficients. It has been writen in Fortran-V for the UNIVAC-1100/80 from JEN and allows for the possibility of graphics for radial profiles and temporal evolution of the main plasma magnitudes, as well in three-dimensional as in two-dimensional representation either on a Calcomp plotter or in the printer. (author)
Numerical model for two-dimensional hydrodynamics and energy transport. [VECTRA code
Energy Technology Data Exchange (ETDEWEB)
Trent, D.S.
1973-06-01
The theoretical basis and computational procedure of the VECTRA computer program are presented. VECTRA (Vorticity-Energy Code for TRansport Analysis) is designed for applying numerical simulation to a broad range of intake/discharge flows in conjunction with power plant hydrological evaluation. The code computational procedure is based on finite-difference approximation of the vorticity-stream function partial differential equations which govern steady flow momentum transport of two-dimensional, incompressible, viscous fluids in conjunction with the transport of heat and other constituents.
Intact coding region of the serotonin transporter gene in obsessive-compulsive disorder
Energy Technology Data Exchange (ETDEWEB)
Altemus, M.; Murphy, D.L.; Greenberg, B. [NIMH, NIH, Bethesda, MD (United States); Lesch, K.P. [Univ. of Wuerzburg (Germany)
1996-07-26
Epidemiologic studies indicate that obsessive-compulsive disorder is genetically transmitted in some families, although no genetic abnormalities have been identified in individuals with this disorder. The selective response of obsessive-compulsive disorder to treatment with agents which block serotonin reuptake suggests the gene coding for the serotonin transporter as a candidate gene. The primary structure of the serotonin-transporter coding region was sequenced in 22 patients with obsessive-compulsive disorder, using direct PCR sequencing of cDNA synthesized from platelet serotonin-transporter mRNA. No variations in amino acid sequence were found among the obsessive-compulsive disorder patients or healthy controls. These results do not support a role for alteration in the primary structure of the coding region of the serotonin-transporter gene in the pathogenesis of obsessive-compulsive disorder. 27 refs.
SCATTER: Source and Transport of Emplaced Radionuclides: Code documentation
International Nuclear Information System (INIS)
Longsine, D.E.
1987-03-01
SCATTER simulated several processes leading to the release of radionuclides to the site subsystem and then simulates transport via the groundwater of the released radionuclides to the biosphere. The processes accounted for to quantify release rates to a ground-water migration path include radioactive decay and production, leaching, solubilities, and the mixing of particles with incoming uncontaminated fluid. Several decay chains of arbitrary length can be considered simultaneously. The release rates then serve as source rates to a numerical technique which solves convective-dispersive transport for each decay chain. The decay chains are allowed to have branches and each member can have a different radioactive factor. Results are cast as radionuclide discharge rates to the accessible environment
The use of Monte Carlo radiation transport codes in radiation physics and dosimetry
CERN. Geneva; Ferrari, Alfredo; Silari, Marco
2006-01-01
Transport and interaction of electromagnetic radiation Interaction models and simulation schemes implemented in modern Monte Carlo codes for the simulation of coupled electron-photon transport will be briefly reviewed. In these codes, photon transport is simulated by using the detailed scheme, i.e., interaction by interaction. Detailed simulation is easy to implement, and the reliability of the results is only limited by the accuracy of the adopted cross sections. Simulations of electron and positron transport are more difficult, because these particles undergo a large number of interactions in the course of their slowing down. Different schemes for simulating electron transport will be discussed. Condensed algorithms, which rely on multiple-scattering theories, are comparatively fast, but less accurate than mixed algorithms, in which hard interactions (with energy loss or angular deflection larger than certain cut-off values) are simulated individually. The reliability, and limitations, of electron-interacti...
Progress on RMC: a Monte Carlo neutron transport code for reactor analysis
International Nuclear Information System (INIS)
Wang, Kan; Li, Zeguang; She, Ding; Liu, Yuxuan; Xu, Qi; Shen, Huayun; Yu, Ganglin
2011-01-01
This paper presents a new 3-D Monte Carlo neutron transport code named RMC (Reactor Monte Carlo code), specifically intended for reactor physics analysis. This code is being developed by Department of Engineering Physics in Tsinghua University and written in C++ and Fortran 90 language with the latest version of RMC 2.5.0. The RMC code uses the method known as the delta-tracking method to simulate neutron transport, the advantages of which include fast simulation in complex geometries and relatively simple handling of complicated geometrical objects. Some other techniques such as computational-expense oriented method and hash-table method have been developed and implemented in RMC to speedup the calculation. To meet the requirements of reactor analysis, the RMC code has the calculational functions including criticality calculation, burnup calculation and also kinetics simulation. In this paper, comparison calculations of criticality problems, burnup problems and transient problems are carried out using RMC code and other Monte Carlo codes, and the results show that RMC performs quite well in these kinds of problems. Based on MPI, RMC succeeds in parallel computation and represents a high speed-up. This code is still under intensive development and the further work directions are mentioned at the end of this paper. (author)
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
International Nuclear Information System (INIS)
Homma, Y.; Moriwaki, H.; Ikeda, K.; Ohdi, S.
2013-01-01
This paper deals with the verification of the 3 dimensional triangular prismatic discrete ordinates transport calculation code ENSEMBLE-TRIZ by comparison with the multi-group Monte Carlo calculation code GMVP in a large fast breeder reactor. The reactor is a 750 MWe electric power sodium cooled reactor. Nuclear characteristics are calculated at the beginning of cycle of an initial core and at the beginning and the end of cycle of an equilibrium core. According to the calculations, the differences between the two methodologies are smaller than 0.0002 Δk in the multiplication factor, relatively about 1% in the control rod reactivity, and 1% in the sodium void reactivity. (authors)
International Nuclear Information System (INIS)
Brenner, D.J.; Prael, R.E.; Little, R.C.
1987-01-01
Realistic simulations of the passage of fast neutrons through tissue require a large quantity of cross-sectional data. What are needed are differential (in particle type, energy and angle) cross sections. A computer code is described which produces such spectra for neutrons above ∼14 MeV incident on light nuclei such as carbon and oxygen. Comparisons have been made with experimental measurements of double-differential secondary charged-particle production on carbon and oxygen at energies from 27 to 60 MeV; they indicate that the model is adequate in this energy range. In order to utilize fully the results of these calculations, they should be incorporated into a neutron transport code. This requires defining a generalized format for describing charged-particle production, putting the calculated results in this format, interfacing the neutron transport code with these data, and charged-particle transport. The design and development of such a program is described. 13 refs., 3 figs
MC++: A parallel, portable, Monte Carlo neutron transport code in C++
International Nuclear Information System (INIS)
Lee, S.R.; Cummings, J.C.; Nolen, S.D.
1997-01-01
MC++ is an implicit multi-group Monte Carlo neutron transport code written in C++ and based on the Parallel Object-Oriented Methods and Applications (POOMA) class library. MC++ runs in parallel on and is portable to a wide variety of platforms, including MPPs, SMPs, and clusters of UNIX workstations. MC++ is being developed to provide transport capabilities to the Accelerated Strategic Computing Initiative (ASCI). It is also intended to form the basis of the first transport physics framework (TPF), which is a C++ class library containing appropriate abstractions, objects, and methods for the particle transport problem. The transport problem is briefly described, as well as the current status and algorithms in MC++ for solving the transport equation. The alpha version of the POOMA class library is also discussed, along with the implementation of the transport solution algorithms using POOMA. Finally, a simple test problem is defined and performance and physics results from this problem are discussed on a variety of platforms
Energy Technology Data Exchange (ETDEWEB)
Bordy, J.M.; Kodeli, I.; Menard, St.; Bouchet, J.L.; Renard, F.; Martin, E.; Blazy, L.; Voros, S.; Bochud, F.; Laedermann, J.P.; Beaugelin, K.; Makovicka, L.; Quiot, A.; Vermeersch, F.; Roche, H.; Perrin, M.C.; Laye, F.; Bardies, M.; Struelens, L.; Vanhavere, F.; Gschwind, R.; Fernandez, F.; Quesne, B.; Fritsch, P.; Lamart, St.; Crovisier, Ph.; Leservot, A.; Antoni, R.; Huet, Ch.; Thiam, Ch.; Donadille, L.; Monfort, M.; Diop, Ch.; Ricard, M
2006-07-01
The purpose of this conference was to describe the present state of computer codes dedicated to radiation transport or radiation source assessment or dosimetry. The presentations have been parted into 2 sessions: 1) methodology and 2) uses in industrial or medical or research domains. It appears that 2 different calculation strategies are prevailing, both are based on preliminary Monte-Carlo calculations with data storage. First, quick simulations made from a database of particle histories built though a previous Monte-Carlo simulation and secondly, a neuronal approach involving a learning platform generated through a previous Monte-Carlo simulation. This document gathers the slides of the presentations.
The beta equilibrium, stability, and transport codes. Applications to the design of stellarators
International Nuclear Information System (INIS)
Bauer, F.; Garabedian, P.; Betancourt, O.; Wakatani, M.
1987-01-01
This book gives a detailed exposition of the available computational methods, documents the codes, and presents many examples showing how to run them and how to interpret the results. A listing of the recently completed BETA transport code is included. Current stellarator experiments are discussed, and the book contains significant applications to the design of major new stellarator experiments that are now in the planning stage
Development of three-dimensional transport code by the double finite element method
International Nuclear Information System (INIS)
Fujimura, Toichiro
1985-01-01
Development of a three-dimensional neutron transport code by the double finite element method is described. Both of the Galerkin and variational methods are adopted to solve the problem, and then the characteristics of them are compared. Computational results of the collocation method, developed as a technique for the vaviational one, are illustrated in comparison with those of an Ssub(n) code. (author)
Radiation transport phenomena and modeling. Part A: Codes; Part B: Applications with examples
Energy Technology Data Exchange (ETDEWEB)
Lorence, L.J. Jr.; Beutler, D.E. [Sandia National Labs., Albuquerque, NM (United States). Simulation Technology Research Dept.
1997-09-01
This report contains the notes from the second session of the 1997 IEEE Nuclear and Space Radiation Effects Conference Short Course on Applying Computer Simulation Tools to Radiation Effects Problems. Part A discusses the physical phenomena modeled in radiation transport codes and various types of algorithmic implementations. Part B gives examples of how these codes can be used to design experiments whose results can be easily analyzed and describes how to calculate quantities of interest for electronic devices.
FLAME: A finite element computer code for contaminant transport n variably-saturated media
International Nuclear Information System (INIS)
Baca, R.G.; Magnuson, S.O.
1992-06-01
A numerical model was developed for use in performance assessment studies at the INEL. The numerical model referred to as the FLAME computer code, is designed to simulate subsurface contaminant transport in a variably-saturated media. The code can be applied to model two-dimensional contaminant transport in an and site vadose zone or in an unconfined aquifer. In addition, the code has the capability to describe transport processes in a porous media with discrete fractures. This report presents the following: description of the conceptual framework and mathematical theory, derivations of the finite element techniques and algorithms, computational examples that illustrate the capability of the code, and input instructions for the general use of the code. The development of the FLAME computer code is aimed at providing environmental scientists at the INEL with a predictive tool for the subsurface water pathway. This numerical model is expected to be widely used in performance assessments for: (1) the Remedial Investigation/Feasibility Study process and (2) compliance studies required by the US Department of energy Order 5820.2A
Srna-Monte Carlo codes for proton transport simulation in combined and voxelized geometries
International Nuclear Information System (INIS)
Ilic, R.D.; Lalic, 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 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. (author)
Srna-Monte Carlo codes for proton transport simulation in combined and voxelized geometries
Ilic, R D; Stankovic, S J
2002-01-01
This paper describes new Monte Carlo codes for proton transport simulations in complex geometrical forms and in materials of different composition. The SRNA codes were developed for three dimensional (3D) dose distribution calculation in proton therapy and dosimetry. The model of these codes is based on the theory of proton multiple scattering and a simple model of compound nucleus decay. The developed package consists of two codes: SRNA-2KG and SRNA-VOX. The first code simulates proton transport in combined geometry that can be described by planes and second order surfaces. The second one uses the voxelized geometry of material zones and is specifically adopted for the application of patient computer tomography data. Transition probabilities for both codes are given by the SRNADAT program. In this paper, we will present the models and algorithms of our programs, as well as the results of the numerical experiments we have carried out applying them, along with the results of proton transport simulation obtaine...
Benchmark test of drift-kinetic and gyrokinetic codes through neoclassical transport simulations
International Nuclear Information System (INIS)
Satake, S.; Sugama, H.; Watanabe, T.-H.; Idomura, Yasuhiro
2009-09-01
Two simulation codes that solve the drift-kinetic or gyrokinetic equation in toroidal plasmas are benchmarked by comparing the simulation results of neoclassical transport. The two codes are the drift-kinetic δf Monte Carlo code (FORTEC-3D) and the gyrokinetic full- f Vlasov code (GT5D), both of which solve radially-global, five-dimensional kinetic equation with including the linear Fokker-Planck collision operator. In a tokamak configuration, neoclassical radial heat flux and the force balance relation, which relates the parallel mean flow with radial electric field and temperature gradient, are compared between these two codes, and their results are also compared with the local neoclassical transport theory. It is found that the simulation results of the two codes coincide very well in a wide rage of plasma collisionality parameter ν * = 0.01 - 10 and also agree with the theoretical estimations. The time evolution of radial electric field and particle flux, and the radial profile of the geodesic acoustic mode frequency also coincide very well. These facts guarantee the capability of GT5D to simulate plasma turbulence transport with including proper neoclassical effects of collisional diffusion and equilibrium radial electric field. (author)
Srna - Monte Carlo codes for proton transport simulation in combined and voxelized geometries
Directory of Open Access Journals (Sweden)
Ilić Radovan D.
2002-01-01
Full Text Available This paper describes new Monte Carlo codes for proton transport simulations in complex geometrical forms and in materials of different composition. The SRNA codes were developed for three dimensional (3D dose distribution calculation in proton therapy and dosimetry. The model of these codes is based on the theory of proton multiple scattering and a simple model of compound nucleus decay. The developed package consists of two codes: SRNA-2KG and SRNA-VOX. The first code simulates proton transport in combined geometry that can be described by planes and second order surfaces. The second one uses the voxelized geometry of material zones and is specifically adopted for the application of patient computer tomography data. Transition probabilities for both codes are given by the SRNADAT program. In this paper, we will present the models and algorithms of our programs, as well as the results of the numerical experiments we have carried out applying them, along with the results of proton transport simulation obtained through the PETRA and GEANT programs. The simulation of the proton beam characterization by means of the Multi-Layer Faraday Cup and spatial distribution of positron emitters obtained by our program indicate the imminent application of Monte Carlo techniques in clinical practice.
ITS Version 6 : the integrated TIGER series of coupled electron/photon Monte Carlo transport codes.
Energy Technology Data Exchange (ETDEWEB)
Franke, Brian Claude; Kensek, Ronald Patrick; Laub, Thomas William
2008-04-01
ITS is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of lineartime-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. Our goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 6, the latest version of ITS, contains (1) improvements to the ITS 5.0 codes, and (2) conversion to Fortran 90. The general user friendliness of the software has been enhanced through memory allocation to reduce the need for users to modify and recompile the code.
FLAME: A finite element computer code for contaminant transport n variably-saturated media
Energy Technology Data Exchange (ETDEWEB)
Baca, R.G.; Magnuson, S.O.
1992-06-01
A numerical model was developed for use in performance assessment studies at the INEL. The numerical model referred to as the FLAME computer code, is designed to simulate subsurface contaminant transport in a variably-saturated media. The code can be applied to model two-dimensional contaminant transport in an and site vadose zone or in an unconfined aquifer. In addition, the code has the capability to describe transport processes in a porous media with discrete fractures. This report presents the following: description of the conceptual framework and mathematical theory, derivations of the finite element techniques and algorithms, computational examples that illustrate the capability of the code, and input instructions for the general use of the code. The development of the FLAME computer code is aimed at providing environmental scientists at the INEL with a predictive tool for the subsurface water pathway. This numerical model is expected to be widely used in performance assessments for: (1) the Remedial Investigation/Feasibility Study process and (2) compliance studies required by the US Department of energy Order 5820.2A.
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)
Development of a relativistic Particle In Cell code PARTDYN for linear accelerator beam transport
Phadte, D.; Patidar, C. B.; Pal, M. K.
2017-04-01
A relativistic Particle In Cell (PIC) code PARTDYN is developed for the beam dynamics simulation of z-continuous and bunched beams. The code is implemented in MATLAB using its MEX functionality which allows both ease of development as well higher performance similar to a compiled language like C. The beam dynamics calculations carried out by the code are compared with analytical results and with other well developed codes like PARMELA and BEAMPATH. The effect of finite number of simulation particles on the emittance growth of intense beams has been studied. Corrections to the RF cavity field expressions were incorporated in the code so that the fields could be calculated correctly. The deviations of the beam dynamics results between PARTDYN and BEAMPATH for a cavity driven in zero-mode have been discussed. The beam dynamics studies of the Low Energy Beam Transport (LEBT) using PARTDYN have been presented.
AlfaMC: A fast alpha particle transport Monte Carlo code
Energy Technology Data Exchange (ETDEWEB)
Peralta, Luis, E-mail: luis@lip.pt [Faculdade de Ciências da Universidade de Lisboa (Portugal); Laboratório de Instrumentação e Física Experimental de Partículas (Portugal); Louro, Alina [Laboratório de Instrumentação e Física Experimental de Partículas (Portugal)
2014-02-11
AlfaMC is a Monte Carlo simulation code for the transport of alpha particles. This code is based on the Continuous Slowing Down Approximation and uses the NIST/ASTAR stopping-power database. The code uses a powerful geometrical package, which allows coding of complex geometries. A flexible histogramming package is used as well, which greatly eases the scoring of results. The code is tailored for microdosimetric applications in which speed is a key factor. Comparison with the SRIM code is made for deposited energy in thin layers and range for air, mylar, aluminum and gold. The general agreement between the two codes is good for beam energies between 1 and 12 MeV. -- Highlights: • AlfaMC is a Monte Carlo program for fast alpha particle transport in matter. • The model is accurate within a few percent in the energy range of 1–12 MeV. • AlfaMC uses a combinatorial geometry package allowing the modeling of complex bodies.
A portable, parallel, object-oriented Monte Carlo neutron transport code in C++
International Nuclear Information System (INIS)
Lee, S.R.; Cummings, J.C.; Nolen, S.D.
1997-01-01
We have developed a multi-group Monte Carlo neutron transport code using C++ and the Parallel Object-Oriented Methods and Applications (POOMA) class library. This transport code, called MC++, currently computes k and α-eigenvalues and is portable to and runs parallel on a wide variety of platforms, including MPPs, clustered SMPs, and individual workstations. It contains appropriate classes and abstractions for particle transport and, through the use of POOMA, for portable parallelism. Current capabilities of MC++ are discussed, along with physics and performance results on a variety of hardware, including all Accelerated Strategic Computing Initiative (ASCI) hardware. Current parallel performance indicates the ability to compute α-eigenvalues in seconds to minutes rather than hours to days. Future plans and the implementation of a general transport physics framework are also discussed
Françoise Benz
2006-01-01
2005-2006 ACADEMIC TRAINING PROGRAMME LECTURE SERIES 27, 28, 29 June 11:00-12:00 - TH Conference Room, bldg. 4 The use of Monte Carlo radiation transport codes in radiation physics and dosimetry F. Salvat Gavalda,Univ. de Barcelona, A. FERRARI, CERN-AB, M. SILARI, CERN-SC Lecture 1. Transport and interaction of electromagnetic radiation F. Salvat Gavalda,Univ. de Barcelona Interaction models and simulation schemes implemented in modern Monte Carlo codes for the simulation of coupled electron-photon transport will be briefly reviewed. Different schemes for simulating electron transport will be discussed. Condensed algorithms, which rely on multiple-scattering theories, are comparatively fast, but less accurate than mixed algorithms, in which hard interactions (with energy loss or angular deflection larger than certain cut-off values) are simulated individually. The reliability, and limitations, of electron-interaction models and multiple-scattering theories will be analyzed. Benchmark comparisons of simu...
International Nuclear Information System (INIS)
Mann, F.M.
1998-01-01
The Tank Waste Remediation System (TWRS) is responsible for the safe storage, retrieval, and disposal of waste currently being held in 177 underground tanks at the Hanford Site. In order to successfully carry out its mission, TWRS must perform environmental analyses describing the consequences of tank contents leaking from tanks and associated facilities during the storage, retrieval, or closure periods and immobilized low-activity tank waste contaminants leaving disposal facilities. Because of the large size of the facilities and the great depth of the dry zone (known as the vadose zone) underneath the facilities, sophisticated computer codes are needed to model the transport of the tank contents or contaminants. This document presents the code selection criteria for those vadose zone analyses (a subset of the above analyses) where the hydraulic properties of the vadose zone are constant in time the geochemical behavior of the contaminant-soil interaction can be described by simple models, and the geologic or engineered structures are complicated enough to require a two-or three dimensional model. Thus, simple analyses would not need to use the fairly sophisticated codes which would meet the selection criteria in this document. Similarly, those analyses which involve complex chemical modeling (such as those analyses involving large tank leaks or those analyses involving the modeling of contaminant release from glass waste forms) are excluded. The analyses covered here are those where the movement of contaminants can be relatively simply calculated from the moisture flow. These code selection criteria are based on the information from the low-level waste programs of the US Department of Energy (DOE) and of the US Nuclear Regulatory Commission as well as experience gained in the DOE Complex in applying these criteria. Appendix table A-1 provides a comparison between the criteria in these documents and those used here. This document does not define the models (that
International Nuclear Information System (INIS)
Viswanathan, H.S.
1995-01-01
The finite element code FEHMN is a three-dimensional finite element heat and mass transport simulator that can handle complex stratigraphy and nonlinear processes such as vadose zone flow, heat flow and solute transport. Scientists at LANL have been developed hydrologic flow and transport models of the Yucca Mountain site using FEHMN. Previous FEHMN simulations have used an equivalent K d model to model solute transport. In this thesis, FEHMN is modified making it possible to simulate the transport of a species with a rigorous chemical model. Including the rigorous chemical equations into FEHMN simulations should provide for more representative transport models for highly reactive chemical species. A fully kinetic formulation is chosen for the FEHMN reactive transport model. Several methods are available to computationally implement a fully kinetic formulation. Different numerical algorithms are investigated in order to optimize computational efficiency and memory requirements of the reactive transport model. The best algorithm of those investigated is then incorporated into FEHMN. The algorithm chosen requires for the user to place strongly coupled species into groups which are then solved for simultaneously using FEHMN. The complete reactive transport model is verified over a wide variety of problems and is shown to be working properly. The simulations demonstrate that gas flow and carbonate chemistry can significantly affect 14 C transport at Yucca Mountain. The simulations also provide that the new capabilities of FEHMN can be used to refine and buttress already existing Yucca Mountain radionuclide transport studies
International Nuclear Information System (INIS)
Biwer, B.M.; LePoire, D.J.; Chen, S.Y.
1996-01-01
The RISKIND computer program was developed for the analysis of radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel (SNF) or other radioactive materials. The code is intended to provide scenario-specific analyses when evaluating alternatives for environmental assessment activities, including those for major federal actions involving radioactive material transport as required by the National Environmental Policy Act (NEPA). As such, rigorous procedures have been implemented to enhance the code's credibility and strenuous efforts have been made to enhance ease of use of the code. To increase the code's reliability and credibility, a new version of RISKIND was produced under a quality assurance plan that covered code development and testing, and a peer review process was conducted. During development of the new version, the flexibility and ease of use of RISKIND were enhanced through several major changes: (1) a Windows trademark point-and-click interface replaced the old DOS menu system, (2) the remaining model input parameters were added to the interface, (3) databases were updated, (4) the program output was revised, and (5) on-line help has been added. RISKIND has been well received by users and has been established as a key component in radiological transportation risk assessments through its acceptance by the U.S. Department of Energy community in recent environmental impact statements (EISs) and its continued use in the current preparation of several EISs
The neutron transport code DTF-Traca users manual and input data
International Nuclear Information System (INIS)
Ahnert, C.
1979-01-01
This is a users manual of the neutron transport code DTF-TRACA, which is a version of the original DTF-IV with some modifications made at JEN. A detailed input data descriptions is given. The new options developed at JEN are included too. (Author) 18 refs
The neutron transport code DTF-Traca users manual and input data
Energy Technology Data Exchange (ETDEWEB)
Ahnert, C.
1979-07-01
This is a users manual of the neutron transport code DTF-TRACA, which is a version of the original DTF-IV with some modifications made at JEN. A detailed input data descriptions is given. The new options developed at JEN are included too. (Author) 18 refs.
TRIDENT-CTR: a two-dimensional transport code for CTR applications
International Nuclear Information System (INIS)
Seed, T.J.
1978-01-01
TRIDENT-CTR is a two-dimensional x-y and r-z geometry multigroup neutral transport code developed at Los Alamos for toroidal calculations. The use of triangular finite elements gives it the geometric flexibility to cope with the nonorthogonal shapes of many toroidal designs of current interest in the CTR community
SQA of finite element method (FEM) codes used for analyses of pit storage/transport packages
Energy Technology Data Exchange (ETDEWEB)
Russel, E. [Lawrence Livermore National Lab., CA (United States)
1997-11-01
This report contains viewgraphs on the software quality assurance of finite element method codes used for analyses of pit storage and transport projects. This methodology utilizes the ISO 9000-3: Guideline for application of 9001 to the development, supply, and maintenance of software, for establishing well-defined software engineering processes to consistently maintain high quality management approaches.
International Nuclear Information System (INIS)
Zazula, J.M.
1983-01-01
The general purpose code BALTORO was written for coupling the three-dimensional Monte-Carlo /MC/ with the one-dimensional Discrete Ordinates /DO/ radiation transport calculations. The quantity of a radiation-induced /neutrons or gamma-rays/ nuclear effect or the score from a radiation-yielding nuclear effect can be analysed in this way. (author)
BOXER3: a three dimensional integral transport code for PHWR supercell
International Nuclear Information System (INIS)
Degweker, S.B.
1985-01-01
This report describes BOXER3, three dimensional computer code for solving the integral transport equation. The code uses a combination of the collision probability and the interface current methods. It uses mixed rectangular and cylinderical coordinates and can therefore treat cylindrical fuel channels and reactivity devices within a rectangular 'supercell' of a Candu PHWR. The report describes the details of computation of collision probabilities and the solution of the neutron balance equations. The latter can be done iteratively or by direct matrix inversion. It is shown that the iteration scheme is convergent. Comparisons of the results of BOXER3 and those obtained by other transport and diffusion codes in one, two and three dimensional geometries are also presented. (author)
A computer code PACTOLE to predict activation and transport of corrosion products in a PWR
International Nuclear Information System (INIS)
Beslu, P.; Frejaville, G.; Lalet, A.
1978-01-01
Theoretical studies on activation and transport of corrosion products in a PWR primary circuit have been concentrated, at CEA on the development of a computer code : PACTOLE. This code takes into account the major phenomena which govern corrosion products transport: 1. Ion solubility is obtained by usual thermodynamics laws in function of water chemistry: pH at operating temperature is calculated by the code. 2. Release rates of base metals, dissolution rates of deposits, precipitation rates of soluble products are derived from solubility variations. 3. Deposition of solid particles is treated by a model taking into account particle size, brownian and turbulent diffusion and inertial effect. Erosion of deposits is accounted for by a semi-empirical model. After a review of calculational models, an application of PACTOLE is presented in view of analyzing the distribution of in core. (author)
Comparative study of boron transport models in NRC Thermal-Hydraulic Code Trace
Energy Technology Data Exchange (ETDEWEB)
Olmo-Juan, Nicolás; Barrachina, Teresa; Miró, Rafael; Verdú, Gumersindo; Pereira, Claubia, E-mail: nioljua@iqn.upv.es, E-mail: tbarrachina@iqn.upv.es, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es, E-mail: claubia@nuclear.ufmg.br [Institute for Industrial, Radiophysical and Environmental Safety (ISIRYM). Universitat Politècnica de València (Spain); Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear
2017-07-01
Recently, the interest in the study of various types of transients involving changes in the boron concentration inside the reactor, has led to an increase in the interest of developing and studying new models and tools that allow a correct study of boron transport. Therefore, a significant variety of different boron transport models and spatial difference schemes are available in the thermal-hydraulic codes, as TRACE. According to this interest, in this work it will be compared the results obtained using the different boron transport models implemented in the NRC thermal-hydraulic code TRACE. To do this, a set of models have been created using the different options and configurations that could have influence in boron transport. These models allow to reproduce a simple event of filling or emptying the boron concentration in a long pipe. Moreover, with the aim to compare the differences obtained when one-dimensional or three-dimensional components are chosen, it has modeled many different cases using only pipe components or a mix of pipe and vessel components. In addition, the influence of the void fraction in the boron transport has been studied and compared under close conditions to BWR commercial model. A final collection of the different cases and boron transport models are compared between them and those corresponding to the analytical solution provided by the Burgers equation. From this comparison, important conclusions are drawn that will be the basis of modeling the boron transport in TRACE adequately. (author)
Validation of the transportation computer codes HIGHWAY, INTERLINE, RADTRAN 4, and RISKIND
Energy Technology Data Exchange (ETDEWEB)
Maheras, S.J.; Pippen, H.K.
1995-05-01
The computer codes HIGHWAY, INTERLINE, RADTRAN 4, and RISKIND were used to estimate radiation doses from the transportation of radioactive material in the Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Environmental Impact Statement. HIGHWAY and INTERLINE were used to estimate transportation routes for truck and rail shipments, respectively. RADTRAN 4 was used to estimate collective doses from incident-free transportation and the risk (probability {times} consequence) from transportation accidents. RISKIND was used to estimate incident-free radiation doses for maximally exposed individuals and the consequences from reasonably foreseeable transportation accidents. The purpose of this analysis is to validate the estimates made by these computer codes; critiques of the conceptual models used in RADTRAN 4 are also discussed. Validation is defined as ``the test and evaluation of the completed software to ensure compliance with software requirements.`` In this analysis, validation means that the differences between the estimates generated by these codes and independent observations are small (i.e., within the acceptance criterion established for the validation analysis). In some cases, the independent observations used in the validation were measurements; in other cases, the independent observations used in the validation analysis were generated using hand calculations. The results of the validation analyses performed for HIGHWAY, INTERLINE, RADTRAN 4, and RISKIND show that the differences between the estimates generated using the computer codes and independent observations were small. Based on the acceptance criterion established for the validation analyses, the codes yielded acceptable results; in all cases the estimates met the requirements for successful validation.
Validation of the transportation computer codes HIGHWAY, INTERLINE, RADTRAN 4, and RISKIND
International Nuclear Information System (INIS)
Maheras, S.J.; Pippen, H.K.
1995-05-01
The computer codes HIGHWAY, INTERLINE, RADTRAN 4, and RISKIND were used to estimate radiation doses from the transportation of radioactive material in the Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Environmental Impact Statement. HIGHWAY and INTERLINE were used to estimate transportation routes for truck and rail shipments, respectively. RADTRAN 4 was used to estimate collective doses from incident-free transportation and the risk (probability x consequence) from transportation accidents. RISKIND was used to estimate incident-free radiation doses for maximally exposed individuals and the consequences from reasonably foreseeable transportation accidents. The purpose of this analysis is to validate the estimates made by these computer codes; critiques of the conceptual models used in RADTRAN 4 are also discussed. Validation is defined as ''the test and evaluation of the completed software to ensure compliance with software requirements.'' In this analysis, validation means that the differences between the estimates generated by these codes and independent observations are small (i.e., within the acceptance criterion established for the validation analysis). In some cases, the independent observations used in the validation were measurements; in other cases, the independent observations used in the validation analysis were generated using hand calculations. The results of the validation analyses performed for HIGHWAY, INTERLINE, RADTRAN 4, and RISKIND show that the differences between the estimates generated using the computer codes and independent observations were small. Based on the acceptance criterion established for the validation analyses, the codes yielded acceptable results; in all cases the estimates met the requirements for successful validation
International Nuclear Information System (INIS)
King, C.M.; Wilhite, E.L.; Root, R.W. Jr.
1985-01-01
The Savannah River Laboratory DOSTOMAN code has been used since 1978 for environmental pathway analysis of potential migration of radionuclides and hazardous chemicals. The DOSTOMAN work is reviewed including a summary of historical use of compartmental models, the mathematical basis for the DOSTOMAN code, examples of exact analytical solutions for simple matrices, methods for numerical solution of complex matrices, and mathematical validation/calibration of the SRL code. The review includes the methodology for application to nuclear and hazardous chemical waste disposal, examples of use of the model in contaminant transport and pathway analysis, a user's guide for computer implementation, peer review of the code, and use of DOSTOMAN at other Department of Energy sites. 22 refs., 3 figs
Open-Source Development of the Petascale Reactive Flow and Transport Code PFLOTRAN
Hammond, G. E.; Andre, B.; Bisht, G.; Johnson, T.; Karra, S.; Lichtner, P. C.; Mills, R. T.
2013-12-01
Open-source software development has become increasingly popular in recent years. Open-source encourages collaborative and transparent software development and promotes unlimited free redistribution of source code to the public. Open-source development is good for science as it reveals implementation details that are critical to scientific reproducibility, but generally excluded from journal publications. In addition, research funds that would have been spent on licensing fees can be redirected to code development that benefits more scientists. In 2006, the developers of PFLOTRAN open-sourced their code under the U.S. Department of Energy SciDAC-II program. Since that time, the code has gained popularity among code developers and users from around the world seeking to employ PFLOTRAN to simulate thermal, hydraulic, mechanical and biogeochemical processes in the Earth's surface/subsurface environment. PFLOTRAN is a massively-parallel subsurface reactive multiphase flow and transport simulator designed from the ground up to run efficiently on computing platforms ranging from the laptop to leadership-class supercomputers, all from a single code base. The code employs domain decomposition for parallelism and is founded upon the well-established and open-source parallel PETSc and HDF5 frameworks. PFLOTRAN leverages modern Fortran (i.e. Fortran 2003-2008) in its extensible object-oriented design. The use of this progressive, yet domain-friendly programming language has greatly facilitated collaboration in the code's software development. Over the past year, PFLOTRAN's top-level data structures were refactored as Fortran classes (i.e. extendible derived types) to improve the flexibility of the code, ease the addition of new process models, and enable coupling to external simulators. For instance, PFLOTRAN has been coupled to the parallel electrical resistivity tomography code E4D to enable hydrogeophysical inversion while the same code base can be used as a third
1987-02-15
82302 F 13211 PT VERDE WPB 82311 F 13212 PT SWIFT WPB 82312 E. 13214 PT THATCHER WPB 82314 E 13218 PT HERRON WPB 82318 C 13232 PT ROBERTS WPB 82332 E...Identifies DOT, FAA Logistica Center, OkIanhea City, as an organization to be billed. 4th Position Code A Ia assigned by DOT, rAA. Identifies appropriation
International Nuclear Information System (INIS)
Blakeman, E.D.
2000-01-01
A software system, GRAVE (Geometry Rendering and Visual Editor), has been developed at the Oak Ridge National Laboratory (ORNL) to perform interactive visualization and development of models used as input to the TORT three-dimensional discrete ordinates radiation transport code. Three-dimensional and two-dimensional visualization displays are included. Display capabilities include image rotation, zoom, translation, wire-frame and translucent display, geometry cuts and slices, and display of individual component bodies and material zones. The geometry can be interactively edited and saved in TORT input file format. This system is an advancement over the current, non-interactive, two-dimensional display software. GRAVE is programmed in the Java programming language and can be implemented on a variety of computer platforms. Three- dimensional visualization is enabled through the Visualization Toolkit (VTK), a free-ware C++ software library developed for geometric and data visual display. Future plans include an extension of the system to read inputs using binary zone maps and combinatorial geometry models containing curved surfaces, such as those used for Monte Carlo code inputs. Also GRAVE will be extended to geometry visualization/editing for the DORT two-dimensional transport code and will be integrated into a single GUI-based system for all of the ORNL discrete ordinates transport codes
Energy Technology Data Exchange (ETDEWEB)
Blakeman, E.D.
2000-05-07
A software system, GRAVE (Geometry Rendering and Visual Editor), has been developed at the Oak Ridge National Laboratory (ORNL) to perform interactive visualization and development of models used as input to the TORT three-dimensional discrete ordinates radiation transport code. Three-dimensional and two-dimensional visualization displays are included. Display capabilities include image rotation, zoom, translation, wire-frame and translucent display, geometry cuts and slices, and display of individual component bodies and material zones. The geometry can be interactively edited and saved in TORT input file format. This system is an advancement over the current, non-interactive, two-dimensional display software. GRAVE is programmed in the Java programming language and can be implemented on a variety of computer platforms. Three- dimensional visualization is enabled through the Visualization Toolkit (VTK), a free-ware C++ software library developed for geometric and data visual display. Future plans include an extension of the system to read inputs using binary zone maps and combinatorial geometry models containing curved surfaces, such as those used for Monte Carlo code inputs. Also GRAVE will be extended to geometry visualization/editing for the DORT two-dimensional transport code and will be integrated into a single GUI-based system for all of the ORNL discrete ordinates transport codes.
Application of the three-dimensional transport code to analysis of the neutron streaming experiment
International Nuclear Information System (INIS)
Chatani, K.; Slater, C.O.
1990-01-01
The neutron streaming through an experimental mock-up of a Clinch River Breeder Reactor (CRBR) prototypic coolant pipe chaseway was recalculated with a three-dimensional discrete ordinates code. The experiment was conducted at the Tower Shielding Facility at Oak Ridge National Laboratory in 1976 and 1977. The measurement of the neutron flux, using Bonner ball detectors, indicated nine orders of attenuation in the empty pipeway, which contained two 90-deg bends and was surrounded by concrete walls. The measurement data were originally analyzed using the DOT3.5 two-dimensional discrete ordinates radiation transport code. However, the results did not agree with measurement data at the bend because of the difficulties in modeling the three-dimensional configurations using two-dimensional methods. The two-dimensional calculations used a three-step procedure in which each of the three legs making the two 90-deg bends was a separate calculation. The experiment was recently analyzed with the TORT three-dimensional discrete ordinates radiation transport code, not only to compare the calculational results with the experimental results, but also to compare with results obtained from analyses in Japan using DOT3.5, MORSE, and ENSEMBLE, which is a three-dimensional discrete ordinates radiation transport code developed in Japan
A user's manual for the three-dimensional Monte Carlo transport code SPARTAN
International Nuclear Information System (INIS)
Bending, R.C.; Heffer, P.J.H.
1975-09-01
SPARTAN is a general-purpose Monte Carlo particle transport code intended for neutron or gamma transport problems in reactor physics, health physics, shielding, and safety studies. The code used a very general geometry system enabling a complex layout to be described and allows the user to obtain physics data from a number of different types of source library. Special tracking and scoring techniques are used to improve the quality of the results obtained. To enable users to run SPARTAN, brief descriptions of the facilities available in the code are given and full details of data input and job control language, as well as examples of complete calculations, are included. It is anticipated that changes may be made to SPARTAN from time to time, particularly in those parts of the code which deal with physics data processing. The load module is identified by a version number and implementation date, and updates of sections of this manual will be issued when significant changes are made to the code. (author)
PFLOTRAN: Reactive Flow & Transport Code for Use on Laptops to Leadership-Class Supercomputers
Energy Technology Data Exchange (ETDEWEB)
Hammond, Glenn E.; Lichtner, Peter C.; Lu, Chuan; Mills, Richard T.
2012-04-18
PFLOTRAN, a next-generation reactive flow and transport code for modeling subsurface processes, has been designed from the ground up to run efficiently on machines ranging from leadership-class supercomputers to laptops. Based on an object-oriented design, the code is easily extensible to incorporate additional processes. It can interface seamlessly with Fortran 9X, C and C++ codes. Domain decomposition parallelism is employed, with the PETSc parallel framework used to manage parallel solvers, data structures and communication. Features of the code include a modular input file, implementation of high-performance I/O using parallel HDF5, ability to perform multiple realization simulations with multiple processors per realization in a seamless manner, and multiple modes for multiphase flow and multicomponent geochemical transport. Chemical reactions currently implemented in the code include homogeneous aqueous complexing reactions and heterogeneous mineral precipitation/dissolution, ion exchange, surface complexation and a multirate kinetic sorption model. PFLOTRAN has demonstrated petascale performance using 2{sup 17} processor cores with over 2 billion degrees of freedom. Accomplishments achieved to date include applications to the Hanford 300 Area and modeling CO{sub 2} sequestration in deep geologic formations.
MCNP: a general Monte Carlo code for neutron and photon transport
Energy Technology Data Exchange (ETDEWEB)
Forster, R.A.; Godfrey, T.N.K.
1985-01-01
MCNP is a very general Monte Carlo neutron photon transport code system with approximately 250 person years of Group X-6 code development invested. It is extremely portable, user-oriented, and a true production code as it is used about 60 Cray hours per month by about 150 Los Alamos users. It has as its data base the best cross-section evaluations available. MCNP contains state-of-the-art traditional and adaptive Monte Carlo techniques to be applied to the solution of an ever-increasing number of problems. Excellent user-oriented documentation is available for all facets of the MCNP code system. Many useful and important variants of MCNP exist for special applications. The Radiation Shielding Information Center (RSIC) in Oak Ridge, Tennessee is the contact point for worldwide MCNP code and documentation distribution. A much improved MCNP Version 3A will be available in the fall of 1985, along with new and improved documentation. Future directions in MCNP development will change the meaning of MCNP to Monte Carlo N Particle where N particle varieties will be transported.
Energy Technology Data Exchange (ETDEWEB)
Ramsdell, J.V. Jr.; Simonen, C.A.; Burk, K.W.
1994-02-01
The purpose of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate radiation doses that individuals may have received from operations at the Hanford Site since 1944. This report deals specifically with the atmospheric transport model, Regional Atmospheric Transport Code for Hanford Emission Tracking (RATCHET). RATCHET is a major rework of the MESOILT2 model used in the first phase of the HEDR Project; only the bookkeeping framework escaped major changes. Changes to the code include (1) significant changes in the representation of atmospheric processes and (2) incorporation of Monte Carlo methods for representing uncertainty in input data, model parameters, and coefficients. To a large extent, the revisions to the model are based on recommendations of a peer working group that met in March 1991. Technical bases for other portions of the atmospheric transport model are addressed in two other documents. This report has three major sections: a description of the model, a user`s guide, and a programmer`s guide. These sections discuss RATCHET from three different perspectives. The first provides a technical description of the code with emphasis on details such as the representation of the model domain, the data required by the model, and the equations used to make the model calculations. The technical description is followed by a user`s guide to the model with emphasis on running the code. The user`s guide contains information about the model input and output. The third section is a programmer`s guide to the code. It discusses the hardware and software required to run the code. The programmer`s guide also discusses program structure and each of the program elements.
International Nuclear Information System (INIS)
De Matteis, A.
1987-01-01
This report describes the fully automatic linkage between the finite difference, two-dimensional code EDGE2D, based on the classical Braginskii partial differential equations of ion transport, and the Monte Carlo code NIMBUS, which solves the integral form of the stationary, linear Boltzmann equation for neutral transport in a plasma. The coupling has been performed for the real poloidal geometry of JET with two belt-limiters and real magnetic configurations with or without a single-null point. The new integrated system starts from the magnetic geometry computed by predictive or interpretative equilibrium codes and yields the plasma and neutrals characteristics in the edge
ETRANS: an energy transport system optimization code for distributed networks of solar collectors
Energy Technology Data Exchange (ETDEWEB)
Barnhart, J.S.
1980-09-01
The optimization code ETRANS was developed at the Pacific Northwest Laboratory to design and estimate the costs associated with energy transport systems for distributed fields of solar collectors. The code uses frequently cited layouts for dish and trough collectors and optimizes them on a section-by-section basis. The optimal section design is that combination of pipe diameter and insulation thickness that yields the minimum annualized system-resultant cost. Among the quantities included in the costing algorithm are (1) labor and materials costs associated with initial plant construction, (2) operating expenses due to daytime and nighttime heat losses, and (3) operating expenses due to pumping power requirements. Two preliminary series of simulations were conducted to exercise the code. The results indicate that transport system costs for both dish and trough collector fields increase with field size and receiver exit temperature. Furthermore, dish collector transport systems were found to be much more expensive to build and operate than trough transport systems. ETRANS itself is stable and fast-running and shows promise of being a highly effective tool for the analysis of distributed solar thermal systems.
International Nuclear Information System (INIS)
Viswanathan, H.S.
1996-08-01
The finite element code FEHMN, developed by scientists at Los Alamos National Laboratory (LANL), is a three-dimensional finite element heat and mass transport simulator that can handle complex stratigraphy and nonlinear processes such as vadose zone flow, heat flow and solute transport. Scientists at LANL have been developing hydrologic flow and transport models of the Yucca Mountain site using FEHMN. Previous FEHMN simulations have used an equivalent Kd model to model solute transport. In this thesis, FEHMN is modified making it possible to simulate the transport of a species with a rigorous chemical model. Including the rigorous chemical equations into FEHMN simulations should provide for more representative transport models for highly reactive chemical species. A fully kinetic formulation is chosen for the FEHMN reactive transport model. Several methods are available to computationally implement a fully kinetic formulation. Different numerical algorithms are investigated in order to optimize computational efficiency and memory requirements of the reactive transport model. The best algorithm of those investigated is then incorporated into FEHMN. The algorithm chosen requires for the user to place strongly coupled species into groups which are then solved for simultaneously using FEHMN. The complete reactive transport model is verified over a wide variety of problems and is shown to be working properly. The new chemical capabilities of FEHMN are illustrated by using Los Alamos National Laboratory's site scale model of Yucca Mountain to model two-dimensional, vadose zone 14 C transport. The simulations demonstrate that gas flow and carbonate chemistry can significantly affect 14 C transport at Yucca Mountain. The simulations also prove that the new capabilities of FEHMN can be used to refine and buttress already existing Yucca Mountain radionuclide transport studies
The OpenMOC method of characteristics neutral particle transport code
International Nuclear Information System (INIS)
Boyd, William; Shaner, Samuel; Li, Lulu; Forget, Benoit; Smith, Kord
2014-01-01
Highlights: • An open source method of characteristics neutron transport code has been developed. • OpenMOC shows nearly perfect scaling on CPUs and 30× speedup on GPUs. • Nonlinear acceleration techniques demonstrate a 40× reduction in source iterations. • OpenMOC uses modern software design principles within a C++ and Python framework. • Validation with respect to the C5G7 and LRA benchmarks is presented. - Abstract: The method of characteristics (MOC) is a numerical integration technique for partial differential equations, and has seen widespread use for reactor physics lattice calculations. The exponential growth in computing power has finally brought the possibility for high-fidelity full core MOC calculations within reach. The OpenMOC code is being developed at the Massachusetts Institute of Technology to investigate algorithmic acceleration techniques and parallel algorithms for MOC. OpenMOC is a free, open source code written using modern software languages such as C/C++ and CUDA with an emphasis on extensible design principles for code developers and an easy to use Python interface for code users. The present work describes the OpenMOC code and illustrates its ability to model large problems accurately and efficiently
Presentation and use of a reactive transport code in porous media
Montarnal, Ph.; Mügler, C.; Colin, J.; Descostes, M.; Dimier, A.; Jacquot, E.
The safety assessment of nuclear waste disposals requires an accurate prediction of the radionuclides and chemical species migration through engineered barriers and geological media. It is therefore necessary to develop and assess qualified and validated tools which integrate both the transport mechanisms through the geological media and the chemical mechanisms governing the mobility of radionuclides. Such a reactive transport simulation tool has been developed in the context of the numerical software platform ALLIANCES. Different component codes are available: PHREEQC and CHESS for the chemical part, CAST3M, MT3D and TRACES for the transport part. A coupling scheme has already been implemented, qualified and validated on numerous configurations involving aqueous speciation, dissolution-precipitation, sorption and surface complexation. Presently, the reactive transport numerical tool is used to simulate realistic configurations. This paper presents two of such applications: the migration of uranium in a soil with various redox conditions and the modelling of clay-cement interactions.
International Nuclear Information System (INIS)
Mori, Takamasa; Nakagawa, Masayuki; Sasaki, Makoto.
1988-11-01
We have developed a group of computer codes to realize the accurate transport calculation by using the multi-group double-differential form cross section. This type of cross section can correctly take account of the energy-angle correlated reaction kinematics. Accordingly, the transport phenomena in materials with highly anisotropic scattering are accurately calculated by using this cross section. They include the following four codes or code systems: PROF-DD : a code system to generate the multi-group double-differential form cross section library by processing basic nuclear data file compiled in the ENDF / B-IV or -V format, ANISN-DD : a one-dimensional transport code based on the discrete ordinate method, DOT-DD : a two-dimensional transport code based on the discrete ordinate method, MORSE-DD : a three-dimensional transport code based on the Monte Carlo method. In addition to these codes, several auxiliary codes have been developed to process calculated results. This report describes the calculation algorithm employed in these codes and how to use them. (author)
Development and preliminary verification of 2-D transport module of radiation shielding code ARES
International Nuclear Information System (INIS)
Zhang Penghe; Chen Yixue; Zhang Bin; Zang Qiyong; Yuan Longjun; Chen Mengteng
2013-01-01
The 2-D transport module of radiation shielding code ARES is two-dimensional neutron and radiation shielding code. The theory model was based on the first-order steady state neutron transport equation, adopting the discrete ordinates method to disperse direction variables. Then a set of differential equations can be obtained and solved with the source iteration method. The 2-D transport module of ARES was capable of calculating k eff and fixed source problem with isotropic or anisotropic scattering in x-y geometry. The theoretical model was briefly introduced and series of benchmark problems were verified in this paper. Compared with the results given by the benchmark, the maximum relative deviation of k eff is 0.09% and the average relative deviation of flux density is about 0.60% in the BWR cells benchmark problem. As for the fixed source problem with isotropic and anisotropic scattering, the results of the 2-D transport module of ARES conform with DORT very well. These numerical results of benchmark problems preliminarily demonstrate that the development process of the 2-D transport module of ARES is right and it is able to provide high precision result. (authors)
Energy Conservation Tests of a Coupled Kinetic-kinetic Plasma-neutral Transport Code
Energy Technology Data Exchange (ETDEWEB)
Stotler, D. P.; Chang, C. S.; Ku, S. H.; Lang, J.; Park, G.
2012-08-29
A Monte Carlo neutral transport routine, based on DEGAS2, has been coupled to the guiding center ion-electron-neutral neoclassical PIC code XGC0 to provide a realistic treatment of neutral atoms and molecules in the tokamak edge plasma. The DEGAS2 routine allows detailed atomic physics and plasma-material interaction processes to be incorporated into these simulations. The spatial pro le of the neutral particle source used in the DEGAS2 routine is determined from the uxes of XGC0 ions to the material surfaces. The kinetic-kinetic plasma-neutral transport capability is demonstrated with example pedestal fueling simulations.
International Nuclear Information System (INIS)
Raske, D.T.; Wang, Z.
1992-01-01
The primary concern governing the design of shipping packages containing radioactive materials is public safety during transport. When these shipments are within the regulatory jurisdiction of the US Department of Energy, the recommended design criterion for the primary containment vessel is either Section III or Section VIII, Division 1, of the ASME Boiler and Pressure Vessel Code, depending on the activity of the contents. The objective of this paper is to discuss the design of a prototypic containment vessel representative of a packaging for the transport of high-level radioactive material
Draft ASME code case on ductile cast iron for transport packaging
Energy Technology Data Exchange (ETDEWEB)
Saegusa, T. [Central Research Inst. of Electric Power Industry, Abiko (Japan); Arai, T. [Central Research Inst. of Electric Power Industry, Yokosuka (Japan); Hirose, M. [Nuclear Fuel Transport Co., Ltd., Tokyo (Japan); Kobayashi, T. [Nippon Chuzo, Kawasaki (Japan); Tezuka, Y. [Mitsubishi Materials Co., Tokyo (Japan); Urabe, N. [Kokan Keisoku K. K., Kawasaki (Japan); Hueggenberg, R. [GNB, Essen (Germany)
2004-07-01
The current Rules for Construction of ''Containment Systems for Storage and Transport Packagings of Spent Nuclear Fuel and High Level Radioactive Material and Waste'' of Division 3 in Section III of ASME Code (2001 Edition) does not include ductile cast iron in its list of materials permitted for use. The Rules specify required fracture toughness values of ferritic steel material for nominal wall thickness 5/8 to 12 inches (16 to 305 mm). New rule for ductile cast iron for transport packaging of which wall thickness is greater than 12 inches (305mm) is required.
The Development of 3D Graphics for Simple Implementation of Photon and Neutron Transport Code
International Nuclear Information System (INIS)
Siangsanan, P.
2014-01-01
The Simple Implementation of Photon and Neutron Transport code (SIPHON) was developed and tested at Office of Atoms for Peace around 1998 using nuclear data from MCNP code. The input of SIPHON is in the form of text file so that user could set the dimension of simulation model with accuracy. Whereas the code can check the correctness of geometry of the model during running time, the point of error will be found only if a simulated particle has crossed the erratic geometry and might take a lot of time to be found in a very complex system. The three-dimensional graphical view was implemented into SIPHON to solve this problem and was found later that it is also useful in educational purpose.
BRYNTRN: A baryon transport computer code, computation procedures and data base
Wilson, John W.; Townsend, Lawrence W.; Chun, Sang Y.; Buck, Warren W.; Khan, Ferdous; Cucinotta, Frank
1988-01-01
The development is described of an interaction data base and a numerical solution to the transport of baryons through the arbitrary shield material based on a straight ahead approximation of the Boltzmann equation. The code is most accurate for continuous energy boundary values but gives reasonable results for discrete spectra at the boundary with even a relatively coarse energy grid (30 points) and large spatial increments (1 cm in H2O).
RIVER-RAD: A computer code for simulating the transport of radionuclides in rivers
Energy Technology Data Exchange (ETDEWEB)
Hetrick, D.M.; McDowell-Boyer, L.M.; Sjoreen, A.L.; Thorne, D.J.; Patterson, M.R.
1992-11-01
A screening-level model, RIVER-RAD, has been developed to assess the potential fate of radionuclides released to rivers. The model is simplified in nature and is intended to provide guidance in determining the potential importance of the surface water pathway, relevant transport mechanisms, and key radionuclides in estimating radiological dose to man. The purpose of this report is to provide a description of the model and a user's manual for the FORTRAN computer code.
Heavy-ion transport codes for radiotherapy and radioprotection in space
Energy Technology Data Exchange (ETDEWEB)
Mancusi, Davide
2006-06-15
Simulation of the transport of heavy ions in matter is a field of nuclear science that has recently received attention in view of its importance for some relevant applications. Accelerated heavy ions can, for example, be used to treat cancers (heavy-ion radiotherapy) and show some superior qualities with respect to more conventional treatment systems, like photons (x-rays) or protons. Furthermore, long-term manned space missions (like a possible future mission to Mars) pose the challenge to protect astronauts and equipment on board against the harmful space radiation environment, where heavy ions can be responsible for a significant share of the exposure risk. The high accuracy expected from a transport algorithm (especially in the case of radiotherapy) and the large amount of semi-empirical knowledge necessary to even state the transport problem properly rule out any analytical approach; the alternative is to resort to numerical simulations in order to build treatment-planning systems for cancer or to aid space engineers in shielding design. This thesis is focused on the description of HIBRAC, a one-dimensional deterministic code optimised for radiotherapy, and PHITS (Particle and Heavy- Ion Transport System), a general-purpose three-dimensional Monte-Carlo code. The structure of both codes is outlined and some relevant results are presented. In the case of PHITS, we also report the first results of an ongoing comprehensive benchmarking program for the main components of the code; we present the comparison of partial charge-changing cross sections for a 400 MeV/n {sup 40}Ar beam impinging on carbon, polyethylene, aluminium, copper, tin and lead targets.
International Nuclear Information System (INIS)
Parks, C.V.; Broadhead, B.L.; Hermann, O.W.; Tang, J.S.; Cramer, S.N.; Gauthey, J.C.; Kirk, B.L.; Roussin, R.W.
1988-07-01
This report provides a preliminary assessment of the computational tools and existing methods used to obtain radiation dose rates from shielded spent nuclear fuel and high-level radioactive waste (HLW). Particular emphasis is placed on analysis tools and techniques applicable to facilities/equipment designed for the transport or storage of spent nuclear fuel or HLW. Applications to cask transport, storage, and facility handling are considered. The report reviews the analytic techniques for generating appropriate radiation sources, evaluating the radiation transport through the shield, and calculating the dose at a desired point or surface exterior to the shield. Discrete ordinates, Monte Carlo, and point kernel methods for evaluating radiation transport are reviewed, along with existing codes and data that utilize these methods. A literature survey was employed to select a cadre of codes and data libraries to be reviewed. The selection process was based on specific criteria presented in the report. Separate summaries were written for several codes (or family of codes) that provided information on the method of solution, limitations and advantages, availability, data access, ease of use, and known accuracy. For each data library, the summary covers the source of the data, applicability of these data, and known verification efforts. Finally, the report discusses the overall status of spent fuel shielding analysis techniques and attempts to illustrate areas where inaccuracy and/or uncertainty exist. The report notes the advantages and limitations of several analysis procedures and illustrates the importance of using adequate cross-section data sets. Additional work is recommended to enable final selection/validation of analysis tools that will best meet the US Department of Energy's requirements for use in developing a viable HLW management system. 188 refs., 16 figs., 27 tabs
The three-dimensional, discrete ordinates neutral particle transport code TORT: An overview
International Nuclear Information System (INIS)
Azmy, Y.Y.
1996-01-01
The centerpiece of the Discrete Ordinates Oak Ridge System (DOORS), the three-dimensional neutral particle transport code TORT is reviewed. Its most prominent features pertaining to large applications, such as adjustable problem parameters, memory management, and coarse mesh methods, are described. Advanced, state-of-the-art capabilities including acceleration and multiprocessing are summarized here. Future enhancement of existing graphics and visualization tools is briefly presented
Heavy-ion transport codes for radiotherapy and radioprotection in space
International Nuclear Information System (INIS)
Mancusi, Davide
2006-06-01
Simulation of the transport of heavy ions in matter is a field of nuclear science that has recently received attention in view of its importance for some relevant applications. Accelerated heavy ions can, for example, be used to treat cancers (heavy-ion radiotherapy) and show some superior qualities with respect to more conventional treatment systems, like photons (x-rays) or protons. Furthermore, long-term manned space missions (like a possible future mission to Mars) pose the challenge to protect astronauts and equipment on board against the harmful space radiation environment, where heavy ions can be responsible for a significant share of the exposure risk. The high accuracy expected from a transport algorithm (especially in the case of radiotherapy) and the large amount of semi-empirical knowledge necessary to even state the transport problem properly rule out any analytical approach; the alternative is to resort to numerical simulations in order to build treatment-planning systems for cancer or to aid space engineers in shielding design. This thesis is focused on the description of HIBRAC, a one-dimensional deterministic code optimised for radiotherapy, and PHITS (Particle and Heavy- Ion Transport System), a general-purpose three-dimensional Monte-Carlo code. The structure of both codes is outlined and some relevant results are presented. In the case of PHITS, we also report the first results of an ongoing comprehensive benchmarking program for the main components of the code; we present the comparison of partial charge-changing cross sections for a 400 MeV/n 40 Ar beam impinging on carbon, polyethylene, aluminium, copper, tin and lead targets
Development of computational two-phase flow analysis code with interfacial area transport equation
International Nuclear Information System (INIS)
Bae, B.U.; Park, G.C.; Yoon, H.Y.; Euh, D.J.; Song, C.H.
2007-01-01
In the two-phase flow analysis with two-fluid model, interfacial area concentration (IAC) is a dominant factor governing the interfacial transfer of momentum and energy. In order to overcome the shortcomings of experimental correlation for IAC, such as the dependency on the flow regime, multi-dimensional computational fluid dynamics (CFD) code was developed with the interfacial area transport equation. The code is based on two-fluid model and simplified marker and cell (SMAC) algorithm using the finite volume method, and the conventional approach in single-phase flow has been modified in order to consider the term of phase change. Also, instead of a static one-dimensional correlation for IAC, the code adopted the one-group interfacial area transport equation which includes source terms with respect to the coalescence and breakup of bubbles, and the phase change such as evaporation or condensation. As benchmark problems of single-phase flow and two-phase flow, the natural convection in rectangular cavity and the subcooled boiling in vertical annulus channel were analyzed, respectively. In the calculation for single-phase flow, the developed code predicted reasonable behavior of buoyancy-driven flow depending on Rayleigh number, so that the robustness in calculation capability of each phase has been confirmed. In the analysis for the subcooled boiling experiment performed in Seoul National University, the calculation results represented the reasonable capability in predicting the multi-dimensional phenomena such as vapor generation and void propagation. (authors)
Development of computational two-phase flow analysis code with interfacial area transport equation
Energy Technology Data Exchange (ETDEWEB)
Bae, B.U.; Park, G.C. [Seoul National Univ., Dept. of Nuclear Engineering (Korea, Republic of); Yoon, H.Y.; Euh, D.J.; Song, C.H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2007-07-01
In the two-phase flow analysis with two-fluid model, interfacial area concentration (IAC) is a dominant factor governing the interfacial transfer of momentum and energy. In order to overcome the shortcomings of experimental correlation for IAC, such as the dependency on the flow regime, multi-dimensional computational fluid dynamics (CFD) code was developed with the interfacial area transport equation. The code is based on two-fluid model and simplified marker and cell (SMAC) algorithm using the finite volume method, and the conventional approach in single-phase flow has been modified in order to consider the term of phase change. Also, instead of a static one-dimensional correlation for IAC, the code adopted the one-group interfacial area transport equation which includes source terms with respect to the coalescence and breakup of bubbles, and the phase change such as evaporation or condensation. As benchmark problems of single-phase flow and two-phase flow, the natural convection in rectangular cavity and the subcooled boiling in vertical annulus channel were analyzed, respectively. In the calculation for single-phase flow, the developed code predicted reasonable behavior of buoyancy-driven flow depending on Rayleigh number, so that the robustness in calculation capability of each phase has been confirmed. In the analysis for the subcooled boiling experiment performed in Seoul National University, the calculation results represented the reasonable capability in predicting the multi-dimensional phenomena such as vapor generation and void propagation. (authors)
Directory of Open Access Journals (Sweden)
Yixue Chen
2017-01-01
Full Text Available ARES is a multidimensional parallel discrete ordinates particle transport code with arbitrary order anisotropic scattering. It can be applied to a wide variety of radiation shielding calculations and reactor physics analysis. ARES uses state-of-the-art solution methods to obtain accurate solutions to the linear Boltzmann transport equation. A multigroup discretization is applied in energy. The code allows multiple spatial discretization schemes and solution methodologies. ARES currently provides diamond difference with or without linear-zero flux fixup, theta weighted, directional theta weighted, exponential directional weighted, and linear discontinuous finite element spatial differencing schemes. Discrete ordinates differencing in angle and spherical harmonics expansion of the scattering source are adopted. First collision source method is used to eliminate or mitigate the ray effects. Traditional source iteration and Krylov iterative method preconditioned with diffusion synthetic acceleration are applied to solve the linear system of equations. ARES uses the Koch-Baker-Alcouffe parallel sweep algorithm to obtain high parallel efficiency. Verification and validation for the ARES transport code system have been done by lots of benchmarks. In this paper, ARES solutions to the HBR-2 benchmark and C5G7 benchmarks are in excellent agreement with published results. Numerical results are presented which demonstrate the accuracy and efficiency of these methods.
PHITS: Particle and heavy ion transport code system, version 2.23
International Nuclear Information System (INIS)
Niita, Koji; Matsuda, Norihiro; Iwamoto, Yosuke; Sato, Tatsuhiko; Nakashima, Hiroshi; Sakamoto, Yukio; Iwase, Hiroshi; Sihver, Lembit
2010-10-01
A Particle and Heavy-Ion Transport code System PHITS has been developed under the collaboration of JAEA (Japan Atomic Energy Agency), RIST (Research Organization for Information Science and Technology) and KEK (High Energy Accelerator Research Organization). PHITS can deal with the transport of all particles (nucleons, nuclei, mesons, photons, and electrons) over wide energy ranges, using several nuclear reaction models and nuclear data libraries. Geometrical configuration of the simulation can be set with GG (General Geometry) or CG (Combinatorial Geometry). Various quantities such as heat deposition, track length and production yields can be deduced from the simulation, using implemented estimator functions called 'tally'. The code also has a function to draw 2D and 3D figures of the calculated results as well as the setup geometries, using a code ANGEL. Because of these features, PHITS has been widely used for various purposes such as designs of accelerator shielding, radiation therapy and space exploration. Recently PHITS introduces an event generator for particle transport parts in the low energy region. Thus, PHITS was completely rewritten for the introduction of the event generator for neutron-induced reactions in energy region less than 20 MeV. Furthermore, several new tallis were incorporated for estimation of the relative biological effects. This document provides a manual of the new PHITS. (author)
Energy Technology Data Exchange (ETDEWEB)
Coste-Delclaux, M
2006-03-15
This document describes the improvements carried out for modelling the self-shielding phenomenon in the multigroup transport code APOLLO2. They concern the space and energy treatment of the slowing-down equation, the setting up of quadrature formulas to calculate reaction rates, the setting-up of a method that treats directly a resonant mixture and the development of a sub-group method. We validate these improvements either in an elementary or in a global way. Now, we obtain, more accurate multigroup reaction rates and we are able to carry out a reference self-shielding calculation on a very fine multigroup mesh. To end, we draw a conclusion and give some prospects on the remaining work. (author)
International Nuclear Information System (INIS)
Ganapol, B.D.; Kornreich, D.E.
1997-01-01
Because of the requirement of accountability and quality control in the scientific world, a demand for high-quality analytical benchmark calculations has arisen in the neutron transport community. The intent of these benchmarks is to provide a numerical standard to which production neutron transport codes may be compared in order to verify proper operation. The overall investigation as modified in the second year renewal application includes the following three primary tasks. Task 1 on two dimensional neutron transport is divided into (a) single medium searchlight problem (SLP) and (b) two-adjacent half-space SLP. Task 2 on three-dimensional neutron transport covers (a) point source in arbitrary geometry, (b) single medium SLP, and (c) two-adjacent half-space SLP. Task 3 on code verification, includes deterministic and probabilistic codes. The primary aim of the proposed investigation was to provide a suite of comprehensive two- and three-dimensional analytical benchmarks for neutron transport theory applications. This objective has been achieved. The suite of benchmarks in infinite media and the three-dimensional SLP are a relatively comprehensive set of one-group benchmarks for isotropically scattering media. Because of time and resource limitations, the extensions of the benchmarks to include multi-group and anisotropic scattering are not included here. Presently, however, enormous advances in the solution for the planar Green's function in an anisotropically scattering medium have been made and will eventually be implemented in the two- and three-dimensional solutions considered under this grant. Of particular note in this work are the numerical results for the three-dimensional SLP, which have never before been presented. The results presented were made possible only because of the tremendous advances in computing power that have occurred during the past decade
Energy Technology Data Exchange (ETDEWEB)
Ganapol, B.D.; Kornreich, D.E. [Univ. of Arizona, Tucson, AZ (United States). Dept. of Nuclear Engineering
1997-07-01
Because of the requirement of accountability and quality control in the scientific world, a demand for high-quality analytical benchmark calculations has arisen in the neutron transport community. The intent of these benchmarks is to provide a numerical standard to which production neutron transport codes may be compared in order to verify proper operation. The overall investigation as modified in the second year renewal application includes the following three primary tasks. Task 1 on two dimensional neutron transport is divided into (a) single medium searchlight problem (SLP) and (b) two-adjacent half-space SLP. Task 2 on three-dimensional neutron transport covers (a) point source in arbitrary geometry, (b) single medium SLP, and (c) two-adjacent half-space SLP. Task 3 on code verification, includes deterministic and probabilistic codes. The primary aim of the proposed investigation was to provide a suite of comprehensive two- and three-dimensional analytical benchmarks for neutron transport theory applications. This objective has been achieved. The suite of benchmarks in infinite media and the three-dimensional SLP are a relatively comprehensive set of one-group benchmarks for isotropically scattering media. Because of time and resource limitations, the extensions of the benchmarks to include multi-group and anisotropic scattering are not included here. Presently, however, enormous advances in the solution for the planar Green`s function in an anisotropically scattering medium have been made and will eventually be implemented in the two- and three-dimensional solutions considered under this grant. Of particular note in this work are the numerical results for the three-dimensional SLP, which have never before been presented. The results presented were made possible only because of the tremendous advances in computing power that have occurred during the past decade.
Fowler, S. J.; Driesner, T.; Kulik, D.; Wagner, T.
2010-12-01
We present a novel computational tool for modelling temporally and spatially varying chemical interactions between hydrothermal fluids and rocks that may affect the long-term performance of geothermal reservoirs. The code is written in C++. It incorporates fluid-rock interaction and scale formation self-consistently, via a modular coupling approach that combines the Complex System Modelling Platform (CSMP++) code for fluid flow in porous and fractured media (Matthai et al., 2007) with the numerical kernel (GEMIPM2K) of the GEM-Selektor Gibbs free energy minimization package (Kulik, Wagner et al., 2007). CSMP++ uses finite element-finite volume spatial discretization, implicit or explicit time discretization, and an operator splitting approach to solve equations. The GEM-Selektor package supports a wide range of equation of state and activity models, facilitating calculation of complex fluid-mineral equilibria. Coupled code input includes temperature, pressure, a charge balance, and total amounts of system chemical elements, as well as domain and boundary condition specifications. Speciation, thermodynamic, and physical properties of the system are output. Critical advantages of the coupled code compared to existing hydrothermal reactive transport models are: (1) simultaneous consideration of complex solid solutions (e.g., clay minerals) and non-ideal aqueous solutions (GEMIPM2K), and (2) a discretization scheme that can be applied to mass and heat transport in irregular, geologically realistic geometries (CSMP++). Each coupled simulation results in a thermodynamically-based description of the geochemical and physical state of a hydrothermal system evolving along a complex P-T-X path. The code design allows for efficient and flexible incorporation of numerical and thermodynamic database improvements. We apply the coupled code to a number of geologic applications to test its accuracy and performance. Kulik, D., Wagner, T. et al. (2007). GEM-Selektor (GEMS-PSI) home
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)
Development of a space radiation Monte Carlo computer simulation based on the FLUKA and ROOT codes
Pinsky, L; Ferrari, A; Sala, P; Carminati, F; Brun, R
2001-01-01
This NASA funded project is proceeding to develop a Monte Carlo-based computer simulation of the radiation environment in space. With actual funding only initially in place at the end of May 2000, the study is still in the early stage of development. The general tasks have been identified and personnel have been selected. The code to be assembled will be based upon two major existing software packages. The radiation transport simulation will be accomplished by updating the FLUKA Monte Carlo program, and the user interface will employ the ROOT software being developed at CERN. The end-product will be a Monte Carlo-based code which will complement the existing analytic codes such as BRYNTRN/HZETRN presently used by NASA to evaluate the effects of radiation shielding in space. The planned code will possess the ability to evaluate the radiation environment for spacecraft and habitats in Earth orbit, in interplanetary space, on the lunar surface, or on a planetary surface such as Mars. Furthermore, it will be usef...
Energy Technology Data Exchange (ETDEWEB)
Baes, C.F. III; Sharp, R.D.; Sjoreen, A.L.; Hermann, O.W.
1984-11-01
TERRA is a computer code which calculates concentrations of radionuclides and ingrowing daughters in surface and root-zone soil, produce and feed, beef, and milk from a given deposition rate at any location in the conterminous United States. The code is fully integrated with seven other computer codes which together comprise a Computerized Radiological Risk Investigation System, CRRIS. Output from either the long range (> 100 km) atmospheric dispersion code RETADD-II or the short range (<80 km) atmospheric dispersion code ANEMOS, in the form of radionuclide air concentrations and ground deposition rates by downwind location, serves as input to TERRA. User-defined deposition rates and air concentrations may also be provided as input to TERRA through use of the PRIMUS computer code. The environmental concentrations of radionuclides predicted by TERRA serve as input to the ANDROS computer code which calculates population and individual intakes, exposures, doses, and risks. TERRA incorporates models to calculate uptake from soil and atmospheric deposition on four groups of produce for human consumption and four groups of livestock feeds. During the environmental transport simulation, intermediate calculations of interception fraction for leafy vegetables, produce directly exposed to atmospherically depositing material, pasture, hay, and silage are made based on location-specific estimates of standing crop biomass. Pasture productivity is estimated by a model which considers the number and types of cattle and sheep, pasture area, and annual production of other forages (hay and silage) at a given location. Calculations are made of the fraction of grain imported from outside the assessment area. TERRA output includes the above calculations and estimated radionuclide concentrations in plant produce, milk, and a beef composite by location.
International Nuclear Information System (INIS)
Baes, C.F. III; Sharp, R.D.; Sjoreen, A.L.; Hermann, O.W.
1984-11-01
TERRA is a computer code which calculates concentrations of radionuclides and ingrowing daughters in surface and root-zone soil, produce and feed, beef, and milk from a given deposition rate at any location in the conterminous United States. The code is fully integrated with seven other computer codes which together comprise a Computerized Radiological Risk Investigation System, CRRIS. Output from either the long range (> 100 km) atmospheric dispersion code RETADD-II or the short range (<80 km) atmospheric dispersion code ANEMOS, in the form of radionuclide air concentrations and ground deposition rates by downwind location, serves as input to TERRA. User-defined deposition rates and air concentrations may also be provided as input to TERRA through use of the PRIMUS computer code. The environmental concentrations of radionuclides predicted by TERRA serve as input to the ANDROS computer code which calculates population and individual intakes, exposures, doses, and risks. TERRA incorporates models to calculate uptake from soil and atmospheric deposition on four groups of produce for human consumption and four groups of livestock feeds. During the environmental transport simulation, intermediate calculations of interception fraction for leafy vegetables, produce directly exposed to atmospherically depositing material, pasture, hay, and silage are made based on location-specific estimates of standing crop biomass. Pasture productivity is estimated by a model which considers the number and types of cattle and sheep, pasture area, and annual production of other forages (hay and silage) at a given location. Calculations are made of the fraction of grain imported from outside the assessment area. TERRA output includes the above calculations and estimated radionuclide concentrations in plant produce, milk, and a beef composite by location
International Nuclear Information System (INIS)
Hiergesell, R.; Taylor, G.
2010-01-01
An investigation was conducted to compare and evaluate contaminant transport results of two model codes, GoldSim and Porflow, using a simple 1-D string of elements in each code. Model domains were constructed to be identical with respect to cell numbers and dimensions, matrix material, flow boundary and saturation conditions. One of the codes, GoldSim, does not simulate advective movement of water; therefore the water flux term was specified as a boundary condition. In the other code, Porflow, a steady-state flow field was computed and contaminant transport was simulated within that flow-field. The comparisons were made solely in terms of the ability of each code to perform contaminant transport. The purpose of the investigation was to establish a basis for, and to validate follow-on work that was conducted in which a 1-D GoldSim model developed by abstracting information from Porflow 2-D and 3-D unsaturated and saturated zone models and then benchmarked to produce equivalent contaminant transport results. A handful of contaminants were selected for the code-to-code comparison simulations, including a non-sorbing tracer and several long- and short-lived radionuclides exhibiting both non-sorbing to strongly-sorbing characteristics with respect to the matrix material, including several requiring the simulation of in-growth of daughter radionuclides. The same diffusion and partitioning coefficients associated with each contaminant and the half-lives associated with each radionuclide were incorporated into each model. A string of 10-elements, having identical spatial dimensions and properties, were constructed within each code. GoldSim's basic contaminant transport elements, Mixing cells, were utilized in this construction. Sand was established as the matrix material and was assigned identical properties (e.g. bulk density, porosity, saturated hydraulic conductivity) in both codes. Boundary conditions applied included an influx of water at the rate of 40 cm/yr at one
3D-TRANS-2003, Workshop on Common Tools and Interfaces for Radiation Transport Codes
International Nuclear Information System (INIS)
2004-01-01
Description: Contents proceedings of Workshop on Common Tools and Interfaces for Deterministic Radiation Transport, for Monte Carlo and Hybrid Codes with a proposal to develop the following: GERALD - A General Environment for Radiation Analysis and Design. GERALD intends to create a unifying software environment where the user can define, solve and analyse a nuclear radiation transport problem using available numerical tools seamlessly. This environment will serve many purposes: teaching, research, industrial needs. It will also help to preserve the existing analytical and numerical knowledge base. This could represent a significant step towards solving the legacy problem. This activity should contribute to attracting young engineers to nuclear science and engineering and contribute to competence and knowledge preservation and management. This proposal was made at the on Workshop on C ommon Tools and Interfaces for Deterministic Radiation Transport, for Monte Carlo and Hybrid Codes , held from 25-26 September 2003 in connection with the conference SNA-2003. A first success with the development of such tools was achieved with the BOT3P2.0 and 3.0 codes providing an easy procedure and mechanism for defining and displaying 3D geometries and materials both in the form of refineable meshes for deterministic codes or Monte Carlo geometries consistent with deterministic models. Advanced SUSD: Improved tools for Sensitivity/Uncertainty Analysis. The development of tools for the analysis and estimation of sensitivities and uncertainties in calculations, or their propagation through complex computational schemes, in the field of neutronics, thermal hydraulics and also thermo-mechanics is of increasing importance for research and engineering applications. These tools allow establishing better margins for engineering designs and for the safe operation of nuclear facilities. Such tools are not sufficiently developed, but their need is increasingly evident in many activities
Contaminant transport in fracture networks with heterogeneous rock matrices. The Picnic code
Energy Technology Data Exchange (ETDEWEB)
Barten, Werner [Paul Scherrer Inst., CH-5232 Villigen PSI (Switzerland); Robinson, Peter C. [QuantiSci Limited, Henley-on-Thames (United Kingdom)
2001-02-01
In the context of safety assessment of radioactive waste repositories, complex radionuclide transport models covering key safety-relevant processes play a major role. In recent Swiss safety assessments, such as Kristallin-I, an important drawback was the limitation in geosphere modelling capability to account for geosphere heterogeneities. In marked contrast to this limitation in modelling capabilities, great effort has been put into investigating the heterogeneity of the geosphere as it impacts on hydrology. Structural geological methods have been used to look at the geometry of the flow paths on a small scale and the diffusion and sorption properties of different rock materials have been investigated. This huge amount of information could however be only partially applied in geosphere transport modelling. To make use of these investigations the 'PICNIC project' was established as a joint cooperation of PSI/Nagra and QuantiSci to provide a new geosphere transport model for Swiss safety assessment of radioactive waste repositories. The new transport code, PICNIC, can treat all processes considered in the older geosphere model RANCH MD generally used in the Kristallin-I study and, in addition, explicitly accounts for the heterogeneity of the geosphere on different spatial scales. The effects and transport phenomena that can be accounted for by PICNIC are a combination of (advective) macro-dispersion due to transport in a network of conduits (legs), micro-dispersion in single legs, one-dimensional or two-dimensional matrix diffusion into a wide range of homogeneous and heterogeneous rock matrix geometries, linear sorption of nuclides in the flow path and the rock matrix and radioactive decay and ingrowth in the case of nuclide chains. Analytical and numerical Laplace transformation methods are integrated in a newly developed hierarchical linear response concept to efficiently account for the transport mechanisms considered which typically act on extremely
Contaminant transport in fracture networks with heterogeneous rock matrices. The Picnic code
International Nuclear Information System (INIS)
Barten, Werner; Robinson, Peter C.
2001-02-01
In the context of safety assessment of radioactive waste repositories, complex radionuclide transport models covering key safety-relevant processes play a major role. In recent Swiss safety assessments, such as Kristallin-I, an important drawback was the limitation in geosphere modelling capability to account for geosphere heterogeneities. In marked contrast to this limitation in modelling capabilities, great effort has been put into investigating the heterogeneity of the geosphere as it impacts on hydrology. Structural geological methods have been used to look at the geometry of the flow paths on a small scale and the diffusion and sorption properties of different rock materials have been investigated. This huge amount of information could however be only partially applied in geosphere transport modelling. To make use of these investigations the 'PICNIC project' was established as a joint cooperation of PSI/Nagra and QuantiSci to provide a new geosphere transport model for Swiss safety assessment of radioactive waste repositories. The new transport code, PICNIC, can treat all processes considered in the older geosphere model RANCH MD generally used in the Kristallin-I study and, in addition, explicitly accounts for the heterogeneity of the geosphere on different spatial scales. The effects and transport phenomena that can be accounted for by PICNIC are a combination of (advective) macro-dispersion due to transport in a network of conduits (legs), micro-dispersion in single legs, one-dimensional or two-dimensional matrix diffusion into a wide range of homogeneous and heterogeneous rock matrix geometries, linear sorption of nuclides in the flow path and the rock matrix and radioactive decay and ingrowth in the case of nuclide chains. Analytical and numerical Laplace transformation methods are integrated in a newly developed hierarchical linear response concept to efficiently account for the transport mechanisms considered which typically act on extremely different
Agent code: Neutron transport benchmark example and extension to 3D lattice geometry
Directory of Open Access Journals (Sweden)
Hursin Mathieu
2005-01-01
Full Text Available The general methodology be hind 2D arbitrary geometry neutron transport AGENT code is the theory of R-functions, which al lows for simple modeling of complex geometries, and the method of characteristics, which solves the integral transport equation along characteristic neutron trajectories. This paper focuses on the extension of the methodology to ac count for 3D lattice geometries. Since the direct application of method of characteristics to 3D non-homogenized core con figuration may re quire a tremendous amount of memory and computing time, an alternative approximate solution based on coupling 2D method of characteristics and 1D diffusion solution is developed. The planar 2D method of characteristics and axial 1D diffusion solutions are coupled through the trans verse leak age. The use of a lower order 1D solution in the axial direction is justified by the fact that more heterogeneity in current PWR and BWR reactor cores occurs in the radial direction than in the axial one. In order to demonstrate the versatility and accuracy of the AGENT code, a 2D heterogeneous lattice problem, C5G7 is described in details. A theoretical description of the coupling methodology for 3D method of characteristics solution is followed by preliminary validation in comparison to the DeCART code.
Benchmarking Heavy Ion Transport Codes FLUKA, HETC-HEDS MARS15, MCNPX, and PHITS
Energy Technology Data Exchange (ETDEWEB)
Ronningen, Reginald Martin [Michigan State University; Remec, Igor [Oak Ridge National Laboratory; Heilbronn, Lawrence H. [University of Tennessee-Knoxville
2013-06-07
Powerful accelerators such as spallation neutron sources, muon-collider/neutrino facilities, and rare isotope beam facilities must be designed with the consideration that they handle the beam power reliably and safely, and they must be optimized to yield maximum performance relative to their design requirements. The simulation codes used for design purposes must produce reliable results. If not, component and facility designs can become costly, have limited lifetime and usefulness, and could even be unsafe. The objective of this proposal is to assess the performance of the currently available codes PHITS, FLUKA, MARS15, MCNPX, and HETC-HEDS that could be used for design simulations involving heavy ion transport. We plan to access their performance by performing simulations and comparing results against experimental data of benchmark quality. Quantitative knowledge of the biases and the uncertainties of the simulations is essential as this potentially impacts the safe, reliable and cost effective design of any future radioactive ion beam facility. Further benchmarking of heavy-ion transport codes was one of the actions recommended in the Report of the 2003 RIA R&D Workshop".
An upgraded version of the nucleon meson transport code: NMTC/JAERI97
Energy Technology Data Exchange (ETDEWEB)
Takada, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Yoshizawa, Nobuaki; Kosako, Kazuaki; Ishibashi, Kenji
1998-02-01
The nucleon-meson transport code NMTC/JAERI is upgraded to NMTC/JAERI97 which has new features not only in physics model and nuclear data but also in computational procedure. NMTC/JAERI97 implements the following two new physics models: an intranuclear cascade model taking account of the in-medium nuclear effects and the preequilibrium calculation model based on the exciton one. For treating the nucleon transport process more accurately, the nucleon-nucleus cross sections are revised to those derived by the systematics of Pearlstein. Moreover, the level density parameter derived by Ignatyuk is included as a new option for particle evaporation calculation. Other than those physical aspects, a new geometry package based on the Combinatorial Geometry with multi-array system and the importance sampling technique are implemented in the code. Tally function is also employed for obtaining such physical quantities as neutron energy spectra, heat deposition and nuclide yield without editing a history file. The resultant NMTC/JAERI97 is tuned to be executed on the UNIX system. This paper explains about the function, physics models and geometry model adopted in NMTC/JAERI97 and guides how to use the code. (author)
Motivation for Using Generalized Geometry in the Time Dependent Transport Code TDKENO
Energy Technology Data Exchange (ETDEWEB)
Dustin Popp; Zander Mausolff; Sedat Goluoglu
2016-04-01
We are proposing to use the code, TDKENO, to model TREAT. TDKENO solves the time dependent, three dimensional Boltzmann transport equation with explicit representation of delayed neutrons. Instead of directly integrating this equation, the neutron flux is factored into two components – a rapidly varying amplitude equation and a slowly varying shape equation and each is solved separately on different time scales. The shape equation is solved using the 3D Monte Carlo transport code KENO, from Oak Ridge National Laboratory’s SCALE code package. Using the Monte Carlo method to solve the shape equation is still computationally intensive, but the operation is only performed when needed. The amplitude equation is solved deterministically and frequently, so the solution gives an accurate time-dependent solution without having to repeatedly We have modified TDKENO to incorporate KENO-VI so that we may accurately represent the geometries within TREAT. This paper explains the motivation behind using generalized geometry, and provides the results of our modifications. TDKENO uses the Improved Quasi-Static method to accomplish this. In this method, the neutron flux is factored into two components. One component is a purely time-dependent and rapidly varying amplitude function, which is solved deterministically and very frequently (small time steps). The other is a slowly varying flux shape function that weakly depends on time and is only solved when needed (significantly larger time steps).
Directory of Open Access Journals (Sweden)
Chapoutier Nicolas
2017-01-01
Full Text Available In the context of the rising of Monte Carlo transport calculations for any kind of application, AREVA recently improved its suite of engineering tools in order to produce efficient Monte Carlo workflow. Monte Carlo codes, such as MCNP or TRIPOLI, are recognized as reference codes to deal with a large range of radiation transport problems. However the inherent drawbacks of theses codes - laboring input file creation and long computation time - contrast with the maturity of the treatment of the physical phenomena. The goals of the recent AREVA developments were to reach similar efficiency as other mature engineering sciences such as finite elements analyses (e.g. structural or fluid dynamics. Among the main objectives, the creation of a graphical user interface offering CAD tools for geometry creation and other graphical features dedicated to the radiation field (source definition, tally definition has been reached. The computations times are drastically reduced compared to few years ago thanks to the use of massive parallel runs, and above all, the implementation of hybrid variance reduction technics. From now engineering teams are capable to deliver much more prompt support to any nuclear projects dealing with reactors or fuel cycle facilities from conceptual phase to decommissioning.
Chapoutier, Nicolas; Mollier, François; Nolin, Guillaume; Culioli, Matthieu; Mace, Jean-Reynald
2017-09-01
In the context of the rising of Monte Carlo transport calculations for any kind of application, AREVA recently improved its suite of engineering tools in order to produce efficient Monte Carlo workflow. Monte Carlo codes, such as MCNP or TRIPOLI, are recognized as reference codes to deal with a large range of radiation transport problems. However the inherent drawbacks of theses codes - laboring input file creation and long computation time - contrast with the maturity of the treatment of the physical phenomena. The goals of the recent AREVA developments were to reach similar efficiency as other mature engineering sciences such as finite elements analyses (e.g. structural or fluid dynamics). Among the main objectives, the creation of a graphical user interface offering CAD tools for geometry creation and other graphical features dedicated to the radiation field (source definition, tally definition) has been reached. The computations times are drastically reduced compared to few years ago thanks to the use of massive parallel runs, and above all, the implementation of hybrid variance reduction technics. From now engineering teams are capable to deliver much more prompt support to any nuclear projects dealing with reactors or fuel cycle facilities from conceptual phase to decommissioning.
International Nuclear Information System (INIS)
Fanaro, L.C.C.B.
1984-01-01
It was developed the BLINDAGE computer code for the radiation transport (neutrons and gammas) calculation. The code uses the removal - diffusion method for neutron transport and point-kernel technique with buil-up factors for gamma-rays. The results obtained through BLINDAGE code are compared with those obtained with the ANISN and SABINE computer codes. (Author) [pt
Parallelization of a three-dimensional whole core transport code DeCART
Energy Technology Data Exchange (ETDEWEB)
Jin Young, Cho; Han Gyu, Joo; Ha Yong, Kim; Moon-Hee, Chang [Korea Atomic Energy Research Institute, Yuseong-gu, Daejon (Korea, Republic of)
2003-07-01
Parallelization of the DeCART (deterministic core analysis based on ray tracing) code is presented that reduces the computational burden of the tremendous computing time and memory required in three-dimensional whole core transport calculations. The parallelization employs the concept of MPI grouping and the MPI/OpenMP mixed scheme as well. Since most of the computing time and memory are used in MOC (method of characteristics) and the multi-group CMFD (coarse mesh finite difference) calculation in DeCART, variables and subroutines related to these two modules are the primary targets for parallelization. Specifically, the ray tracing module was parallelized using a planar domain decomposition scheme and an angular domain decomposition scheme. The parallel performance of the DeCART code is evaluated by solving a rodded variation of the C5G7MOX three dimensional benchmark problem and a simplified three-dimensional SMART PWR core problem. In C5G7MOX problem with 24 CPUs, a speedup of maximum 21 is obtained on an IBM Regatta machine and 22 on a LINUX Cluster in the MOC kernel, which indicates good parallel performance of the DeCART code. In the simplified SMART problem, the memory requirement of about 11 GBytes in the single processor cases reduces to 940 Mbytes with 24 processors, which means that the DeCART code can now solve large core problems with affordable LINUX clusters. (authors)
Overview of development and design of MPACT: Michigan parallel characteristics transport code
International Nuclear Information System (INIS)
Kochunas, B.; Collins, B.; Jabaay, D.; Downar, T. J.; Martin, W. R.
2013-01-01
MPACT (Michigan Parallel Characteristics Transport Code) is a new reactor analysis tool. It is being developed by students and research staff at the University of Michigan to be used for an advanced pin-resolved transport capability within VERA (Virtual Environment for Reactor Analysis). VERA is the end-user reactor simulation tool being produced by the Consortium for the Advanced Simulation of Light Water Reactors (CASL). The MPACT development project is itself unique for the way it is changing how students do research to achieve the instructional and research goals of an academic institution, while providing immediate value to industry. The MPACT code makes use of modern lean/agile software processes and extensive testing to maintain a level of productivity and quality required by CASL. MPACT's design relies heavily on object-oriented programming concepts and design patterns and is programmed in Fortran 2003. These designs are explained and illustrated as to how they can be readily extended to incorporate new capabilities and research ideas in support of academic research objectives. The transport methods currently implemented in MPACT include the 2-D and 3-D method of characteristics (MOC) and 2-D and 3-D method of collision direction probabilities (CDP). For the cross section resonance treatment, presently the subgroup method and the new embedded self-shielding method (ESSM) are implemented within MPACT. (authors)
ACCEPT: three-dimensional electron/photon Monte Carlo transport code using combinatorial geometry
Energy Technology Data Exchange (ETDEWEB)
Halbleib, J.A. Sr.
1979-05-01
The ACCEPT code provides experimenters and theorists with a method for the routine solution of coupled electron/photon transport through three-dimensional multimaterial geometries described by the combinational method. Emphasis is placed upon operational simplicity without sacrificing the rigor of the model. ACCEPT combines condensed-history electron Monte Carlo with conventional single-scattering photon Monte Carlo in order to describe the transport of all generations of particles from several MeV down to 1.0 and 10.0 keV for electrons and photons, respectively. The model is more accurate at the higher energies with a less rigorous description of the particle cascade at energies where the shell structure of the transport media becomes important. Flexibility of construction permits the user to tailor the model to specific applications and to extend the capabilities of the model to more sophisticated applications through relatively simple update procedures. The ACCEPT code is currently running on the CDC-7600 (66000) where the bulk of the cross-section data and the statistical variables are stored in Large Core Memory (Extended Core Storage).
ACCEPT: three-dimensional electron/photon Monte Carlo transport code using combinatorial geometry
International Nuclear Information System (INIS)
Halbleib, J.A. Sr.
1979-05-01
The ACCEPT code provides experimenters and theorists with a method for the routine solution of coupled electron/photon transport through three-dimensional multimaterial geometries described by the combinational method. Emphasis is placed upon operational simplicity without sacrificing the rigor of the model. ACCEPT combines condensed-history electron Monte Carlo with conventional single-scattering photon Monte Carlo in order to describe the transport of all generations of particles from several MeV down to 1.0 and 10.0 keV for electrons and photons, respectively. The model is more accurate at the higher energies with a less rigorous description of the particle cascade at energies where the shell structure of the transport media becomes important. Flexibility of construction permits the user to tailor the model to specific applications and to extend the capabilities of the model to more sophisticated applications through relatively simple update procedures. The ACCEPT code is currently running on the CDC-7600 (66000) where the bulk of the cross-section data and the statistical variables are stored in Large Core Memory
PRESTO-II: a low-level waste environmental transport and risk assessment code
Energy Technology Data Exchange (ETDEWEB)
Fields, D.E.; Emerson, C.J.; Chester, R.O.; Little, C.A.; Hiromoto, G.
1986-04-01
PRESTO-II (Prediction of Radiation Effects from Shallow Trench Operations) is a computer code designed for the evaluation of possible health effects from shallow-land and, waste-disposal trenches. The model is intended to serve as a non-site-specific screening model for assessing radionuclide transport, ensuing exposure, and health impacts to a static local population for a 1000-year period following the end of disposal operations. Human exposure scenarios considered include normal releases (including leaching and operational spillage), human intrusion, and limited site farming or reclamation. Pathways and processes of transit from the trench to an individual or population include ground-water transport, overland flow, erosion, surface water dilution, suspension, atmospheric transport, deposition, inhalation, external exposure, and ingestion of contaminated beef, milk, crops, and water. Both population doses and individual doses, as well as doses to the intruder and farmer, may be calculated. Cumulative health effects in terms of cancer deaths are calculated for the population over the 1000-year period using a life-table approach. Data are included for three example sites: Barnwell, South Carolina; Beatty, Nevada; and West Valley, New York. A code listing and example input for each of the three sites are included in the appendices to this report.
PRESTO-II: a low-level waste environmental transport and risk assessment code
International Nuclear Information System (INIS)
Fields, D.E.; Emerson, C.J.; Chester, R.O.; Little, C.A.; Hiromoto, G.
1986-04-01
PRESTO-II (Prediction of Radiation Effects from Shallow Trench Operations) is a computer code designed for the evaluation of possible health effects from shallow-land and, waste-disposal trenches. The model is intended to serve as a non-site-specific screening model for assessing radionuclide transport, ensuing exposure, and health impacts to a static local population for a 1000-year period following the end of disposal operations. Human exposure scenarios considered include normal releases (including leaching and operational spillage), human intrusion, and limited site farming or reclamation. Pathways and processes of transit from the trench to an individual or population include ground-water transport, overland flow, erosion, surface water dilution, suspension, atmospheric transport, deposition, inhalation, external exposure, and ingestion of contaminated beef, milk, crops, and water. Both population doses and individual doses, as well as doses to the intruder and farmer, may be calculated. Cumulative health effects in terms of cancer deaths are calculated for the population over the 1000-year period using a life-table approach. Data are included for three example sites: Barnwell, South Carolina; Beatty, Nevada; and West Valley, New York. A code listing and example input for each of the three sites are included in the appendices to this report
Particle and heavy ion transport code system, PHITS, version 2.52
International Nuclear Information System (INIS)
Sato, Tatsuhiko; Matsuda, Norihiro; Hashimoto, Shintaro; Iwamoto, Yosuke; Noda, Shusaku; Ogawa, Tatsuhiko; Nakashima, Hiroshi; Fukahori, Tokio; Okumura, Keisuke; Kai, Tetsuya; Niita, Koji; Iwase, Hiroshi; Chiba, Satoshi; Furuta, Takuya; Sihver, Lembit
2013-01-01
An upgraded version of the Particle and Heavy Ion Transport code System, PHITS2.52, was developed and released to the public. The new version has been greatly improved from the previously released version, PHITS2.24, in terms of not only the code itself but also the contents of its package, such as the attached data libraries. In the new version, a higher accuracy of simulation was achieved by implementing several latest nuclear reaction models. The reliability of the simulation was improved by modifying both the algorithms for the electron-, positron-, and photon-transport simulations and the procedure for calculating the statistical uncertainties of the tally results. Estimation of the time evolution of radioactivity became feasible by incorporating the activation calculation program DCHAIN-SP into the new package. The efficiency of the simulation was also improved as a result of the implementation of shared-memory parallelization and the optimization of several time-consuming algorithms. Furthermore, a number of new user-support tools and functions that help users to intuitively and effectively perform PHITS simulations were developed and incorporated. Due to these improvements, PHITS is now a more powerful tool for particle transport simulation applicable to various research and development fields, such as nuclear technology, accelerator design, medical physics, and cosmic-ray research. (author)
The new deterministic 3-D radiation transport code Multitrans: C5G7 MOX fuel assembly benchmark
International Nuclear Information System (INIS)
Kotiluoto, P.
2003-01-01
The novel deterministic three-dimensional radiation transport code MultiTrans is based on combination of the advanced tree multigrid technique and the simplified P3 (SP3) radiation transport approximation. In the tree multigrid technique, an automatic mesh refinement is performed on material surfaces. The tree multigrid is generated directly from stereo-lithography (STL) files exported by computer-aided design (CAD) systems, thus allowing an easy interface for construction and upgrading of the geometry. The deterministic MultiTrans code allows fast solution of complicated three-dimensional transport problems in detail, offering a new tool for nuclear applications in reactor physics. In order to determine the feasibility of a new code, computational benchmarks need to be carried out. In this work, MultiTrans code is tested for a seven-group three-dimensional MOX fuel assembly transport benchmark without spatial homogenization (NEA C5G7 MOX). (author)
OpenGeoSys-GEMS: Hybrid parallelization of a reactive transport code with MPI and threads
Kosakowski, G.; Kulik, D. A.; Shao, H.
2012-04-01
OpenGeoSys-GEMS is a generic purpose reactive transport code based on the operator splitting approach. The code couples the Finite-Element groundwater flow and multi-species transport modules of the OpenGeoSys (OGS) project (http://www.ufz.de/index.php?en=18345) with the GEM-Selektor research package to model thermodynamic equilibrium of aquatic (geo)chemical systems utilizing the Gibbs Energy Minimization approach (http://gems.web.psi.ch/). The combination of OGS and the GEM-Selektor kernel (GEMS3K) is highly flexible due to the object-oriented modular code structures and the well defined (memory based) data exchange modules. Like other reactive transport codes, the practical applicability of OGS-GEMS is often hampered by the long calculation time and large memory requirements. • For realistic geochemical systems which might include dozens of mineral phases and several (non-ideal) solid solutions the time needed to solve the chemical system with GEMS3K may increase exceptionally. • The codes are coupled in a sequential non-iterative loop. In order to keep the accuracy, the time step size is restricted. In combination with a fine spatial discretization the time step size may become very small which increases calculation times drastically even for small 1D problems. • The current version of OGS is not optimized for memory use and the MPI version of OGS does not distribute data between nodes. Even for moderately small 2D problems the number of MPI processes that fit into memory of up-to-date workstations or HPC hardware is limited. One strategy to overcome the above mentioned restrictions of OGS-GEMS is to parallelize the coupled code. For OGS a parallelized version already exists. It is based on a domain decomposition method implemented with MPI and provides a parallel solver for fluid and mass transport processes. In the coupled code, after solving fluid flow and solute transport, geochemical calculations are done in form of a central loop over all finite
The EGS4 Code System: Solution of Gamma-ray and Electron Transport Problems
Nelson, W. R.; Namito, Yoshihito
1990-03-01
In this paper we present an overview of the EGS4 Code System -- a general purpose package for the Monte Carlo simulation of the transport of electrons and photons. During the last 10-15 years EGS has been widely used to design accelerators and detectors for high-energy physics. More recently the code has been found to be of tremendous use in medical radiation physics and dosimetry. The problem-solving capabilities of EGS4 will be demonstrated by means of a variety of practical examples. To facilitate this review, we will take advantage of a new add-on package, called SHOWGRAF, to display particle trajectories in complicated geometries. These are shown as 2-D laser pictures in the written paper and as photographic slides of a 3-D high-resolution color monitor during the oral presentation. 11 refs., 15 figs.
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)
Geometry system used in the General Monte Carlo transport code SPARTAN
International Nuclear Information System (INIS)
Bending, R.C.; Easter, P.G.
1974-01-01
The geometry routines used in the general-purpose, three-dimensional particle transport code SPARTAN are described. The code is designed to deal with the very complex geometries encountered in lattice cell and fuel handling calculations, health physics, and shielding problems. Regions of the system being studied may be represented by simple shapes (spheres, cylinders, and so on) or by multinomial surfaces of any order, and many simple shapes may be combined to make up a complex layout. The geometry routines are designed to allow the program to carry out a number of tasks (such as sampling for a random point or tracking a path through several regions) in any order, so that the use of the routines is not restricted to a particular tracking or scoring method. Routines for reading, checking, and printing the data are included. (U.S.)
International Nuclear Information System (INIS)
Fletcher, J.K.
1987-12-01
The computer code MARC/PN provides a solution of the multigroup transport equation by expanding the flux in spherical harmonics. The coefficients of the series so obtained satisfy linked first order differential equations, and on eliminating terms associated with odd parity harmonics a second order system results which can be solved by established finite difference or finite element techniques. This report describes modifications incorporated in MARC/PN to allow for anisotropic scattering, and the modelling of irregular exterior boundaries in the finite element option. The latter development leads to substantial reductions in problem size, particularly for three dimensions. Also, links to an interactive graphics mesh generator (SUPERTAB) have been added. The final section of the report contains results from problems showing the effects of anisotropic scatter and the ability of the code to model irregular geometries. (author)
Tripoli-3: monte Carlo transport code for neutral particles - version 3.5 - users manual
International Nuclear Information System (INIS)
Vergnaud, Th.; Nimal, J.C.; Chiron, M.
2001-01-01
The TRIPOLI-3 code applies the Monte Carlo method to neutron, gamma-ray and coupled neutron and gamma-ray transport calculations in three-dimensional geometries, either in steady-state conditions or having a time dependence. It can be used to study problems where there is a high flux attenuation between the source zone and the result zone (studies of shielding configurations or source driven sub-critical systems, with fission being taken into account), as well as problems where there is a low flux attenuation (neutronic calculations -- in a fuel lattice cell, for example -- where fission is taken into account, usually with the calculation on the effective multiplication factor, fine structure studies, numerical experiments to investigate methods approximations, etc). TRIPOLI-3 has been operational since 1995 and is the version of the TRIPOLI code that follows on from TRIPOLI-2; it can be used on SUN, RISC600 and HP workstations and on PC using the Linux or Windows/NT operating systems. The code uses nuclear data libraries generated using the THEMIS/NJOY system. The current libraries were derived from ENDF/B6 and JEF2. There is also a response function library based on a number of evaluations, notably the dosimetry libraries IRDF/85, IRDF/90 and also evaluations from JEF2. The treatment of particle transport is the same in version 3.5 as in version 3.4 of the TRIPOLI code; but the version 3.5 is more convenient for preparing the input data and for reading the output. The french version of the user's manual exists. (authors)
A GPU-based Monte Carlo dose calculation code for photon transport in a voxel phantom
Energy Technology Data Exchange (ETDEWEB)
Bellezzo, M.; Do Nascimento, E.; Yoriyaz, H., E-mail: mbellezzo@gmail.br [Instituto de Pesquisas Energeticas e Nucleares / CNEN, Av. Lineu Prestes 2242, Cidade Universitaria, 05508-000 Sao Paulo (Brazil)
2014-08-15
As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo method has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this paper, we present the CUBMC code, a GPU-based Mc photon transport algorithm for dose calculation under the Compute Unified Device Architecture platform. The simulation of physical events is based on the algorithm used in Penelope, and the cross section table used is the one generated by the Material routine, als present in Penelope code. Photons are transported in voxel-based geometries with different compositions. To demonstrate the capabilities of the algorithm developed in the present work four 128 x 128 x 128 voxel phantoms have been considered. One of them is composed by a homogeneous water-based media, the second is composed by bone, the third is composed by lung and the fourth is composed by a heterogeneous bone and vacuum geometry. Simulations were done considering a 6 MeV monoenergetic photon point source. There are two distinct approaches that were used for transport simulation. The first of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon stop in the frontier will be considered depending on the material changing across the photon travel line. Dose calculations using these methods are compared for validation with Penelope and MCNP5 codes. Speed-up factors are compared using a NVidia GTX 560-Ti GPU card against a 2.27 GHz Intel Xeon CPU processor. (Author)
A GPU-based Monte Carlo dose calculation code for photon transport in a voxel phantom
International Nuclear Information System (INIS)
Bellezzo, M.; Do Nascimento, E.; Yoriyaz, H.
2014-08-01
As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo method has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this paper, we present the CUBMC code, a GPU-based Mc photon transport algorithm for dose calculation under the Compute Unified Device Architecture platform. The simulation of physical events is based on the algorithm used in Penelope, and the cross section table used is the one generated by the Material routine, als present in Penelope code. Photons are transported in voxel-based geometries with different compositions. To demonstrate the capabilities of the algorithm developed in the present work four 128 x 128 x 128 voxel phantoms have been considered. One of them is composed by a homogeneous water-based media, the second is composed by bone, the third is composed by lung and the fourth is composed by a heterogeneous bone and vacuum geometry. Simulations were done considering a 6 MeV monoenergetic photon point source. There are two distinct approaches that were used for transport simulation. The first of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon stop in the frontier will be considered depending on the material changing across the photon travel line. Dose calculations using these methods are compared for validation with Penelope and MCNP5 codes. Speed-up factors are compared using a NVidia GTX 560-Ti GPU card against a 2.27 GHz Intel Xeon CPU processor. (Author)
International Nuclear Information System (INIS)
Wells, F.H.; Powell, R.G.
1980-10-01
The Code of Practice and design principles for portable and transportable radiological protection systems are presented in three parts. Part 1 specifies the requirement for Radiological Protection Instrumentation (RPI) including operational characteristics and the effects of both a radiation and non-radiation environment. Part 2 satisfies the requirement for RPI equipment as regards the overall design, the availability, the reliability, the information display, the human factors, the power supplies, the manufacture and quality assurance, the testing and the cost. Part 3 deals with the supply, location and operation of the RPI equipment. (U.K.)
A new nuclide transport model in soil in the GENII-LIN health physics code
Teodori, F.
2017-11-01
The nuclide soil transfer model, originally included in the GENII-LIN software system, was intended for residual contamination from long term activities and from waste form degradation. Short life nuclides were supposed absent or at equilibrium with long life parents. Here we present an enhanced soil transport model, where short life nuclide contributions are correctly accounted. This improvement extends the code capabilities to handle incidental release of contaminant to soil, by evaluating exposure since the very beginning of the contamination event, before the radioactive decay chain equilibrium is reached.
The fusion code XGC: Enabling kinetic study of multi-scale edge turbulent transport in ITER
Energy Technology Data Exchange (ETDEWEB)
D' Azevedo, Eduardo [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Abbott, Stephen [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Koskela, Tuomas [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Worley, Patrick [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ku, Seung-Hoe [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Ethier, Stephane [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Yoon, Eisung [Rensselaer Polytechnic Inst., Troy, NY (United States); Shephard, Mark [Rensselaer Polytechnic Inst., Troy, NY (United States); Hager, Robert [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Lang, Jianying [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Intel Corporation, Santa Clara, CA (United States); Choi, Jong [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Podhorszki, Norbert [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Klasky, Scott [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Parashar, Manish [Rutgers Univ., Piscataway, NJ (United States); Chang, Choong-Seock [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)
2017-01-01
The XGC fusion gyrokinetic code combines state-of-the-art, portable computational and algorithmic technologies to enable complicated multiscale simulations of turbulence and transport dynamics in ITER edge plasma on the largest US open-science computer, the CRAY XK7 Titan, at its maximal heterogeneous capability, which have not been possible before due to a factor of over 10 shortage in the time-to-solution for less than 5 days of wall-clock time for one physics case. Frontier techniques such as nested OpenMP parallelism, adaptive parallel I/O, staging I/O and data reduction using dynamic and asynchronous applications interactions, dynamic repartitioning.
Nupack, the new ASME code for radioactive material transportation packaging containments
International Nuclear Information System (INIS)
Turula, P.
1998-01-01
The American Society of Mechanical Engineers (ASME) has added a new division to the nuclear construction section of its Boiler and Pressure Vessel Code (B and PVC). This Division, commonly referred to as Nupack, has been written to provide a consistent set of technical requirements for containment vessels of transportation packagings for high-level radioactive materials. This paper provides an introduction to Nupack, discusses some of its technical provisions, and describes how it can be used for the design and construction of packaging components. Nupack's general provisions and design requirements are emphasized, while treatment of materials, fabrication and inspection is left for another paper
One-dimensional transport code for one-group problems in plane geometry
International Nuclear Information System (INIS)
Bareiss, E.H.; Chamot, C.
1970-09-01
Equations and results are given for various methods of solution of the one-dimensional transport equation for one energy group in plane geometry with inelastic scattering and an isotropic source. After considerable investigation, a matrix method of solution was found to be faster and more stable than iteration procedures. A description of the code is included which allows for up to 24 regions, 250 points, and 16 angles such that the product of the number of angles and the number of points is less than 600
Resuspension of toxic aerosol using MATHEW--ADPIC wind field--transport and diffusion codes
International Nuclear Information System (INIS)
Porch, W.M.
1979-01-01
Computer codes have been written which estimate toxic aerosol resuspension based on computed deposition from a primary source, wind, and surface characteristics. The primary deposition pattern and the transport, diffusion, and redeposition of the resuspended toxic aerosol are calculated using a mass-consistent wind field model including topography (MATHEW) and a particle-in-cell diffusion and transport model (ADPIC) which were developed at LLL. The source term for resuspended toxic aerosol is determined by multiplying the total aerosol flux as a function of wind speed by the area of highest concentration and the fraction of suspended material estimated to be toxic. Preliminary calculations based on a test problem at the Nevada Test Site determined an hourly averaged maximum resuspension factor of 10 -4 for a 15 m/sec wind which is within an admittedly large range of resuspension factor measurements using experimental data
Recent Improvements of Particle and Heavy Ion Transport code System: PHITS
Sato, Tatsuhiko; Niita, Koji; Iwamoto, Yosuke; Hashimoto, Shintaro; Ogawa, Tatsuhiko; Furuta, Takuya; Abe, Shin-ichiro; Kai, Takeshi; Matsuda, Norihiro; Okumura, Keisuke; Kai, Tetsuya; Iwase, Hiroshi; Sihver, Lembit
2017-09-01
The Particle and Heavy Ion Transport code System, PHITS, has been developed under the collaboration of several research institutes in Japan and Europe. This system can simulate the transport of most particles with energy levels up to 1 TeV (per nucleon for ion) using different nuclear reaction models and data libraries. More than 2,500 registered researchers and technicians have used this system for various applications such as accelerator design, radiation shielding and protection, medical physics, and space- and geo-sciences. This paper summarizes the physics models and functions recently implemented in PHITS, between versions 2.52 and 2.88, especially those related to source generation useful for simulating brachytherapy and internal exposures of radioisotopes.
International Nuclear Information System (INIS)
Choi, Sung Hoon; Kwark, Min Su; Shim, Hyung Jin
2012-01-01
As The Monte Carlo (MC) particle transport analysis for a complex system such as research reactor, accelerator, and fusion facility may require accurate modeling of the complicated geometry. Its manual modeling by using the text interface of a MC code to define the geometrical objects is tedious, lengthy and error-prone. This problem can be overcome by taking advantage of modeling capability of the computer aided design (CAD) system. There have been two kinds of approaches to develop MC code systems utilizing the CAD data: the external format conversion and the CAD kernel imbedded MC simulation. The first approach includes several interfacing programs such as McCAD, MCAM, GEOMIT etc. which were developed to automatically convert the CAD data into the MCNP geometry input data. This approach makes the most of the existing MC codes without any modifications, but implies latent data inconsistency due to the difference of the geometry modeling system. In the second approach, a MC code utilizes the CAD data for the direct particle tracking or the conversion to an internal data structure of the constructive solid geometry (CSG) and/or boundary representation (B-rep) modeling with help of a CAD kernel. MCNP-BRL and OiNC have demonstrated their capabilities of the CAD-based MC simulations. Recently we have developed a CAD-based geometry processing module for the MC particle simulation by using the OpenCASCADE (OCC) library. In the developed module, CAD data can be used for the particle tracking through primitive CAD surfaces (hereafter the CAD-based tracking) or the internal conversion to the CSG data structure. In this paper, the performances of the text-based model, the CAD-based tracking, and the internal CSG conversion are compared by using an in-house MC code, McSIM, equipped with the developed CAD-based geometry processing module
Cullen, D
2000-01-01
TART2000 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input Preparation, running Monte Carlo calculations, and analysis of output results. TART2000 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART2000 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART2000 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART2000 and its data files.
International Nuclear Information System (INIS)
Cullen, D.E
2000-01-01
TART2000 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input Preparation, running Monte Carlo calculations, and analysis of output results. TART2000 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART2000 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART2000 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART2000 and its data files
On the Way to Future's High Energy Particle Physics Transport Code
Bíró, Gábor; Futó, Endre
2015-01-01
High Energy Physics (HEP) needs a huge amount of computing resources. In addition data acquisition, transfer, and analysis require a well developed infrastructure too. In order to prove new physics disciplines it is required to higher the luminosity of the accelerator facilities, which produce more-and-more data in the experimental detectors. Both testing new theories and detector R&D are based on complex simulations. Today have already reach that level, the Monte Carlo detector simulation takes much more time than real data collection. This is why speed up of the calculations and simulations became important in the HEP community. The Geant Vector Prototype (GeantV) project aims to optimize the most-used particle transport code applying parallel computing and to exploit the capabilities of the modern CPU and GPU architectures as well. With the maximized concurrency at multiple levels the GeantV is intended to be the successor of the Geant4 particle transport code that has been used since two decades succe...
The TORT three-dimensional discrete ordinates neutron/photon transport code (TORT version 3)
Energy Technology Data Exchange (ETDEWEB)
Rhoades, W.A.; Simpson, D.B.
1997-10-01
TORT calculates the flux or fluence of neutrons and/or photons throughout three-dimensional systems due to particles incident upon the system`s external boundaries, due to fixed internal sources, or due to sources generated by interaction with the system materials. The transport process is represented by the Boltzman transport equation. The method of discrete ordinates is used to treat the directional variable, and a multigroup formulation treats the energy dependence. Anisotropic scattering is treated using a Legendre expansion. Various methods are used to treat spatial dependence, including nodal and characteristic procedures that have been especially adapted to resist numerical distortion. A method of body overlay assists in material zone specification, or the specification can be generated by an external code supplied by the user. Several special features are designed to concentrate machine resources where they are most needed. The directional quadrature and Legendre expansion can vary with energy group. A discontinuous mesh capability has been shown to reduce the size of large problems by a factor of roughly three in some cases. The emphasis in this code is a robust, adaptable application of time-tested methods, together with a few well-tested extensions.
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)
GEOTHER: a two-phase fluid-flow and heat-transport code
International Nuclear Information System (INIS)
1983-04-01
GEOTHER is a three-dimensional geothermal reservoir simulation code. The model describes heat transport and flow of a single component, two-phase fluid in porous media. It is based on the continuity equations for steam and water, which are reduced to two nonlinear partial differential equations in which the dependent variables are fluid pressure and enthalpy. These equations, describing three-dimensional effects, are approximated using finite-difference techniques and are solved using an iterative technique. The nonlinear coefficients are calculated using Newton-Raphson iteration, and an option is provided for using either upstream or midpoint weighting on the mobility terms. GEOTHER can be used to simulate the fluid-thermal interaction in rock that can be approximated by a porous media representation. It can simulate heat transport and the flow of compressed water, two-phase mixtures, and super-heated steam in porous media over a temperature range of 10 to 300 0 C. In addition, it can treat the conversion from single- to two-phase flow, and vice versa. It can be used for evaluation of a near repository spatial scale and a time scale of a few years to thousands of years. The model can be used to investigate temperature and fluid pressure changes in response to thermal loading by waste materials. In Section 1.5 of this document the code custodianship and control is described along with the status of verification, validation and peer review of this report
MCNP: a general Monte Carlo code for neutron and photon transport
International Nuclear Information System (INIS)
1979-11-01
The general-purpose Monte Carlo code MCNP ca be used for neutron, photon, or coupled neutron-photon transport, including the capability to calculate eigenvalues for critical systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces (elliptical tori). Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation are accounted for. Thermal neutrons are described by both the free-gas and S(α,β) models. For photons, the code takes account of incoherent and coherent scattering, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. MCNP includes an elaborate, interactive plotting capability that allows the user to view his input geometry to help check for setup errors. Standard features which are available to improve computational efficiency include geometry splitting and Russian roulette, weight cutoff with Russian roulette, correlated sampling, analog capture or capture by weight reduction, the exponential transformation, energy splitting, forced collisions in designated cells, flux estimates at point or ring detectors, deterministically transporting pseudo-particles to designated regions, track-length estimators, source biasing, and several parameter cutoffs. Extensive summary information is provided to help the user better understand the physics and Monte Carlo simulation of his problem. The standard, user-defined output of MCNP includes two-way current as a function of direction across any set of surfaces or surface segments in the problem. Flux across any set of surfaces or surface segments is available. 58 figures, 28 tables
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
Core2D. A code for non-isothermal water flow and reactive solute transport. Users manual version 2
International Nuclear Information System (INIS)
Samper, J.; Juncosa, R.; Delgado, J.; Montenegro, L.
2000-01-01
Understanding natural groundwater quality patterns, quantifying groundwater pollution and assessing the effects of waste disposal, require modeling tools accounting for water flow, and transport of heat and dissolved species as well as their complex interactions with solid and gases phases. This report contains the users manual of CORE ''2D Version V.2.0, a COde for modeling water flow (saturated and unsaturated), heat transport and multicomponent Reactive solute transport under both local chemical equilibrium and kinetic conditions. it is an updated and improved version of CORE-LE-2D V0 (Samper et al., 1988) which in turns is an extended version of TRANQUI, a previous reactive transport code (ENRESA, 1995). All these codes were developed within the context of Research Projects funded by ENRESA and the European Commission. (Author)
Core 2D. A code for non-isothermal water flow and reactive solute transport. Users manual version 2
Energy Technology Data Exchange (ETDEWEB)
Samper, J.; Juncosa, R.; Delgado, J.; Montenegro, L. [Universidad de A Coruna (Spain)
2000-07-01
Understanding natural groundwater quality patterns, quantifying groundwater pollution and assessing the effects of waste disposal, require modeling tools accounting for water flow, and transport of heat and dissolved species as well as their complex interactions with solid and gases phases. This report contains the users manual of CORE ''2D Version V.2.0, a COde for modeling water flow (saturated and unsaturated), heat transport and multicomponent Reactive solute transport under both local chemical equilibrium and kinetic conditions. it is an updated and improved version of CORE-LE-2D V0 (Samper et al., 1988) which in turns is an extended version of TRANQUI, a previous reactive transport code (ENRESA, 1995). All these codes were developed within the context of Research Projects funded by ENRESA and the European Commission. (Author)
High performance 3D neutron transport on peta scale and hybrid architectures within APOLLO3 code
International Nuclear Information System (INIS)
Jamelot, E.; Dubois, J.; Lautard, J-J.; Calvin, C.; Baudron, A-M.
2011-01-01
APOLLO3 code is a common project of CEA, AREVA and EDF for the development of a new generation system for core physics analysis. We present here the parallelization of two deterministic transport solvers of APOLLO3: MINOS, a simplified 3D transport solver on structured Cartesian and hexagonal grids, and MINARET, a transport solver based on triangular meshes on 2D and prismatic ones in 3D. We used two different techniques to accelerate MINOS: a domain decomposition method, combined with an accelerated algorithm using GPU. The domain decomposition is based on the Schwarz iterative algorithm, with Robin boundary conditions to exchange information. The Robin parameters influence the convergence and we detail how we optimized the choice of these parameters. MINARET parallelization is based on angular directions calculation using explicit message passing. Fine grain parallelization is also available for each angular direction using shared memory multithreaded acceleration. Many performance results are presented on massively parallel architectures using more than 103 cores and on hybrid architectures using some tens of GPUs. This work contributes to the HPC development in reactor physics at the CEA Nuclear Energy Division. (author)
PyMercury: Interactive Python for the Mercury Monte Carlo Particle Transport Code
International Nuclear Information System (INIS)
Iandola, F.N.; O'Brien, M.J.; Procassini, R.J.
2010-01-01
Monte Carlo particle transport applications are often written in low-level languages (C/C++) for optimal performance on clusters and supercomputers. However, this development approach often sacrifices straightforward usability and testing in the interest of fast application performance. To improve usability, some high-performance computing applications employ mixed-language programming with high-level and low-level languages. In this study, we consider the benefits of incorporating an interactive Python interface into a Monte Carlo application. With PyMercury, a new Python extension to the Mercury general-purpose Monte Carlo particle transport code, we improve application usability without diminishing performance. In two case studies, we illustrate how PyMercury improves usability and simplifies testing and validation in a Monte Carlo application. In short, PyMercury demonstrates the value of interactive Python for Monte Carlo particle transport applications. In the future, we expect interactive Python to play an increasingly significant role in Monte Carlo usage and testing.
International Nuclear Information System (INIS)
Simmons, C.S.; Cole, C.R.
1985-05-01
This document was written to provide guidance to managers and site operators on how ground-water transport codes should be selected for assessing burial site performance. There is a need for a formal approach to selecting appropriate codes from the multitude of potentially useful ground-water transport codes that are currently available. Code selection is a problem that requires more than merely considering mathematical equation-solving methods. These guidelines are very general and flexible and are also meant for developing systems simulation models to be used to assess the environmental safety of low-level waste burial facilities. Code selection is only a single aspect of the overall objective of developing a systems simulation model for a burial site. The guidance given here is mainly directed toward applications-oriented users, but managers and site operators need to be familiar with this information to direct the development of scientifically credible and defensible transport assessment models. Some specific advice for managers and site operators on how to direct a modeling exercise is based on the following five steps: identify specific questions and study objectives; establish costs and schedules for achieving answers; enlist the aid of professional model applications group; decide on approach with applications group and guide code selection; and facilitate the availability of site-specific data. These five steps for managers/site operators are discussed in detail following an explanation of the nine systems model development steps, which are presented first to clarify what code selection entails
International Nuclear Information System (INIS)
Yabusaki, S.; Cole, C.; Monti, A.M.; Gupta, S.K.
1987-04-01
Part of the safety analysis is evaluating groundwater flow through the repository and the host rock to the accessible environment by developing mathematical or analytical models and numerical computer codes describing the flow mechanisms. This need led to the establishment of an international project called HYDROCOIN (HYDROlogic COde INtercomparison) organized by the Swedish Nuclear Power Inspectorate, a forum for discussing techniques and strategies in subsurface hydrologic modeling. The major objective of the present effort, HYDROCOIN Level 1, is determining the numerical accuracy of the computer codes. The definition of each case includes the input parameters, the governing equations, the output specifications, and the format. The Coupled Fluid, Energy, and Solute Transport (CFEST) code was applied to solve cases 1, 2, 4, 5, and 7; the Finite Element Three-Dimensional Groundwater (FE3DGW) Flow Model was used to solve case 6. Case 3 has been ignored because unsaturated flow is not pertinent to SRP. This report presents the Level 1 results furnished by the project teams. The numerical accuracy of the codes is determined by (1) comparing the computational results with analytical solutions for cases that have analytical solutions (namely cases 1 and 4), and (2) intercomparing results from codes for cases which do not have analytical solutions (cases 2, 5, 6, and 7). Cases 1, 2, 6, and 7 relate to flow analyses, whereas cases 4 and 5 require nonlinear solutions. 7 refs., 71 figs., 9 tabs
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)
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
Space applications of the MITS electron-photon Monte Carlo transport code system
International Nuclear Information System (INIS)
Kensek, R.P.; Lorence, L.J.; Halbleib, J.A.; Morel, J.E.
1996-01-01
The MITS multigroup/continuous-energy electron-photon Monte Carlo transport code system has matured to the point that it is capable of addressing more realistic three-dimensional adjoint applications. It is first employed to efficiently predict point doses as a function of source energy for simple three-dimensional experimental geometries exposed to simulated uniform isotropic planar sources of monoenergetic electrons up to 4.0 MeV. Results are in very good agreement with experimental data. It is then used to efficiently simulate dose to a detector in a subsystem of a GPS satellite due to its natural electron environment, employing a relatively complex model of the satellite. The capability for survivability analysis of space systems is demonstrated, and results are obtained with and without variance reduction
MCPT: A Monte Carlo code for simulation of photon transport in tomographic scanners
International Nuclear Information System (INIS)
Prettyman, T.H.; Gardner, R.P.; Verghese, K.
1990-01-01
MCPT is a special-purpose Monte Carlo code designed to simulate photon transport in tomographic scanners. Variance reduction schemes and sampling games present in MCPT were selected to characterize features common to most tomographic scanners. Combined splitting and biasing (CSB) games are used to systematically sample important detection pathways. An efficient splitting game is used to tally particle energy deposition in detection zones. The pulse height distribution of each detector can be found by convolving the calculated energy deposition distribution with the detector's resolution function. A general geometric modelling package, HERMETOR, is used to describe the geometry of the tomographic scanners and provide MCPT information needed for particle tracking. MCPT's modelling capabilites are described and preliminary experimental validation is presented. (orig.)
Nupack, the new Asme code for radioactive material transportation packaging containments
International Nuclear Information System (INIS)
Turula, P.
1998-01-01
The American Society of Mechanical Engineers (ASME) has added a new division to the nuclear construction section of its Boiler and Pressure Vessel Code (B and PVC). This Division, commonly referred to as 'Nupack', has been written to provide a consistent set of technical requirements for containment vessels of transportation packagings for high-level radioactive materials. This paper provides an introduction to Nupack, discusses some of its technical provisions, and describes how it can be used the design and construction of packaging components. Nupack's general provisions and design requirements are emphasized, while treatment of materials, fabrication and inspection is left for another paper. Participation in the Nupack development work described in this paper was supported by the U.S. Department of Energy. (authors)
Applications of Transport/Reaction Codes to Problems in Cell Modeling; TOPICAL
International Nuclear Information System (INIS)
MEANS, SHAWN A.; RINTOUL, MARK DANIEL; SHADID, JOHN N.
2001-01-01
We demonstrate two specific examples that show how our exiting capabilities in solving large systems of partial differential equations associated with transport/reaction systems can be easily applied to outstanding problems in computational biology. First, we examine a three-dimensional model for calcium wave propagation in a Xenopus Laevis frog egg and verify that a proposed model for the distribution of calcium release sites agrees with experimental results as a function of both space and time. Next, we create a model of the neuron's terminus based on experimental observations and show that the sodium-calcium exchanger is not the route of sodium's modulation of neurotransmitter release. These state-of-the-art simulations were performed on massively parallel platforms and required almost no modification of existing Sandia codes
Pandya, Tara M.; Johnson, Seth R.; Evans, Thomas M.; Davidson, Gregory G.; Hamilton, Steven P.; Godfrey, Andrew T.
2016-03-01
This work discusses the implementation, capabilities, and validation of Shift, a massively parallel Monte Carlo radiation transport package authored at Oak Ridge National Laboratory. Shift has been developed to scale well from laptops to small computing clusters to advanced supercomputers and includes features such as support for multiple geometry and physics engines, hybrid capabilities for variance reduction methods such as the Consistent Adjoint-Driven Importance Sampling methodology, advanced parallel decompositions, and tally methods optimized for scalability on supercomputing architectures. The scaling studies presented in this paper demonstrate good weak and strong scaling behavior for the implemented algorithms. Shift has also been validated and verified against various reactor physics benchmarks, including the Consortium for Advanced Simulation of Light Water Reactors' Virtual Environment for Reactor Analysis criticality test suite and several Westinghouse AP1000® problems presented in this paper. These benchmark results compare well to those from other contemporary Monte Carlo codes such as MCNP5 and KENO.
Multitasking the three-dimensional transport code TORT on CRAY platforms
International Nuclear Information System (INIS)
Azmy, Y.Y.
1996-01-01
The multitasking options in the three-dimensional neutral particle transport code TORT originally implemented for Cray's CTSS operating system are revived and extended to run on Cray Y/MP and C90 computers using the UNICOS operating system. These include two coarse-grained domain decompositions; across octants, and across directions within an octant, termed Octant Parallel (OP), and Direction Parallel (DP), respectively. Parallel performance of the DP is significantly enhanced by increasing the task grain size and reducing load imbalance via dynamic scheduling of the discrete angles among the participating tasks. Substantial Wall Clock speedup factors, approaching 4.5 using 8 tasks, have been measured in a time-sharing environment, and generally depend on the test problem specifications, number of tasks, and machine loading during execution
Modeling a TRIGA Mark II reactor using the Attila three-dimensional deterministic transport code
International Nuclear Information System (INIS)
Keller, S.T.; Palmer, T.S.; Wareing, T.A.
2005-01-01
A benchmark model of a TRIGA reactor constructed using materials and dimensions similar to existing TRIGA reactors was analyzed using MCNP and the recently developed deterministic transport code Attila TM . The benchmark reactor requires no MCNP modeling approximations, yet is sufficiently complex to validate the new modeling techniques. Geometric properties of the benchmark reactor are specified for use by Attila TM with CAD software. Materials are treated individually in MCNP. Materials used in Attila TM that are clad are homogenized. Attila TM uses multigroup energy discretization. Two cross section libraries were constructed for comparison. A 16 group library collapsed from the SCALE 4.4.a 238 group library provided better results than a seven group library calculated with WIMS-ANL. Values of the k-effective eigenvalue and scalar flux as a function of location and energy were calculated by the two codes. The calculated values for k-effective and spatially averaged neutron flux were found to be in good agreement. Flux distribution by space and energy also agreed well. Attila TM results could be improved with increased spatial and angular resolution and revised energy group structure. (authors)
A pre- and post-processor for the ICOOL muon transport code
International Nuclear Information System (INIS)
Fawley, W.M.
2001-01-01
ICOOL[1] is a Fortran77 macroparticle transport code widely used by researchers to study the front end of a neutrino factory/muon collider[2]. In part due to the desire that ICOOL be usable over multiple computer platforms and operating systems, the code uses simple text files for input/output services. This choice together with user-driven requests for greater and greater choice of lattice element type and configuration has led to ICOOL input decks becoming rather difficult to compose and modify easily. Moreover, the lack of a standard graphical post-processor has prevented many ICOOL users from extracting all but the most simple results from the output files. Here I present two attempts to improve this situation: First, a simple but quite general graphical pre-processor (NIME) written in the Tcl/TK[3] to permit users to write and maintain ASCII-formatted input files by use of simple macro definitions and expansions. Second, an interactive post-processor written in Fortran90 and NCAR graphics, which allows users to define, extract, and then examine the behavior of various particle subsets. In this paper I show some examples of use of both the pre- and post-processor for a standard ICOOL run
Development Of A Parallel Performance Model For The THOR Neutral Particle Transport Code
Energy Technology Data Exchange (ETDEWEB)
Yessayan, Raffi; Azmy, Yousry; Schunert, Sebastian
2017-02-01
The THOR neutral particle transport code enables simulation of complex geometries for various problems from reactor simulations to nuclear non-proliferation. It is undergoing a thorough V&V requiring computational efficiency. This has motivated various improvements including angular parallelization, outer iteration acceleration, and development of peripheral tools. For guiding future improvements to the code’s efficiency, better characterization of its parallel performance is useful. A parallel performance model (PPM) can be used to evaluate the benefits of modifications and to identify performance bottlenecks. Using INL’s Falcon HPC, the PPM development incorporates an evaluation of network communication behavior over heterogeneous links and a functional characterization of the per-cell/angle/group runtime of each major code component. After evaluating several possible sources of variability, this resulted in a communication model and a parallel portion model. The former’s accuracy is bounded by the variability of communication on Falcon while the latter has an error on the order of 1%.
Spallation neutron production and the current intra-nuclear cascade and transport codes
Filges, D.; Goldenbaum, F.; Enke, M.; Galin, J.; Herbach, C.-M.; Hilscher, D.; Jahnke, U.; Letourneau, A.; Lott, B.; Neef, R.-D.; Nünighoff, K.; Paul, N.; Péghaire, A.; Pienkowski, L.; Schaal, H.; Schröder, U.; Sterzenbach, G.; Tietze, A.; Tishchenko, V.; Toke, J.; Wohlmuther, M.
A recent renascent interest in energetic proton-induced production of neutrons originates largely from the inception of projects for target stations of intense spallation neutron sources, like the planned European Spallation Source (ESS), accelerator-driven nuclear reactors, nuclear waste transmutation, and also from the application for radioactive beams. In the framework of such a neutron production, of major importance is the search for ways for the most efficient conversion of the primary beam energy into neutron production. Although the issue has been quite successfully addressed experimentally by varying the incident proton energy for various target materials and by covering a huge collection of different target geometries --providing an exhaustive matrix of benchmark data-- the ultimate challenge is to increase the predictive power of transport codes currently on the market. To scrutinize these codes, calculations of reaction cross-sections, hadronic interaction lengths, average neutron multiplicities, neutron multiplicity and energy distributions, and the development of hadronic showers are confronted with recent experimental data of the NESSI collaboration. Program packages like HERMES, LCS or MCNPX master the prevision of reaction cross-sections, hadronic interaction lengths, averaged neutron multiplicities and neutron multiplicity distributions in thick and thin targets for a wide spectrum of incident proton energies, geometrical shapes and materials of the target generally within less than 10% deviation, while production cross-section measurements for light charged particles on thin targets point out that appreciable distinctions exist within these models.
Criticality coefficient calculation for a small PWR using Monte Carlo Transport Code
Energy Technology Data Exchange (ETDEWEB)
Trombetta, Debora M.; Su, Jian, E-mail: dtrombetta@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Chirayath, Sunil S., E-mail: sunilsc@tamu.edu [Department of Nuclear Engineering and Nuclear Security Science and Policy Institute, Texas A and M University, TX (United States)
2015-07-01
Computational models of reactors are increasingly used to predict nuclear reactor physics parameters responsible for reactivity changes which could lead to accidents and losses. In this work, preliminary results for criticality coefficient calculation using the Monte Carlo transport code MCNPX were presented for a small PWR. The computational modeling developed consists of the core with fuel elements, radial reflectors, and control rods inside a pressure vessel. Three different geometries were simulated, a single fuel pin, a fuel assembly and the core, with the aim to compare the criticality coefficients among themselves.The criticality coefficients calculated were: Doppler Temperature Coefficient, Coolant Temperature Coefficient, Coolant Void Coefficient, Power Coefficient, and Control Rod Worth. The coefficient values calculated by the MCNP code were compared with literature results, showing good agreement with reference data, which validate the computational model developed and allow it to be used to perform more complex studies. Criticality Coefficient values for the three simulations done had little discrepancy for almost all coefficients investigated, the only exception was the Power Coefficient. Preliminary results presented show that simple modelling as a fuel assembly can describe changes at almost all the criticality coefficients, avoiding the need of a complex core simulation. (author)
Spallation neutron production and the current intra-nuclear cascade and transport codes
International Nuclear Information System (INIS)
Filges, D.; Goldenbaum, F.
2001-01-01
A recent renascent interest in energetic proton-induced production of neutrons originates largely from the inception of projects for target stations of intense spallation neutron sources, like the planned European Spallation Source (ESS), accelerator-driven nuclear reactors, nuclear waste transmutation, and also from the application for radioactive beams. In the framework of such a neutron production, of major importance is the search for ways for the most efficient conversion of the primary beam energy into neutron production. Although the issue has been quite successfully addressed experimentally by varying the incident proton energy for various target materials and by covering a huge collection of different target geometries --providing an exhaustive matrix of benchmark data-- the ultimate challenge is to increase the predictive power of transport codes currently on the market. To scrutinize these codes, calculations of reaction cross-sections, hadronic interaction lengths, average neutron multiplicities, neutron multiplicity and energy distributions, and the development of hadronic showers are confronted with recent experimental data of the NESSI collaboration. Program packages like HERMES, LCS or MCNPX master the prevision of reaction cross-sections, hadronic interaction lengths, averaged neutron multiplicities and neutron multiplicity distributions in thick and thin targets for a wide spectrum of incident proton energies, geometrical shapes and materials of the target generally within less than 10% deviation, while production cross-section measurements for light charged particles on thin targets point out that appreciable distinctions exist within these models. (orig.)
EBQ code: transport of space-charge beams in axially symmetric devices
International Nuclear Information System (INIS)
Paul, A.C.
1982-11-01
Such general-purpose space charge codes as EGUN, BATES, WOLF, and TRANSPORT do not gracefully accommodate the simulation of relativistic space-charged beams propagating a long distance in axially symmetric devices where a high degree of cancellation has occurred between the self-magnetic and self-electric forces of the beam. The EBQ code was written specifically to follow high current beam particles where space charge is important in long distance flight in axially symmetric machines possessing external electric and magnetic field. EBQ simultaneously tracks all trajectories so as to allow procedures for charge deposition based on inter-ray separations. The orbits are treated in Cartesian geometry (position and momentum) with z as the independent variable. Poisson's equation is solved in cylindrical geometry on an orthogonal rectangular mesh. EBQ can also handle problems involving multiple ion species where the space charge from each must be included. Such problems arise in the design of ion sources where different charge and mass states are present
International Nuclear Information System (INIS)
Hsiao, Ming-Yuan; Werley, K.A.; Ling, Kuok-Mee.
1988-05-01
A one-and-a-quarter-dimensional transport code, which includes radial as well as some two-dimensional effects for field-reversed configurations, is described. The set of transport equations is transformed to a set of new independent and dependent variables and is solved as a coupled initial-boundary value problem. The code simulation includes both the closed and open field regions. The axial effects incorporated include global axial force balance, axial losses in the open field region, and flux surface averaging over the closed field region. Input, output, and structure of the code are described in detail. A typical example of the code results is also given. 20 refs., 21 figs., 7 tabs
Penelope - a code system for Monte Carlo simulation of electron and photon transport
International Nuclear Information System (INIS)
2003-01-01
Radiation is used in many applications of modern technology. Its proper handling requires competent knowledge of the basic physical laws governing its interaction with matter. To ensure its safe use, appropriate tools for predicting radiation fields and doses, as well as pertinent regulations, are required. One area of radiation physics that has received much attention concerns electron-photon transport in matter. PENELOPE is a modern, general-purpose Monte Carlo tool for simulating the transport of electrons and photons, which is applicable for arbitrary materials and in a wide energy range. PENELOPE provides quantitative guidance for many practical situations and techniques, including electron and X-ray spectroscopies, electron microscopy and microanalysis, biophysics, dosimetry, medical diagnostics and radiotherapy, as well as radiation damage and shielding. These proceedings contain the extensively revised teaching notes of the second workshop/training course on PENELOPE held in 2003, along with a detailed description of the improved physic models, numerical algorithms and structure of the code system. (author)
International Nuclear Information System (INIS)
Mosca, P.
2009-12-01
The deterministic transport codes solve the stationary Boltzmann equation in a discretized energy formalism called multigroup. The transformation of continuous data in a multigroup form is obtained by averaging the highly variable cross sections of the resonant isotopes with the solution of the self-shielding models and the remaining ones with the coarse energy spectrum of the reactor type. So far the error of such an approach could only be evaluated retrospectively. To remedy this, we studied in this thesis a set of methods to control a priori the accuracy and the cost of the multigroup transport computation. The energy mesh optimisation is achieved using a two step process: the creation of a reference mesh and its optimized condensation. In the first stage, by refining locally and globally the energy mesh, we seek, on a fine energy mesh with subgroup self-shielding, a solution equivalent to a reference solver (Monte Carlo or pointwise deterministic solver). In the second step, once fixed the number of groups, depending on the acceptable computational cost, and chosen the most appropriate self-shielding models to the reactor type, we look for the best bounds of the reference mesh minimizing reaction rate errors by the particle swarm optimization algorithm. This new approach allows us to define new meshes for fast reactors as accurate as the currently used ones, but with fewer groups. (author)
Energy Technology Data Exchange (ETDEWEB)
Kostin, Mikhail [Michigan State Univ., East Lansing, MI (United States); Mokhov, Nikolai [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Niita, Koji [Research Organization for Information Science and Technology, Ibaraki-ken (Japan)
2013-09-25
A parallel computing framework has been developed to use with general-purpose radiation transport codes. The framework was implemented as a C++ module that uses MPI for message passing. It is intended to be used with older radiation transport codes implemented in Fortran77, Fortran 90 or C. The module is significantly independent of radiation transport codes it can be used with, and is connected to the codes by means of a number of interface functions. The framework was developed and tested in conjunction with the MARS15 code. It is possible to use it with other codes such as PHITS, FLUKA and MCNP after certain adjustments. Besides the parallel computing functionality, the framework offers a checkpoint facility that allows restarting calculations with a saved checkpoint file. The checkpoint facility can be used in single process calculations as well as in the parallel regime. The framework corrects some of the known problems with the scheduling and load balancing found in the original implementations of the parallel computing functionality in MARS15 and PHITS. The framework can be used efficiently on homogeneous systems and networks of workstations, where the interference from the other users is possible.
ITS Version 3.0: The Integrated TIGER Series of coupled electron/photon Monte Carlo transport codes
International Nuclear Information System (INIS)
Halbleib, J.A.; Kensek, R.P.; Valdez, G.D.; Mehlhorn, T.A.; Seltzer, S.M.; Berger, M.J.
1993-01-01
ITS is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of linear time-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields. It combines operational simplicity and physical accuracy in order to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Flexibility of construction permits tailoring of the codes to specific applications and extension of code capabilities to more complex applications through simple update procedures
Monte-Carlo Impurity transport simulations in the edge of the DIII-D tokamak using the MCI code
International Nuclear Information System (INIS)
Evans, T.E.; Mahdavi, M.A.; Sager, G.T.; West, W.P.; Fenstermacher, M.E.; Meyer, W.H.; Porter, G.D.
1995-07-01
A Monte-Carlo Impurity (MCI) transport code is used to follow trace impurities through multiple ionization states in realistic 2-D tokamak geometries. The MCI code is used to study impurity transport along the open magnetic field lines of the Scrape-off Layer (SOL) and to understand how impurities get into the core from the SOL. An MCI study concentrating on the entrainment of carbon impurities ions by deuterium background plasma into the DIII-D divertor is discussed. MCI simulation results are compared to experimental DIII-D carbon measurements
Simunek, J.; Jacques, D.; Mayer, K. U.; Gerard, F.
2016-12-01
A large number of organic matter degradation, CO2 transport and dissolved organic matter models have been developed during the last decades. However, organic matter degradation models are in many cases hard-coded in terms of pools, kinetics and dependency on environmental variables. The input of the model user is typically limited to the adjustment of input parameters. In addition, the coupling with geochemical soil processes including aqueous speciation, sorption and colloid-facilitated transport are not incorporated in many of these models. Furthermore, these models are combined with simplified representations of flow and transport processes. We illustrate the capability of generic reactive transport codes to overcome these shortcomings. The formulations of reactive transport codes include a physics-based continuum representation of flow and transport processes, while biogeochemical reactions can be described as equilibrium processes and/or kinetic reaction networks. The flexibility of these type of codes allows for straightforward extension of reaction networks with new model components and in such a way facilitates an application-tailored implementation of organic matter degradation models and related processes. A numerical benchmark involving two reactive transport codes (HPx and MIN3P) demonstrates how the process-based simulation of transient variably saturated water flow, solute transport, heat transfer and diffusion in the gas phase can be combined with a flexible implementation of a soil organic matter degradation model. The benchmark includes the production of leachable organic matter and inorganic carbon in the aqueous and gaseous phases, as well as different decomposition functions with first-order, linear dependence or nonlinear dependence on a biomass pool. In addition, we show how processes such as local bioturbation (biodiffusion) can be included implicitly through a Fickian formulation of transport of soil organic matter. Coupling soil organic
Development of a 1.5D plasma transport code for coupling to full orbit runaway electron simulations
Lore, J. D.; Del Castillo-Negrete, D.; Baylor, L.; Carbajal, L.
2017-10-01
A 1.5D (1D radial transport + 2D equilibrium geometry) plasma transport code is being developed to simulate runaway electron generation, mitigation, and avoidance by coupling to the full-orbit kinetic electron transport code KORC. The 1.5D code solves the time-dependent 1D flux surface averaged transport equations with sources for plasma density, pressure, and poloidal magnetic flux, along with the Grad-Shafranov equilibrium equation for the 2D flux surface geometry. Disruption mitigation is simulated by introducing an impurity neutral gas `pellet', with impurity densities and electron cooling calculated from ionization, recombination, and line emission rate coefficients. Rapid cooling of the electrons increases the resistivity, inducing an electric field which can be used as an input to KORC. The runaway electron current is then included in the parallel Ohm's law in the transport equations. The 1.5D solver will act as a driver for coupled simulations to model effects such as timescales for thermal quench, runaway electron generation, and pellet impurity mixtures for runaway avoidance. Current progress on the code and details of the numerical algorithms will be presented. Work supported by the US DOE under DE-AC05-00OR22725.
Energy Technology Data Exchange (ETDEWEB)
Walsh, J. A. [Department of Nuclear Science and Engineering, Massachusetts Institute of Technology, NW12-312 Albany, St. Cambridge, MA 02139 (United States); Palmer, T. S. [Department of Nuclear Engineering and Radiation Health Physics, Oregon State University, 116 Radiation Center, Corvallis, OR 97331 (United States); Urbatsch, T. J. [XTD-5: Air Force Systems, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)
2013-07-01
A new method for generating discrete scattering cross sections to be used in charged particle transport calculations is investigated. The method of data generation is presented and compared to current methods for obtaining discrete cross sections. The new, more generalized approach allows greater flexibility in choosing a cross section model from which to derive discrete values. Cross section data generated with the new method is verified through a comparison with discrete data obtained with an existing method. Additionally, a charged particle transport capability is demonstrated in the time-dependent Implicit Monte Carlo radiative transfer code package, Milagro. The implementation of this capability is verified using test problems with analytic solutions as well as a comparison of electron dose-depth profiles calculated with Milagro and an already-established electron transport code. An initial investigation of a preliminary integration of the discrete cross section generation method with the new charged particle transport capability in Milagro is also presented. (authors)
Present status of transport code development based on Monte Carlo method
International Nuclear Information System (INIS)
Nakagawa, Masayuki
1985-01-01
The present status of development in Monte Carlo code is briefly reviewed. The main items are the followings; Application fields, Methods used in Monte Carlo code (geometry spectification, nuclear data, estimator and variance reduction technique) and unfinished works, Typical Monte Carlo codes and Merits of continuous energy Monte Carlo code. (author)
International Nuclear Information System (INIS)
Calloo, A.A.
2012-01-01
In reactor physics, calculation schemes with deterministic codes are validated with respect to a reference Monte Carlo code. The remaining biases are attributed to the approximations and models induced by the multigroup theory (self-shielding models and expansion of the scattering law using Legendre polynomials) to represent physical phenomena (resonant absorption and scattering anisotropy respectively). This work focuses on the relevance of a polynomial expansion to model the scattering law. Since the outset of reactor physics, the latter has been expanded on a truncated Legendre polynomial basis. However, the transfer cross sections are highly anisotropic, with non-zero values for a very small range of the cosine of the scattering angle. Besides, the finer the energy mesh and the lighter the scattering nucleus, the more exacerbated is the peaked shape of this cross section. As such, the Legendre expansion is less suited to represent the scattering law. Furthermore, this model induces negative values which are non-physical. In this work, various scattering laws are briefly described and the limitations of the existing model are pointed out. Hence, piecewise-constant functions have been used to represent the multigroup scattering cross section. This representation requires a different model for the diffusion source. The discrete ordinates method which is widely employed to solve the transport equation has been adapted. Thus, the finite volume method for angular discretization has been developed and implemented in Paris environment which hosts the S n solver, Snatch. The angular finite volume method has been compared to the collocation method with Legendre moments to ensure its proper performance. Moreover, unlike the latter, this method is adapted for both the Legendre moments and the piecewise-constant functions representations of the scattering cross section. This hybrid-source method has been validated for different cases: fuel cell in infinite lattice
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
International Nuclear Information System (INIS)
Ahnert, C.; Aragones, J. M.
1981-01-01
This Is a users manual of the neutron transport code TWOTRAN-TRACA, which is a version of the original TWOTRAN-GG from the Los Alamos Laboratory, with some modifications made at JEN. A detailed input data description is given as well as the new modifications developed at JEN. (Author) 8 refs
Energy Technology Data Exchange (ETDEWEB)
Carneiro, Ana [Vanderbilt University; Airey, David [University of Tennessee Health Science Center, Memphis; Thompson, Brent [Vanderbilt University; Zhu, C [Vanderbilt University; Rinchik, Eugene M [ORNL; Lu, Lu [University of Tennessee Health Science Center, Memphis; Chesler, Elissa J [ORNL; Erikson, Keith [University of North Carolina; Blakely, Randy [Vanderbilt University
2009-01-01
The human serotonin (5-hydroxytryptamine, 5-HT) transporter (hSERT, SLC6A4) figures prominently in the etiology or treatment of many prevalent neurobehavioral disorders including anxiety, alcoholism, depression, autism and obsessive-compulsive disorder (OCD). Here we utilize naturally occurring polymorphisms in recombinant inbred (RI) lines to identify novel phenotypes associated with altered SERT function. The widely used mouse strain C57BL/6J, harbors a SERT haplotype defined by two nonsynonymous coding variants (Gly39 and Lys152 (GK)). At these positions, many other mouse lines, including DBA/2J, encode Glu39 and Arg152 (ER haplotype), assignments found also in hSERT. Synaptosomal 5-HT transport studies revealed reduced uptake associated with the GK variant. Heterologous expression studies confirmed a reduced SERT turnover rate for the GK variant. Experimental and in silico approaches using RI lines (C57Bl/6J X DBA/2J=BXD) identifies multiple anatomical, biochemical and behavioral phenotypes specifically impacted by GK/ER variation. Among our findings are multiple traits associated with anxiety and alcohol consumption, as well as of the control of dopamine (DA) signaling. Further bioinformatic analysis of BXD phenotypes, combined with biochemical evaluation of SERT knockout mice, nominates SERT-dependent 5-HT signaling as a major determinant of midbrain iron homeostasis that, in turn, dictates ironregulated DA phenotypes. Our studies provide a novel example of the power of coordinated in vitro, in vivo and in silico approaches using murine RI lines to elucidate and quantify the system-level impact of gene variation.
Carneiro, Ana M D; Airey, David C; Thompson, Brent; Zhu, Chong-Bin; Lu, Lu; Chesler, Elissa J; Erikson, Keith M; Blakely, Randy D
2009-02-10
The human serotonin (5-hydroxytryptamine, 5-HT) transporter (hSERT, SLC6A4) figures prominently in the etiology and treatment of many prevalent neurobehavioral disorders including anxiety, alcoholism, depression, autism, and obsessive-compulsive disorder (OCD). Here, we use naturally occurring polymorphisms in recombinant inbred (RI) lines to identify multiple phenotypes associated with altered SERT function. The widely used mouse strain C57BL/6J, harbors a SERT haplotype defined by 2 nonsynonymous coding variants [Gly-39 and Lys-152 (GK)]. At these positions, many other mouse lines, including DBA/2J, encode, respectively, Glu-39 and Arg-152 (ER haplotype), amino acids found also in hSERT. Ex vivo synaptosomal 5-HT transport studies revealed reduced uptake associated with the GK variant, a finding confirmed by in vitro heterologous expression studies. Experimental and in silico approaches using RI lines (C57BL/6J x DBA/2J = BXD) identify multiple anatomical, biochemical, and behavioral phenotypes specifically impacted by GK/ER variation. Among our findings are several traits associated with alcohol consumption and multiple traits associated with dopamine signaling. Further bioinformatic analysis of BXD phenotypes, combined with biochemical evaluation of SERT knockout mice, nominates SERT-dependent 5-HT signaling as a major determinant of midbrain iron homeostasis that, in turn, dictates iron-regulated DA phenotypes. Our studies provide an example of the power of coordinated in vitro, in vivo, and in silico approaches using mouse RI lines to elucidate and quantify the system-level impact of gene variation.
Del Sorbo, Dario; Brodrick, Jonathan P.; Read, Martin P.; Holec, Milan; Debayle, Arnaud; Loiseau, Pascal; Kingham, Robert J.; Nicolai, Philippe; Feugeas, Jean-Luc; Tikhonchuk, Vladimir T.; Ridgers, Christopher P.
2017-10-01
Hydrodynamics simulations relevant to inertial confinement fusion require a detailed description of energy transport, in particular by electrons. This may be nonlocal if, as is commonly the case, the plasma is not in local thermodynamic equilibrium (i.e. if the electron mean free path is long compared to the temperature scale-length). In this case, a kinetic model of electron thermal transport is required. Some of the most successful approaches to nonlocal transport (SNB & M1 models) are systematically compared against Vlasov-Foker-Planck & Particle-in-Cell codes, extending benchmarking beyond the 1D unmagnetized case and studying situations of immediate relevance to ICF.
Modelling of a general purpose irradiation chamber using a Monte Carlo particle transport code
International Nuclear Information System (INIS)
Dhiyauddin Ahmad Fauzi; Sheik, F.O.A.; Nurul Fadzlin Hasbullah
2013-01-01
Full-text: The aim of this research is to stimulate the effectiveness use of a general purpose irradiation chamber to contain pure neutron particles obtained from a research reactor. The secondary neutron and gamma particles dose discharge from the chamber layers will be used as a platform to estimate the safe dimension of the chamber. The chamber, made up of layers of lead (Pb), shielding, polyethylene (PE), moderator and commercial grade aluminium (Al) cladding is proposed for the use of interacting samples with pure neutron particles in a nuclear reactor environment. The estimation was accomplished through simulation based on general Monte Carlo N-Particle transport code using Los Alamos MCNPX software. Simulations were performed on the model of the chamber subjected to high neutron flux radiation and its gamma radiation product. The model of neutron particle used is based on the neutron source found in PUSPATI TRIGA MARK II research reactor which holds a maximum flux value of 1 x 10 12 neutron/ cm 2 s. The expected outcomes of this research are zero gamma dose in the core of the chamber and neutron dose rate of less than 10 μSv/ day discharge from the chamber system. (author)
Revision of Collisional-Radiative Models and Neutral-Transport Code for Hydrogen and Helium Species
International Nuclear Information System (INIS)
Sawada, Keiji; Goto, Motoshi
2013-01-01
We have been developing collisional-radiative models and a neutral-transport code for hydrogen and helium species, which are used to investigate fusion plasmas. Collisional-radiative models of atomic hydrogen and helium have been applied to a helium-hydrogen RF plasma at Shinshu University, Japan, to test whether these models reproduce the observed emission intensities. The electron temperature and density are determined from visible emission line intensities of helium atom considering photoexcitation from the ground state to singlet P states, which is accompanied by radiation trapping. From the observed hydrogen Balmer γ line intensity, which is hardly affected by photoexcitation, the atomic hydrogen density is determined using a hydrogen collisional-radiative model that ignores photoexcitation. The atomic hydrogen temperature, which reproduces Balmer α and β line intensities, is determined using an iterative hydrogen atom collisional-radiative model that calculates photoexcitation rates. R-Matrix cross sections for n≤5 are used in the model. The hope is hoped that precise cross sections for higher-lying levels will be produced to determine the atomic density in fusion plasmas
A massively parallel method of characteristic neutral particle transport code for GPUs
International Nuclear Information System (INIS)
Boyd, W. R.; Smith, K.; Forget, B.
2013-01-01
Over the past 20 years, parallel computing has enabled computers to grow ever larger and more powerful while scientific applications have advanced in sophistication and resolution. This trend is being challenged, however, as the power consumption for conventional parallel computing architectures has risen to unsustainable levels and memory limitations have come to dominate compute performance. Heterogeneous computing platforms, such as Graphics Processing Units (GPUs), are an increasingly popular paradigm for solving these issues. This paper explores the applicability of GPUs for deterministic neutron transport. A 2D method of characteristics (MOC) code - OpenMOC - has been developed with solvers for both shared memory multi-core platforms as well as GPUs. The multi-threading and memory locality methodologies for the GPU solver are presented. Performance results for the 2D C5G7 benchmark demonstrate 25-35 x speedup for MOC on the GPU. The lessons learned from this case study will provide the basis for further exploration of MOC on GPUs as well as design decisions for hardware vendors exploring technologies for the next generation of machines for scientific computing. (authors)
International Nuclear Information System (INIS)
Orsi, R.
2003-01-01
Bot3p consists of a set of standard Fortran 77 language programs that gives the users of the deterministic transport codes Dort and Tort some useful diagnostic tools to prepare and check the geometry of their input data files for both Cartesian and cylindrical geometries including graphical display modules. Bot3p produces at the same time the geometrical and material distribution data for the deterministic transport codes Twodant and Threedant and, only in three-dimensional (3D) Cartesian geometry, for the Monte Carlo Transport Code MCNP. This makes it possible to compare directly for the same geometry the effects stemming from the use of different data libraries and solution approaches on transport analysis results. Through the use of Bot3p, radiation transport problems with complex 3D geometrical structures can be modelled easily, as a relatively small amount of engineer-time is required and refinement is achieved by changing few parameters. This tool is useful for solving very large challenging problems. (author)
International Nuclear Information System (INIS)
Rahatgaonkar, P. S.; Datta, D.; Malhotra, P. K.; Ghadge, S. G.
2012-01-01
Prediction of groundwater movement and contaminant transport in soil is an important problem in many branches of science and engineering. This includes groundwater hydrology, environmental engineering, soil science, agricultural engineering and also nuclear engineering. Specifically, in nuclear engineering it is applicable in the design of spent fuel storage pools and waste management sites in the nuclear power plants. Ground water modeling involves the simulation of flow and contaminant transport by groundwater flow. In the context of contaminated soil and groundwater system, numerical simulations are typically used to demonstrate compliance with regulatory standard. A one-dimensional Computational Fluid Dynamics code GFLOW had been developed based on the Finite Difference Method for simulating groundwater flow and contaminant transport through saturated and unsaturated soil. The code is validated with the analytical model and the benchmarking cases available in the literature. (authors)
International Nuclear Information System (INIS)
Yamano, Naoki; Minami, Kazuyoshi; Koyama, Kinji; Naito, Yoshitaka.
1989-03-01
A modular code system RADHEAT-V4 has been developed for performing precisely neutron and photon transport analyses, and shielding safety evaluations. The system consists of the functional modules for producing coupled multi-group neutron and photon cross section sets, for analyzing the neutron and photon transport, and for calculating the atom displacement and the energy deposition due to radiations in nuclear reactor or shielding material. A precise method named Direct Angular Representation (DAR) has been developed for eliminating an error associated with the method of the finite Legendre expansion in evaluating angular distributions of cross sections and radiation fluxes. The DAR method implemented in the code system has been described in detail. To evaluate the accuracy and applicability of the code system, some test calculations on strong anisotropy problems have been performed. From the results, it has been concluded that RADHEAT-V4 is successfully applicable to evaluating shielding problems accurately for fission and fusion reactors and radiation sources. The method employed in the code system is very effective in eliminating negative values and oscillations of angular fluxes in a medium having an anisotropic source or strong streaming. Definitions of the input data required in various options of the code system and the sample problems are also presented. (author)
Zamani, K.; Bombardelli, F. A.
2014-12-01
Verification of geophysics codes is imperative to avoid serious academic as well as practical consequences. In case that access to any given source code is not possible, the Method of Manufactured Solution (MMS) cannot be employed in code verification. In contrast, employing the Method of Exact Solution (MES) has several practical advantages. In this research, we first provide four new one-dimensional analytical solutions designed for code verification; these solutions are able to uncover the particular imperfections of the Advection-diffusion-reaction equation, such as nonlinear advection, diffusion or source terms, as well as non-constant coefficient equations. After that, we provide a solution of Burgers' equation in a novel setup. Proposed solutions satisfy the continuity of mass for the ambient flow, which is a crucial factor for coupled hydrodynamics-transport solvers. Then, we use the derived analytical solutions for code verification. To clarify gray-literature issues in the verification of transport codes, we designed a comprehensive test suite to uncover any imperfection in transport solvers via a hierarchical increase in the level of tests' complexity. The test suite includes hundreds of unit tests and system tests to check vis-a-vis the portions of the code. Examples for checking the suite start by testing a simple case of unidirectional advection; then, bidirectional advection and tidal flow and build up to nonlinear cases. We design tests to check nonlinearity in velocity, dispersivity and reactions. The concealing effect of scales (Peclet and Damkohler numbers) on the mesh-convergence study and appropriate remedies are also discussed. For the cases in which the appropriate benchmarks for mesh convergence study are not available, we utilize symmetry. Auxiliary subroutines for automation of the test suite and report generation are designed. All in all, the test package is not only a robust tool for code verification but it also provides comprehensive
Energy Technology Data Exchange (ETDEWEB)
Huet, B.M.; Prevost, J.H.; Scherer, G.W. [Princeton Univ., NJ (United States)
2007-07-01
A modular reactive transport code is proposed to analyze the reactivity of cement in CO{sub 2} saturated brine. The coupling of the transport module and the geochemical module within Dynaflow{sup TM} is derived. Both modules are coupled in a sequential iterative approach to accurately model: (1) mineral dissolution/precipitation and (2) porosity dependent transport properties. Results of the model reproduce qualitatively the dissolution of cement hydrates (C-H, C-S-H, AFm, AFt) and intermediate products (CaCO{sub 3}) into the brine. Slight discrepancies between modeling and experimental results were found concerning the dynamics of the mineral zoning. Results suggest that the power law relationship to model effective transport properties from porosity values is not accurate for very reactive case. (authors)
International Nuclear Information System (INIS)
Huet, B.M.; Prevost, J.H.; Scherer, G.W.
2007-01-01
A modular reactive transport code is proposed to analyze the reactivity of cement in CO 2 saturated brine. The coupling of the transport module and the geochemical module within Dynaflow TM is derived. Both modules are coupled in a sequential iterative approach to accurately model: (1) mineral dissolution/precipitation and (2) porosity dependent transport properties. Results of the model reproduce qualitatively the dissolution of cement hydrates (C-H, C-S-H, AFm, AFt) and intermediate products (CaCO 3 ) into the brine. Slight discrepancies between modeling and experimental results were found concerning the dynamics of the mineral zoning. Results suggest that the power law relationship to model effective transport properties from porosity values is not accurate for very reactive case. (authors)
LASER-R a computer code for reactor cell and burnup calculations in neutron transport theory
International Nuclear Information System (INIS)
Cristian, I.; Cirstoiu, B.; Dumitrache, I.; Cepraga, D.
1976-04-01
The LASER-R code is an IBM 370/135 version of the Westinghouse code, LASER, based on the THERMOS and MUFT codes developped by Poncelet. It can be used to perform thermal reactor cell calculations and burnup calculations. The cell exhibits 3-4 concentric areas: fuel, cladding, moderator and scattering ring. Besides directions for use, a short description of the physical model, numerical methods and output is presented
International Nuclear Information System (INIS)
Jevremovic, Tatjana; Hursin, Mathieu; Satvat, Nader; Hopkins, John; Xiao, Shanjie; Gert, Godfree
2006-01-01
The AGENT (Arbitrary Geometry Neutron Transport) an open-architecture reactor modeling tool is deterministic neutron transport code for two or three-dimensional heterogeneous neutronic design and analysis of the whole reactor cores regardless of geometry types and material configurations. The AGENT neutron transport methodology is applicable to all generations of nuclear power and research reactors. It combines three theories: (1) the theory of R-functions used to generate real three-dimensional whole-cores of square, hexagonal or triangular cross sections, (2) the planar method of characteristics used to solve isotropic neutron transport in non-homogenized 2D) reactor slices, and (3) the one-dimensional diffusion theory used to couple the planar and axial neutron tracks through the transverse leakage and angular mesh-wise flux values. The R-function-geometrical module allows a sequential building of the layers of geometry and automatic sub-meshing based on the network of domain functions. The simplicity of geometry description and selection of parameters for accurate treatment of neutron propagation is achieved through the Boolean algebraic hierarchically organized simple primitives into complex domains (both being represented with corresponding domain functions). The accuracy is comparable to Monte Carlo codes and is obtained by following neutron propagation through real geometrical domains that does not require homogenization or simplifications. The efficiency is maintained through a set of acceleration techniques introduced at all important calculation levels. The flux solution incorporates power iteration with two different acceleration techniques: Coarse Mesh Re-balancing (CMR) and Coarse Mesh Finite Difference (CMFD). The stand-alone originally developed graphical user interface of the AGENT code design environment allows the user to view and verify input data by displaying the geometry and material distribution. The user can also view the output data such
International Nuclear Information System (INIS)
Avery, A.F.; Locke, H.F.
1992-03-01
In 1985 the Reactor Physics Committee of the Nuclear Energy Agency initiated an intercomparison of codes for the calculation of the performance of shielding for the transportation of spent reactor fuel. The results of the application of a range of codes to the prediction of the dose-rates in the four theoretical benchmarks set to examine the attenuation of radiation through a variety of cask geometries are presented in this report. The contributions from neutrons, fission product gamma-rays and secondary gamma-rays are tabulated separately, and grouped according to the type of method of calculation employed. A brief discussion is included for each set of results, and overall comparisons of the methods, codes, and nuclear data are made. A number of conclusions are drawn on the advantages and disadvantages of the various methods of calculation, based upon the results of their application to these four benchmark problems
Aurora T: a Monte Carlo code for transportation of neutral atoms in a toroidal plasma
International Nuclear Information System (INIS)
Bignami, A.; Chiorrini, R.
1982-01-01
This paper contains a short description of Aurora code. This code have been developed at Princeton with Monte Carlo method for calculating neutral gas in cylindrical plasma. In this work subroutines such one can take in account toroidal geometry are developed
Representation of inhomogeneities in the flow and transport codes d3f and r3t
International Nuclear Information System (INIS)
Schneider, Anke
2013-09-01
The codes d 3 f and r 3 t are well established for modelling density-driven flow and nuclide transport in the far field of repositories for hazardous material in deep geological formations. While originally intended to be applied to the overburden of a salt dome they were adapted to alternative host media such as crystalline rock or mudstone by including fractures into an otherwise porous medium. However, only discrete fractures or fracture networks with a rather limited number of fractures could be dealt with. Networks of smaller fractures - so-called background fractures - can easily consist of hundreds and thousands of significant individual fractures in a model domain and were therefore beyond the scope of d 3 f and r 3 t. One way to circumvent this problem is to replace a discrete fracture network with an equivalent porous medium. While this is a task in itself the codes had also numerically adapted to be to cope with the new methods. This report describes approaches and results of this work. In groundwater flow simulation fractures are usually modelled as lower dimensional objects. But especially in the case of density driven flow situations may occur where the validity of this assumption has to be proved. Here a special approach was developed and implemented that allows an adaptive resolution of the layers. Of central relevance in this respect is the development of local refinement or coarsening criteria, an adaptive discretisation that allows an adaptive transition from low-dimensional to equidimensional modelling of the fractures, and an adaptive multigrid algorithm Furthermore, discretisation methods of higher order for the mixed parabolic-hyperbolic problems were developed. New filtering algebraic multigrid methods as efficient solvers for the large linear equation systems were implemented. The parallelisation was improved by implementation of a parallel communication layer (pcl). For the estimation of parameters for these systems by inverse modelling
International Nuclear Information System (INIS)
Grieshemer, D.P.; Gill, D.F.; Nease, B.R.; Carpenter, D.C.; Joo, H.; Millman, D.L.; Sutton, T.M.; Stedry, M.H.; Dobreff, P.S.; Trumbull, T.H.; Caro, E.
2013-01-01
MC21 is a continuous-energy Monte Carlo radiation transport code for the calculation of the steady-state spatial distributions of reaction rates in three-dimensional models. The code supports neutron and photon transport in fixed source problems, as well as iterated-fission-source (eigenvalue) neutron transport problems. MC21 has been designed and optimized to support large-scale problems in reactor physics, shielding, and criticality analysis applications. The code also supports many in-line reactor feedback effects, including depletion, thermal feedback, xenon feedback, eigenvalue search, and neutron and photon heating. MC21 uses continuous-energy neutron/nucleus interaction physics over the range from 10 -5 eV to 20 MeV. The code treats all common neutron scattering mechanisms, including fast-range elastic and non-elastic scattering, and thermal- and epithermal-range scattering from molecules and crystalline materials. For photon transport, MC21 uses continuous-energy interaction physics over the energy range from 1 keV to 100 GeV. The code treats all common photon interaction mechanisms, including Compton scattering, pair production, and photoelectric interactions. All of the nuclear data required by MC21 is provided by the NDEX system of codes, which extracts and processes data from EPDL-, ENDF-, and ACE-formatted source files. For geometry representation, MC21 employs a flexible constructive solid geometry system that allows users to create spatial cells from first- and second-order surfaces. The system also allows models to be built up as hierarchical collections of previously defined spatial cells, with interior detail provided by grids and template overlays. Results are collected by a generalized tally capability which allows users to edit integral flux and reaction rate information. Results can be collected over the entire problem or within specific regions of interest through the use of phase filters that control which particles are allowed to score each
TART96: a coupled neutron-photon 3-D, combinatorial geometry Monte Carlo transport code
International Nuclear Information System (INIS)
Cullen, D.E.
1996-11-01
The original TARTND has been used and distributed from LLNL for many years. TART95, released in July 1995, was the first version of the code designed to be used on virtually any computer. TART96 is designed to extend the general utility of the code to more areas of application, by concentrating on improving the physics used by the code. TART96 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART96 and its data files
MCNP: a general Monte Carlo code for neutron and photon transport. Version 3A. Revision 2
International Nuclear Information System (INIS)
Briesmeister, J.F.
1986-09-01
This manual is a practical guide for the use of our general-purpose Monte Carlo code MCNP. The first chapter is a primer for the novice user. The second chapter describes the mathematics, data, physics, and Monte Carlo simulation found in MCNP. This discussion is not meant to be exhaustive - details of the particular techniques and of the Monte Carlo method itself will have to be found elsewhere. The third chapter shows the user how to prepare input for the code. The fourth chapter contains several examples, and the fifth chapter explains the output. The appendices show how to use MCNP on particular computer systems at the Los Alamos National Laboratory and also give details about some of the code internals that those who wish to modify the code may find useful. 57 refs
An Eulerian transport-dispersion model of passive effluents: the Difeul code
International Nuclear Information System (INIS)
Wendum, D.
1994-11-01
R and D has decided to develop an Eulerian diffusion model easy to adapt to meteorological data coming from different sources: for instance the ARPEGE code of Meteo-France or the MERCURE code of EDF. We demand this in order to be able to apply the code in independent cases: a posteriori studies of accidental releases from nuclear power plants ar large or medium scale, simulation of urban pollution episodes within the ''Reactive Atmospheric Flows'' research project. For simplicity reasons, the numerical formulation of our code is the same as the one used in Meteo-France's MEDIA model. The numerical tests presented in this report show the good performance of those schemes. In order to illustrate the method by a concrete example a fictitious release from Saint-Laurent has been simulated at national scale: the results of this simulation agree quite well with those of the trajectory model DIFTRA. (author). 6 figs., 4 tabs
International Nuclear Information System (INIS)
Rockhold, M.L.; Wurstner, S.K.
1991-03-01
The objective of this work was to test the ability of the PORFLO-3 computer code to simulate water infiltration and solute transport in dry soils. Data from a field-scale unsaturated zone flow and transport experiment, conducted near Las Cruces, New Mexico, were used for model validation. A spatial moment analysis was used to provide a quantitative basis for comparing the mean simulated and observed flow behavior. The scope of this work was limited to two-dimensional simulations of the second experiment at the Las Cruces trench site. Three simulation cases are presented. The first case represents a uniform soil profile, with homogeneous, isotropic hydraulic and transport properties. The second and third cases represent single stochastic realizations of randomly heterogeneous hydraulic conductivity fields, generated from the cumulative probability distribution of the measured data. Two-dimensional simulations produced water content changes that matched the observed data reasonably well. Models that explicitly incorporated heterogeneous hydraulic conductivity fields reproduced the characteristics of the observed data somewhat better than a uniform, homogeneous model. Improved predictions of water content changes at specific spatial locations were obtained by adjusting the soil hydraulic properties. The results of this study should only be considered a qualitative validation of the PORFLO-3 code. However, the results of this study demonstrate the importance of site-specific data for model calibration. Applications of the code for waste management and remediation activities will require site-specific data for model calibration before defensible predictions of unsaturated flow and containment transport can be made. 23 refs., 16 figs., 3 tabs
International Nuclear Information System (INIS)
Chepe P, M.; Xolocostli M, J. V.; Gomez T, A. M.; Del Valle G, E.
2015-09-01
The deterministic transport codes for analysis of nuclear reactors have been used for several years already, these codes have evolved in terms of the methodology used and the degree of accuracy, because at the present time has more computer power. In this paper, the transport code used considers the classical technique of multi-group for discretization energy, for space discretization uses the nodal methods, while for the angular discretization the discrete ordinates method is used; so that presents the development and implementation of a set of numerical quadratures of SQ N type symmetrical with the same weight for each angular direction and these are compared with the quadratures of EQ N type. The two sets of numerical quadratures were implemented in the program AZTRAN to a problem with isotropic medium in XYZ geometry, in steady state using the nodal method RTN-0 (Raviart-Thomas-Nedelec). The analyzed results correspond to the effective multiplication factor k eff and neutron angular flux with approximations from S 4 to S 16 . (Author)
International Nuclear Information System (INIS)
Eslinger, Paul W.; Engel, David W.; Gerhardstein, Lawrence H.; Lopresti, Charles A.; Nichols, William E.; Strenge, Dennis L.
2001-12-01
One activity of the Department of Energy's Groundwater/Vadose Zone Integration Project is an assessment of cumulative impacts from Hanford Site wastes on the subsurface environment and the Columbia River. Through the application of a system assessment capability (SAC), decisions for each cleanup and disposal action will be able to take into account the composite effect of other cleanup and disposal actions. The SAC has developed a suite of computer programs to simulate the migration of contaminants (analytes) present on the Hanford Site and to assess the potential impacts of the analytes, including dose to humans, socio-cultural impacts, economic impacts, and ecological impacts. The general approach to handling uncertainty in the SAC computer codes is a Monte Carlo approach. Conceptually, one generates a value for every stochastic parameter in the code (the entire sequence of modules from inventory through transport and impacts) and then executes the simulation, obtaining an output value, or result. This document provides user instructions for the SAC codes that handle inventory tracking, release of contaminants to the environment, and transport of contaminants through the unsaturated zone, saturated zone, and the Columbia River
International Nuclear Information System (INIS)
Chepe P, M.; Xolocostli M, J. V.; Gomez T, A. M.; Del Valle G, E.
2016-09-01
Due to the current computing power, the deterministic codes for analyzing nuclear reactors that have been used for several years are becoming more relevant, since much more precise solution techniques can be used; the last century would have been very difficult, since memory and processor capacities were very limited or had high prices on the components. In this work we analyze the effect of the anisotropic dispersion of the effective dispersion section, compared to the isotropic dispersion. The anisotropy implementation was carried out in the AZTRAN transport code, which is part of the AZTLAN platform for nuclear reactors analysis (in development). The AZTRAN code solves the Boltzmann transport equation in one, two and three dimensions at steady state, using the multi-group technique for energy discretization, the RTN-0 nodal method in spatial discretization and for angular discretization the discrete ordinates without considering anisotropy originally. The effect of the anisotropy dispersion on the effective multiplication factor and the axial and radial power on a fuel assembly BWR type are analyzed. (Author)
International Nuclear Information System (INIS)
Fujimura, Toichiro
1996-01-01
A three-dimensional neutron transport code DFEM has been developed by the double finite element method to analyze reactor cores with complex geometry as large fast reactors. Solution algorithm is based on the double finite element method in which the space and angle finite elements are employed. A reactor core system can be divided into some triangular and/or quadrangular prism elements, and the spatial distribution of neutron flux in each element is approximated with linear basis functions. As for the angular variables, various basis functions are applied, and their characteristics were clarified by comparison. In order to enhance the accuracy, a general method is derived to remedy the truncation errors at reflective boundaries, which are inherent in the conventional FEM. An adaptive acceleration method and the source extrapolation method were applied to accelerate the convergence of the iterations. The code structure is outlined and explanations are given on how to prepare input data. A sample input list is shown for reference. The eigenvalue and flux distribution for real scale fast reactors and the NEA benchmark problems were presented and discussed in comparison with the results of other transport codes. (author)
Energy Technology Data Exchange (ETDEWEB)
Vergnaud, Th.; Nimal, J.C.; Chiron, M
2001-07-01
The TRIPOLI-3 code applies the Monte Carlo method to neutron, gamma-ray and coupled neutron and gamma-ray transport calculations in three-dimensional geometries, either in steady-state conditions or having a time dependence. It can be used to study problems where there is a high flux attenuation between the source zone and the result zone (studies of shielding configurations or source driven sub-critical systems, with fission being taken into account), as well as problems where there is a low flux attenuation (neutronic calculations -- in a fuel lattice cell, for example -- where fission is taken into account, usually with the calculation on the effective multiplication factor, fine structure studies, numerical experiments to investigate methods approximations, etc). TRIPOLI-3 has been operational since 1995 and is the version of the TRIPOLI code that follows on from TRIPOLI-2; it can be used on SUN, RISC600 and HP workstations and on PC using the Linux or Windows/NT operating systems. The code uses nuclear data libraries generated using the THEMIS/NJOY system. The current libraries were derived from ENDF/B6 and JEF2. There is also a response function library based on a number of evaluations, notably the dosimetry libraries IRDF/85, IRDF/90 and also evaluations from JEF2. The treatment of particle transport is the same in version 3.5 as in version 3.4 of the TRIPOLI code; but the version 3.5 is more convenient for preparing the input data and for reading the output. The french version of the user's manual exists. (authors)
DELTA : a computer code for determination of efficiency of particulate matter and aerosol transport
International Nuclear Information System (INIS)
Picini, P.; Caropreso, G.; Antonini, A.; Galuppi, G.; Sbrana, M.; Bardone, G.; Malvestuto, V.; Ricotta, A.
1996-04-01
In the Part I of this paper a mathematical model to calculate the sampling and transport efficiencies (both in laminar and turbulent condition) of any sampling and transport system decomposable in several cylindrical elemental component is presented. In the Part II an experimental facility built in Casaccia ENEA laboratory is described and the measures carried out to validate the model are reported
International Nuclear Information System (INIS)
Suteau, C.; Chiron, M.; Luneville, L.; Berger, L.; Huver, M.
2003-01-01
The M.E.R.C.U.R.E. calculation code (version 6.3) simulate the photons transport from 15 keV to 10 MeV in three dimensional geometries between volume sources and calculation points. It is based in the integration of attenuation punctual nuclei in straight line with accumulation factors. The accumulation factors take into account the following physical phenomena: photoelectric effect, coherent diffusion, incoherent diffusion, pairs production, radiation secondary sources coming from Bremsstrahlung and fluorescence. The code determines the accumulation factor of a succession of several screens with an innovative iterative method. M.E.R.C.U.R.E. -6.3 integers the punctual nuclei by a Monte Carlo method for which it automatically determines the importance distributions. The results of this code are compared with these ones of the Sn T.W.O.D.A.N.T. code in two one-dimensional configurations. One includes five screens composed of four different materials and the other one three screens. In the configuration with three screens, the second screen is of an infinitesimal thickness. (N.C.)
Reactor Dosimetry Applications Using RAPTOR-M3G:. a New Parallel 3-D Radiation Transport Code
Longoni, Gianluca; Anderson, Stanwood L.
2009-08-01
The numerical solution of the Linearized Boltzmann Equation (LBE) via the Discrete Ordinates method (SN) requires extensive computational resources for large 3-D neutron and gamma transport applications due to the concurrent discretization of the angular, spatial, and energy domains. This paper will discuss the development RAPTOR-M3G (RApid Parallel Transport Of Radiation - Multiple 3D Geometries), a new 3-D parallel radiation transport code, and its application to the calculation of ex-vessel neutron dosimetry responses in the cavity of a commercial 2-loop Pressurized Water Reactor (PWR). RAPTOR-M3G is based domain decomposition algorithms, where the spatial and angular domains are allocated and processed on multi-processor computer architectures. As compared to traditional single-processor applications, this approach reduces the computational load as well as the memory requirement per processor, yielding an efficient solution methodology for large 3-D problems. Measured neutron dosimetry responses in the reactor cavity air gap will be compared to the RAPTOR-M3G predictions. This paper is organized as follows: Section 1 discusses the RAPTOR-M3G methodology; Section 2 describes the 2-loop PWR model and the numerical results obtained. Section 3 addresses the parallel performance of the code, and Section 4 concludes this paper with final remarks and future work.
Energy Technology Data Exchange (ETDEWEB)
Watabe, Naoto; Suzuki, Hiroshi [Central Research Inst. of Electric Power Industry, Abiko, Chiba (Japan). Abiko Research Lab
1999-03-01
CRIEPI has been trying to adapt the probabilistic safety assessment (PSA) method to a safety assessment of radioactive materials (RAM) transport in Japan. As the new step of environmental risk assessment, the authors decided to adopt the `INTERTRAN 2 code` as the body for development works. Tow different routes at the hypothetical area which partially reflects the regional situation and traffic situation in Japan were selected in the trial calculation for the purpose of investigating the adaptability of `INTERTRAN 2`. Shuttle transport of hypothetical LLW containers was established in this case study. The collective dose in trial calculation was evaluated in both cases of `Incident Free mode` and `Accident mode`, and their subdivisions of collective dose were accorded to the definition of `INTERTRAN 2`. As for the difference of distance and population for two routes, it was demonstrated that collective dose data were properly derived from the route characteristics. From these results, it was confirmed that `INTERTRAN 2` code was adaptable to RAM transport in Japan, although there are some problems to be solved from the viewpoint of practical use and refining with a probabilistic approach. (M.N.)
Spallation integral experiment analysis by high energy nucleon-meson transport code
Energy Technology Data Exchange (ETDEWEB)
Takada, Hiroshi; Meigo, Shin-ichiro; Sasa, Toshinobu; Fukahori, Tokio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Yoshizawa, Nobuaki; Furihata, Shiori; Belyakov-Bodin, V.I.; Krupny, G.I.; Titarenko, Y.E.
1997-03-01
Reaction rate distributions were measured with various activation detectors on the cylindrical surface of the thick tungsten target of 20 cm in diameter and 60 cm in length bombarded with the 0.895 and 1.21 GeV protons. The experimental results were analyzed with the Monte Carlo simulation code systems of NMTC/JAERI-MCNP-4A, LAHET and HERMES. It is confirmed that those code systems can represent the reaction rate distributions with the C/E ratio of 0.6 to 1.4 at the positions up to 30 cm from beam incident surface. (author)
INTERTRAN 2 - A computer code for calculating the risk from transportation of radioactive materials
International Nuclear Information System (INIS)
Ericsson, A.M.; Jaernry, C.
1993-01-01
In this paper a description of IAEA Coordinated Research Program (CRP) dealing with the updating of the computer code INTERTRAN is given. The paper includes a summary of the work performed by several member states within the CRP as well as gives a description of the final product that will be presented to the IAEA. (J.P.N.)
BLAZE-DEM: A GPU based Polyhedral DEM particle transport code
CSIR Research Space (South Africa)
Govender, Nicolin
2013-05-01
Full Text Available This paper introduces the BLAZE-DEM code that is based on the Discrete Element Method (DEM) and specifically targeted for Graphical Processing Unit (GPU) platforms. BLAZE-DEM uses actual polyhedral particle representations as opposed to multi...
Energy Technology Data Exchange (ETDEWEB)
Iwamoto, Yosuke, E-mail: iwamoto.yosuke@jaea.go.jp; Ogawa, Tatsuhiko
2017-04-01
Because primary knock-on atoms (PKAs) create point defects and clusters in materials that are irradiated with neutrons, it is important to validate the calculations of recoil cross section spectra that are used to estimate radiation damage in materials. Here, the recoil cross section spectra of fission- and fusion-relevant materials were calculated using the Event Generator Mode (EGM) of the Particle and Heavy Ion Transport code System (PHITS) and also using the data processing code NJOY2012 with the nuclear data libraries TENDL2015, ENDF/BVII.1, and JEFF3.2. The heating number, which is the integral of the recoil cross section spectra, was also calculated using PHITS-EGM and compared with data extracted from the ACE files of TENDL2015, ENDF/BVII.1, and JENDL4.0. In general, only a small difference was found between the PKA spectra of PHITS + TENDL2015 and NJOY + TENDL2015. From analyzing the recoil cross section spectra extracted from the nuclear data libraries using NJOY2012, we found that the recoil cross section spectra were incorrect for {sup 72}Ge, {sup 75}As, {sup 89}Y, and {sup 109}Ag in the ENDF/B-VII.1 library, and for {sup 90}Zr and {sup 55}Mn in the JEFF3.2 library. From analyzing the heating number, we found that the data extracted from the ACE file of TENDL2015 for all nuclides were problematic in the neutron capture region because of incorrect data regarding the emitted gamma energy. However, PHITS + TENDL2015 can calculate PKA spectra and heating numbers correctly.
International Nuclear Information System (INIS)
Iwamoto, Y.; Ogawa, T.
2016-01-01
The modelling of the damage in materials irradiated by neutrons is needed for understanding the mechanism of radiation damage in fission and fusion reactor facilities. The molecular dynamics simulations of damage cascades with full atomic interactions require information about the energy distribution of the Primary Knock on Atoms (PKAs). The most common process to calculate PKA energy spectra under low-energy neutron irradiation is to use the nuclear data processing code NJOY2012. It calculates group-to-group recoil cross section matrices using nuclear data libraries in ENDF data format, which is energy and angular recoil distributions for many reactions. After the NJOY2012 process, SPKA6C is employed to produce PKA energy spectra combining recoil cross section matrices with an incident neutron energy spectrum. However, intercomparison with different processes and nuclear data libraries has not been studied yet. Especially, the higher energy (~5 MeV) of the incident neutrons, compared to fission, leads to many reaction channels, which produces a complex distribution of PKAs in energy and type. Recently, we have developed the event generator mode (EGM) in the Particle and Heavy Ion Transport code System PHITS for neutron incident reactions in the energy region below 20 MeV. The main feature of EGM is to produce PKA with keeping energy and momentum conservation in a reaction. It is used for event-by-event analysis in application fields such as soft error analysis in semiconductors, micro dosimetry in human body, and estimation of Displacement per Atoms (DPA) value in metals and so on. The purpose of this work is to specify differences of PKA spectra and heating number related with kerma between different calculation method using PHITS-EGM and NJOY2012+SPKA6C with different libraries TENDL-2015, ENDF/B-VII.1 and JENDL-4.0 for fusion relevant materials
Energy Technology Data Exchange (ETDEWEB)
Onishi, Yasuo; Yokuda, Satoru T.
2013-03-28
Pacific Northwest National Laboratory initiated the application of the time-varying, one-dimensional sediment-contaminant transport code, TODAM (Time-dependent, One-dimensional, Degradation, And Migration) to simulate the cesium migration and accumulation in the Ukedo River in Fukushima. This report describes the preliminary TODAM simulation results of the Ukedo River model from the location below the Ougaki Dam to the river mouth at the Pacific Ocean. The major findings of the 100-hour TODAM simulation of the preliminary Ukedo River modeling are summarized as follows:
Energy Technology Data Exchange (ETDEWEB)
Wiengarten, T.; Kleimann, J.; Fichtner, H. [Institut für Theoretische Physik IV, Ruhr-Universität Bochum (Germany); Kühl, P.; Kopp, A.; Heber, B. [Institut für Experimentelle und Angewandte Physik, Christian-Albrecht-Universität zu Kiel (Germany); Kissmann, R. [Institut für Astro- und Teilchenphysik, Universität Innsbruck (Austria)
2014-06-10
The transport of energetic particles such as cosmic rays is governed by the properties of the plasma being traversed. While these properties are rather poorly known for galactic and interstellar plasmas due to the lack of in situ measurements, the heliospheric plasma environment has been probed by spacecraft for decades and provides a unique opportunity for testing transport theories. Of particular interest for the three-dimensional (3D) heliospheric transport of energetic particles are structures such as corotating interaction regions, which, due to strongly enhanced magnetic field strengths, turbulence, and associated shocks, can act as diffusion barriers on the one hand, but also as accelerators of low energy CRs on the other hand as well. In a two-fold series of papers, we investigate these effects by modeling inner-heliospheric solar wind conditions with a numerical magnetohydrodynamic (MHD) setup (this paper), which will serve as an input to a transport code employing a stochastic differential equation approach (second paper). In this first paper, we present results from 3D MHD simulations with our code CRONOS: for validation purposes we use analytic boundary conditions and compare with similar work by Pizzo. For a more realistic modeling of solar wind conditions, boundary conditions derived from synoptic magnetograms via the Wang-Sheeley-Arge (WSA) model are utilized, where the potential field modeling is performed with a finite-difference approach in contrast to the traditional spherical harmonics expansion often utilized in the WSA model. Our results are validated by comparing with multi-spacecraft data for ecliptical (STEREO-A/B) and out-of-ecliptic (Ulysses) regions.
International Nuclear Information System (INIS)
Asai, Kiyoshi; Shinozawa, Naohisa; Ishikawa, Hirohiko; Chino, Masamichi; Hayashi, Takashi
1983-02-01
Three computer codes MATHEW, ADPIC of LLNL and GAMPUL of JAERI for prediction of wind field, concentration and external exposure rate of airborne radioactive materials are vectorized and the results are presented. Using the continuous equation of incompressible flow as a constraint, the MATHEW calculates the three dimensional wind field by a variational method. Using the particle-in -cell method, the ADPIC calculates the advection and diffusion of radioactive materials in three dimensional wind field and terrain, and gives the concentration of the materials in each cell of the domain. The GAMPUL calculates the external exposure rate assuming Gaussian plume type distribution of concentration. The vectorized code MATHEW attained 7.8 times speedup by a vector processor FACOM230-75 APU. The ADPIC and GAMPUL are estimated to attain 1.5 and 4 times speedup respectively on CRAY-1 type vector processor. (author)
International Nuclear Information System (INIS)
Franke, B.C.; Kensek, R.P.; Prinja, A.K.
2013-01-01
Stochastic-media simulations require numerous boundary crossings. We consider two Monte Carlo electron transport approaches and evaluate accuracy with numerous material boundaries. In the condensed-history method, approximations are made based on infinite-medium solutions for multiple scattering over some track length. Typically, further approximations are employed for material-boundary crossings where infinite-medium solutions become invalid. We have previously explored an alternative 'condensed transport' formulation, a Generalized Boltzmann-Fokker-Planck (GBFP) method, which requires no special boundary treatment but instead uses approximations to the electron-scattering cross sections. Some limited capabilities for analog transport and a GBFP method have been implemented in the Integrated Tiger Series (ITS) codes. Improvements have been made to the condensed history algorithm. The performance of the ITS condensed-history and condensed-transport algorithms are assessed for material-boundary crossings. These assessments are made both by introducing artificial material boundaries and by comparison to analog Monte Carlo simulations. (authors)
SPHERE: a spherical-geometry multimaterial electron/photon Monte Carlo transport code
International Nuclear Information System (INIS)
Halbleib, J.A. Sr.
1977-06-01
SPHERE provides experimenters and theorists with a method for the routine solution of coupled electron/photon transport through multimaterial configurations possessing spherical symmetry. Emphasis is placed upon operational simplicity without sacrificing the rigor of the model. SPHERE combines condensed-history electron Monte Carlo with conventional single-scattering photon Monte Carlo in order to describe the transport of all generations of particles from several MeV down to 1.0 and 10.0 keV for electrons and photons, respectively. The model is more accurate at the higher energies, with a less rigorous description of the particle cascade at energies where the shell structure of the transport media becomes important. Flexibility of construction permits the user to tailor the model to specific applications and to extend the capabilities of the model to more sophisticated applications through relatively simple update procedures. 8 figs., 3 tables
International Nuclear Information System (INIS)
Pölz, Stefan; Laubersheimer, Sven; Eberhardt, Jakob S; Harrendorf, Marco A; Keck, Thomas; Benzler, Andreas; Breustedt, Bastian
2013-01-01
The basic idea of Voxel2MCNP is to provide a framework supporting users in modeling radiation transport scenarios using voxel phantoms and other geometric models, generating corresponding input for the Monte Carlo code MCNPX, and evaluating simulation output. Applications at Karlsruhe Institute of Technology are primarily whole and partial body counter calibration and calculation of dose conversion coefficients. A new generic data model describing data related to radiation transport, including phantom and detector geometries and their properties, sources, tallies and materials, has been developed. It is modular and generally independent of the targeted Monte Carlo code. The data model has been implemented as an XML-based file format to facilitate data exchange, and integrated with Voxel2MCNP to provide a common interface for modeling, visualization, and evaluation of data. Also, extensions to allow compatibility with several file formats, such as ENSDF for nuclear structure properties and radioactive decay data, SimpleGeo for solid geometry modeling, ImageJ for voxel lattices, and MCNPX’s MCTAL for simulation results have been added. The framework is presented and discussed in this paper and example workflows for body counter calibration and calculation of dose conversion coefficients is given to illustrate its application. (paper)
International Nuclear Information System (INIS)
Peek, J.M.; Halbleib, J.A.
1983-04-01
The electron stopping and range data now used in the TIGER and TIGERP electron-transport codes are extracted and compared with other data for these processes. At the smallest collision energies treated by these codes, E approx. 1 keV, the stopping-power is estimated to be accurate for small-Z targets, to be about 25 percent too small for Z near 36 and to be a factor of three too small for Z > 79. These errors decrease with increasing E and the largest error for any target is roughly 20 percent for E = 10 keV. The closely related continuous-slowing-down range is estimated, at 1 keV, to be about 25 percent too small for small-Z targets and a factor of 2 too large for large-Z targets. The electron-transport problem of reflection from planer surfaces is re-investigated with improved stopping-power data. The effects of this change for the examples considered were about the size of the statistical uncertainties in the calculation, 1 to 2 percent
International Nuclear Information System (INIS)
Armand, Patrick
1995-01-01
The aim of this work consists in the Fluid Mechanics and aerosol Physics coupling. In the first part, the order of magnitude analysis of the particle dynamics is done. This particle is embedded in a non-uniform unsteady flow. Flow approximations around the inclusion are described. Corresponding aerodynamic drag formulae are expressed. Possible situations related to the problem data are extensively listed. In the second part, one studies the turbulent particles transport. Eulerian approach which is particularly well adapted to industrial codes is preferred in comparison with the Lagrangian methods. One chooses the two-fluid formalism in which career gas-particles slip is taken into account. Turbulence modelling gets through a k-epsilon modulated by the inclusions action on the flow. The model is implemented In a finite elements code. Finally, In the third part, one validates the modelling in laminar and turbulent cases. We compare simulations to various experiments (settling battery, inertial impaction in a bend, jets loaded with glass beads particles) which are taken in the literature or done by ourselves at the laboratory. The results are very close. It is a good point when it is thought of the particles transport model and associated software future use. (author) [fr
Energy Technology Data Exchange (ETDEWEB)
Ikushima, Takeshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1998-05-01
A computer code system CASKET (CASK thermal and structural analyses and Evaluation code system) for the thermal and structural analyses which are indispensable for radioactive material transport and/or storage cask designs has been developed. The CASKET is a simplified computer code system to perform parametric analyses on sensitivity evaluations in designing a cask and conducting its safety analysis. Main features of the CASKET are as follow: (1) it is capable to perform impact analysis of casks with shock absorbers, (2) it is capable to perform impact analysis of casks with fins. (3) puncture analysis of casks is capable, (4) rocking analysis of casks during seismic load is capable, (5) material property data library are provided for impact analysis of casks, (6) material property data library are provided for thermal analysis of casks, (7) fin energy absorption data library are provided for impact analysis of casks with fins are and (8) not only main frame computers (OS MSP) but also work stations (OS UNIX) and personal computers (OS Windows 3.1) are available. In the paper, brief illustrations of calculation methods are presented. Some calculation results are compared with experimental ones to confirm the computer programs are useful for thermal and structural analyses. (author)
International Nuclear Information System (INIS)
Yoon, Churl; Tak, Nam Il; Lim, Hong Sik
2010-01-01
One of the unique features of a Very High Temperature Gas Cooled Reactor (VHTR) is Vented Low Pressure Containment (VLPC) containing two separate vent paths where both have two gravity operated relief valves in a series. Because VLPC strategy allows the release of a relatively small amount of radioactive fission products(FP) into the environment during the blowdown phase, behavior analyses of the fission products circulating in the primary coolant loop and in the containment are major consideration factors for safety evaluation. For thermal-fluid analysis of a Very High Temperature Gas Cooled Reactor (VHTR), the GAMMA(GAs Multicomponent Mixture Analysis)+ code is under development. The MAEROS model is the multicomponent aerosol module of the CONTAIN code, and has been widely used for aerosol behavior analysis. For the first work of FP module development, the MAEROS model had been implemented as an independent module and examined against some analytic solutions and experimental data by Yoo et al. In this study, an aerosol transport model and a turbulent resuspension model were additionally implemented in the FP module of the GAMMA+ code and verified for FP analysis of a VHTR
Application of Inverse Gamma Transport to Material Thickness Identification with SGRD Code
Directory of Open Access Journals (Sweden)
Humbert Philippe
2017-01-01
Full Text Available SGRD (Spectroscopy, Gamma rays, Rapid, Deterministic code is used to infer the dimensions of a one dimensional model of a shielded gamma ray source. The method is based on the simulation of the uncollided leakage current of discrete gamma lines that are produced by nuclear decay. Experimentally, the unscattered gamma lines leakage current is obtained by processing high precision gamma spectroscopy measurements. The material thicknesses are computed with SGRD using a fast ray-tracing algorithm embedded in a non-linear multidimensional iterative optimization procedure that minimizes the error metric between calculated and measured signatures. For verification, numerical results on a test problem are presented.
MESTRN: A Deterministic Meson-Muon Transport Code for Space Radiation
Blattnig, Steve R.; Norbury, John W.; Norman, Ryan B.; Wilson, John W.; Singleterry, Robert C., Jr.; Tripathi, Ram K.
2004-01-01
A safe and efficient exploration of space requires an understanding of space radiations, so that human life and sensitive equipment can be protected. On the way to these sensitive sites, the radiation fields are modified in both quality and quantity. Many of these modifications are thought to be due to the production of pions and muons in the interactions between the radiation and intervening matter. A method used to predict the effects of the presence of these particles on the transport of radiation through materials is developed. This method was then used to develop software, which was used to calculate the fluxes of pions and muons after the transport of a cosmic ray spectrum through aluminum and water. Software descriptions are given in the appendices.
Projection of the Cost-Effectiveness of PIMs for Particle Transport Codes
International Nuclear Information System (INIS)
CHRISTOPHER, THOMAS WOODS
2003-01-01
PIM (Processor in Memory) architectures are being proposed for future supercomputers, because they reduce the problems that SMP MMPs have with latency. However, they do not meet the SMP MPP balance factors. Being relatively processor rich and memory starved, it is unclear whether an ASCI application could run on them, either as-is or with recoding. The KBA (Koch-Baker-Alcouffe) algorithm (Koch, 1992) for particle transport (radiation transport) is shown not to fit on PIMs as written. When redesigned with a 3-D allocation of cells to PIMs, the resulting algorithm is projected to execute an order of magnitude faster and more cost-effectively than the KBA algorithm, albeit with high initial hardware costs
Radiation transport code with adaptive Mesh Refinement: acceleration techniques and applications
International Nuclear Information System (INIS)
Velarde, Pedro; Garcia-Fernaandez, Carlos; Portillo, David; Barbas, Alfonso
2011-01-01
We present a study of acceleration techniques for solving Sn radiation transport equations with Adaptive Mesh Refinement (AMR). Both DSA and TSA are considered, taking into account the influence of the interaction between different levels of the mesh structure and the order of approximation in angle. A Hybrid method is proposed in order to obtain better convergence rate and lower computer times. Some examples are presented relevant to ICF and X ray secondary sources. (author)
MNM1D: A Numerical Code for Colloid Transport in Porous Media: Implementation and Validation
Tiziana Tosco; Rajandrea Sethi
2009-01-01
Problem statement: Understanding the mechanisms that control the transport and fate of colloidal particles in subsurface environments is a crucial issue faced by several researchers in the last years. In many cases, natural colloids have been shown to play a major role in the spreading of strongly sorbing contaminants, while manufactured micro-and nanoparticles, which are nowadays widely spread in the subsurface, can be toxic themselves. On the other hand, in recent years studies have been ad...
International Nuclear Information System (INIS)
White, Morgan C.
2000-01-01
The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V and V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second, the ability to
Energy Technology Data Exchange (ETDEWEB)
White, Morgan C. [Univ. of Florida, Gainesville, FL (United States)
2000-07-01
The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V&V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second
A new philosophy for calibrating oil well logging tools based on neutron transport codes
International Nuclear Information System (INIS)
Butler, J.; Clayton, C.G.
1984-01-01
The current practice of calibrating neutron borehole logging probes is limited by an inability to match calibration conditions to those which pertain in an operational situation. In addition, test boreholes are expensive to construct and, when natural materials are used, rely on an exact correspondence in composition and in structure between the materials of the test facility and representative samples which may not be valid. Now that neutron tansport codes have been developed to a point at which they are able to cope with realistic, complex situations an alternative approach to calibration can be considered. The basis of this philosophy is the construction of a limited number of calibration facilities which are composed of artificial rocks of controlled but variable porosity and accurately known nuclear characteristics
Yamoto, S.; Bonnin, X.; Homma, Y.; Inoue, H.; Hoshino, K.; Hatayama, A.; Pitts, R. A.
2017-11-01
In order to obtain a better understanding of tungsten (W) transport processes, we are developing the Monte-Carlo W transport code IMPGYRO. The code has the following characteristics which are important for calculating W transport: (1) the exact Larmor motion of W ions is computed so that the effects of drifts are automatically taken into account; (2) Coulomb collisions between W impurities and background plasma ions are modelled using the Binary Collision Model which provides more precise kinetic calculations of the friction and thermal forces. By using the IMPGYRO code, the W production/transport in the ITER geometry has been calculated under two different divertor operation modes (Case A: partially detached state and Case B: high recycling state) obtained from the SOLPS-ITER code suite calculation without the effect of drifts. The results of the W-density in the upstream SOL (scrape-off layer) strongly depend on the divertor operation mode. From the comparison of the W impurity transport between Case A and Case B, obtaining a partially detached state is shown to be effective to reduce W-impurities in the upstream SOL. The limitations of the employed model and the validity of the above results are discussed and future problems are summarized for further applications of IMPGYRO code to ITER plasmas.
A Monte Carlo transport code study of the space radiation environment using FLUKA and ROOT
Wilson, T; Carminati, F; Brun, R; Ferrari, A; Sala, P; Empl, A; MacGibbon, J
2001-01-01
We report on the progress of a current study aimed at developing a state-of-the-art Monte-Carlo computer simulation of the space radiation environment using advanced computer software techniques recently available at CERN, the European Laboratory for Particle Physics in Geneva, Switzerland. By taking the next-generation computer software appearing at CERN and adapting it to known problems in the implementation of space exploration strategies, this research is identifying changes necessary to bring these two advanced technologies together. The radiation transport tool being developed is tailored to the problem of taking measured space radiation fluxes impinging on the geometry of any particular spacecraft or planetary habitat and simulating the evolution of that flux through an accurate model of the spacecraft material. The simulation uses the latest known results in low-energy and high-energy physics. The output is a prediction of the detailed nature of the radiation environment experienced in space as well a...
Computer codes for automatic tuning of the beam transport at the UNILAC
International Nuclear Information System (INIS)
Dahl, L.; Ehrich, A.
1984-01-01
For application in routine operation fully automatic computer controlled algorithms are developed for tuning of beam transport elements at the Unilac. Computations, based on emittance measurements, simulate the beam behaviour and evaluate quadrupole settings, in order to produce defined beam properties at specified positions along the accelerator. The interactive program is controlled using a graphic display on which the beam emittances and envelopes are plotted. To align the beam onto the ion-optical axis of the accelerator two automatic computer controlled procedures have been developed. The misalignment of the beam is determined by variation of quadrupole or steering magnet settings with simultaneous measurement of the beam distribution on profile grids. According to the result a pair of steering magnet settings are adjusted to bend the beam on the axis. The effects of computer controlled tuning on beam quality and operation are reported
Di Gioia, F; Aprile, A; Sabella, E; Santamaria, P; Pardossi, A; Miceli, A; De Bellis, L; Nutricati, E
2017-09-01
Boron (B) is essential for plant growth, however its excess in soil and/or in irrigation water can severely compromise plant growth and yield. The goal of this work was to determine whether grafting onto 'Arnold', a commercial interspecific hybrid (Solanum lycopersicum × S. habrochaites) rootstock, which in a previous study was found to be tolerant to salt stress, could improve tomato (S. lycopersicum L. 'Ikram') tolerance to excess B, and whether this effect is associated with an exclusion mechanism. Non-grafted, self-grafted and grafted plants were hydroponically grown in a greenhouse with B concentration in the nutrient solution of 0.27 (control), 5, 10 and 15 mg·l -1 . A transcription analysis was carried out on SlNIP5 and SlBOR1 genes, which encode putative B transporters. Grafting 'Ikram' onto 'Arnold' rootstock reduced B concentration in leaf tissue of plants exposed to B concentrations of 10-15 mg·l -1 . At high B levels, SlNIP5 was down-regulated in all grafting combinations, while SlBOR1 was down-regulated only in the roots of plants grafted onto 'Arnold'. We conclude that grafting the susceptible tomato cultivar 'Ikram' onto the commercial rootstock 'Arnold' improved tolerance to excess B by reducing expression of genes encoding for B transporters at the root level, thus partially reducing the root uptake of B and its accumulation in the shoot. © 2017 German Botanical Society and The Royal Botanical Society of the Netherlands.
Energy Technology Data Exchange (ETDEWEB)
Sugino, Kazuteru [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center
1998-07-01
As a tool to perform a fast reactor core calculations with high accuracy, NSHEX the nodal transport calculation code for three-dimensional hexagonal-Z geometry is under development. To improve the practical applicability of NSHEX, for instance, in its application to safety analysis and commercial reactor core design studies, we investigated the basic theory used in it, improved the program performance, and evaluated its applicability to the analysis of commercial reactor cores. The current studies show the following: (1) An improvement in the treatment of radial leakage in the radial nodal coupling equation bettered calculational convergence for safety analysis calculation, so the applicability of NSHEX to safety analysis was improved. (2) As a result of comparison of results from NSHEX and the standard core calculation code, it was confirmed that there was consistency between them. (3) According to the evaluation of the effect due to the difference of calculational condition, it was found that the calculation under appropriate nodal expansion orders and Sn orders correspond to the one under most detailed condition. However further investigation is required to reduce the uncertainty in calculational results due to the treatment of high order flux moments. (4) A whole core version of NSHEX enabling calculation for any FBR core geometry has been developed, this improved general applicability of NSHEX. (5) An investigation of the applicability of the rebalance method to acceleration clarified that this improved calculational convergence and it was effective. (J.P.N.)
Energy Technology Data Exchange (ETDEWEB)
Williams, M.L.; Yuecel, A.; Nadkarny, S.
1988-05-01
The HEATING6 heat conduction code is modified to (a) read the multigroup particle fluxes from a two-dimensional DOT-IV neutron- photon transport calculation, (b) interpolate the fluxes from the DOT-IV variable (optional) mesh to the HEATING6 control volume mesh, and (c) fold the interpolated fluxes with kerma factors to obtain a nuclear heating source for the heat conduction equation. The modified HEATING6 is placed as a module in the ORNL discrete ordinates system (DOS), and has been renamed DOS-HEATING6. DOS-HEATING6 provides the capability for determining temperature distributions due to nuclear heating in complex, multi-dimensional systems. All of the original capabilities of HEATING6 are retained for the nuclear heating calculation; e.g., generalized boundary conditions (convective, radiative, finned, fixed temperature or heat flux), temperature and space dependent thermal properties, steady-state or transient analysis, general geometry description, etc. The numerical techniques used in the code are reviewed and the user input instructions and JCL to perform DOS-HEATING6 calculations are presented. Finally a sample problem involving coupled DOT-IV and DOS-HEATING6 calculations of a complex space-reactor configurations described, and the input and output of the calculations are listed. 10 refs., 11 figs., 6 tabs.
Shi, Xue-Ming; Peng, Xian-Jue
2016-09-01
Fusion science and technology has made progress in the last decades. However, commercialization of fusion reactors still faces challenges relating to higher fusion energy gain, irradiation-resistant material, and tritium self-sufficiency. Fusion Fission Hybrid Reactors (FFHR) can be introduced to accelerate the early application of fusion energy. Traditionally, FFHRs have been classified as either breeders or transmuters. Both need partition of plutonium from spent fuel, which will pose nuclear proliferation risks. A conceptual design of a Fusion Fission Hybrid Reactor for Energy (FFHR-E), which can make full use of natural uranium with lower nuclear proliferation risk, is presented. The fusion core parameters are similar to those of the International Thermonuclear Experimental Reactor. An alloy of natural uranium and zirconium is adopted in the fission blanket, which is cooled by light water. In order to model blanket burnup problems, a linkage code MCORGS, which couples MCNP4B and ORIGEN-S, is developed and validated through several typical benchmarks. The average blanket energy Multiplication and Tritium Breeding Ratio can be maintained at 10 and 1.15 respectively over tens of years of continuous irradiation. If simple reprocessing without separation of plutonium from uranium is adopted every few years, FFHR-E can achieve better neutronic performance. MCORGS has also been used to analyze the ultra-deep burnup model of Laser Inertial Confinement Fusion Fission Energy (LIFE) from LLNL, and a new blanket design that uses Pb instead of Be as the neutron multiplier is proposed. In addition, MCORGS has been used to simulate the fluid transmuter model of the In-Zinerater from Sandia. A brief comparison of LIFE, In-Zinerater, and FFHR-E will be given.
Energy Technology Data Exchange (ETDEWEB)
Franke, Brian Claude; Kensek, Ronald Patrick; Laub, Thomas William
2005-09-01
ITS is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of linear time-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. Our goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 5.0, the latest version of ITS, contains (1) improvements to the ITS 3.0 continuous-energy codes, (2) multigroup codes with adjoint transport capabilities, (3) parallel implementations of all ITS codes, (4) a general purpose geometry engine for linking with CAD or other geometry formats, and (5) the Cholla facet geometry library. Moreover, the general user friendliness of the software has been enhanced through increased internal error checking and improved code portability.
International Nuclear Information System (INIS)
Seed, T.J.; Miller, W.F. Jr.; Brinkley, F.W. Jr.
1977-03-01
TRIDENT solves the two-dimensional-multigroup-transport equations in rectangular (x-y) and cylindrical (r-z) geometries using a regular triangular mesh. Regular and adjoint, inhomogeneous and homogeneous (k/sub eff/ and eigenvalue searches) problems subject to vacuum, reflective, white, or source boundary conditions are solved. General anisotropic scattering is allowed and anisotropic-distributed sources are permitted. The discrete-ordinates approximation is used for the neutron directional variables. An option is included to append a fictitious source to the discrete-ordinates equations that is defined such that spherical-harmonics solutions (in x-y geometry) or spherical-harmonics-like solutions (in r-z geometry) are obtained. A spatial-finite-element method is used in which the angular flux is expressed as a linear polynomial in each triangle that is discontinous at triangle boundaries. Unusual Features of the program: Provision is made for creation of standard interface output files for S/sub N/ constants, angle-integrated (scalar) fluxes, and angular fluxes. Standard interface input files for S/sub N/ constants, inhomogeneous sources, cross sections, and the scalar flux may be read. Flexible edit options as well as a dump and restart capability are provided
Directory of Open Access Journals (Sweden)
Emerson A Castilho-Martins
Full Text Available Leishmania (L. amazonensis uses arginine to synthesize polyamines to support its growth and survival. Here we describe the presence of two gene copies, arranged in tandem, that code for the arginine transporter. Both copies show similar Open Reading Frames (ORFs, which are 93% similar to the L. (L. donovani AAP3 gene, but their 5' and 3' UTR's have distinct regions. According to quantitative RT-PCR, the 5.1 AAP3 mRNA amount was increased more than 3 times that of the 4.7 AAP3 mRNA along the promastigote growth curve. Nutrient deprivation for 4 hours and then supplemented or not with arginine (400 µM resulted in similar 4.7 AAP3 mRNA copy-numbers compared to the starved and control parasites. Conversely, the 5.1 AAP3 mRNA copy-numbers increased in the starved parasites but not in ones supplemented with arginine (p<0.05. These results correlate with increases in amino acid uptake. Both Meta1 and arginase mRNAs remained constant with or without supplementation. The same starvation experiment was performed using a L. (L. amazonensis null knockout for arginase (arg(- and two other mutants containing the arginase ORF with (arg(-/ARG or without the glycosomal addressing signal (arg(-/argΔSKL. The arg(- and the arg(-/argΔSKL mutants did not show the same behavior as the wild-type (WT parasite or the arg(-/ARG mutant. This can be an indicative that the internal pool of arginine is also important for controlling transporter expression and function. By inhibiting mRNA transcription or/and mRNA maturation, we showed that the 5.1 AAP3 mRNA did not decay after 180 min, but the 4.7 AAP3 mRNA presented a half-life decay of 32.6 +/- 5.0 min. In conclusion, parasites can regulate amino acid uptake by increasing the amount of transporter-coding mRNA, possibly by regulating the mRNA half-life in an environment where the amino acid is not present or is in low amounts.
International Nuclear Information System (INIS)
Sauter, O.; Harvey, R.W.; Hinton, F.L.
1993-10-01
A new 3-D Fokker-Planck code, CQL, which solves the Fokker-Planck equations with two velocity coordinates and one spatial coordinate parallel to the magnetic field lines B/B, has been developed. This code enables us to model the parallel transport for low, intermediate and high collisional regime. The physical model, the possible relevant applications of the code as well as a first application, the computation of the neoclassical resistivity for various collisionalities and aspect ratios in tokamak geometry are presented. (author) 3 figs., 3 refs
Energy Technology Data Exchange (ETDEWEB)
Both, J.P.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B
2003-07-01
This manual relates to Version 4.3 TRIPOLI-4 code. TRIPOLI-4 is a computer code simulating the transport of neutrons, photons, electrons and positrons. It can be used for radiation shielding calculations (long-distance propagation with flux attenuation in non-multiplying media) and neutronic calculations (fissile medium, criticality or sub-criticality basis). This makes it possible to calculate k{sub eff} (for criticality), flux, currents, reaction rates and multi-group cross-sections. TRIPOLI-4 is a three-dimensional code that uses the Monte-Carlo method. It allows for point-wise description in terms of energy of cross-sections and multi-group homogenized cross-sections and features two modes of geometrical representation: surface and combinatorial. The code uses cross-section libraries in ENDF/B format (such as JEF2-2, ENDF/B-VI and JENDL) for point-wise description cross-sections in APOTRIM format (from the APOLLO2 code) or a format specific to TRIPOLI-4 for multi-group description. (authors)
Energy Technology Data Exchange (ETDEWEB)
Cupini, E. [ENEA, Centro Ricerche `Ezio Clementel`, Bologna (Italy). Dipt. Innovazione; Borgia, M.G. [ENEA, Centro Ricerche `Ezio Clementel`, Bologna (Italy). Dipt. Energia; Premuda, M. [Consiglio Nazionale delle Ricerche, Bologna (Italy). Ist. FISBAT
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.
Energy Technology Data Exchange (ETDEWEB)
1974-01-01
The code outlines general requirements for pollution prevention and provides guidelines for corrosion protection of mild steel tanks, pipe and fitting assemblies, and for storage tank installations. The transportation and delivery of petroleum fuels are discussed, and operating procedures are suggested.
Energy Technology Data Exchange (ETDEWEB)
Lao, Lang L. [General Atomics; St John, Holger [General Atomics; Staebler, Gary M. [General Atomics; Snyder, Phil B. [General Atomics
2010-08-20
This report describes the work done under U.S. Department of Energy grant number DE-FG02-07ER54935 for the period ending July 31, 2010. The goal of this project was to provide predictive transport analysis to the PTRANSP code. Our contribution to this effort consisted of three parts: (a) a predictive solver suitable for use with highly non-linear transport models and installation of the turbulent confinement models GLF23 and TGLF, (b) an interface of this solver with the PTRANSP code, and (c) initial development of an EPED1 edge pedestal model interface with PTRANSP. PTRANSP has been installed locally on this cluster by importing a complete PTRANSP build environment that always contains the proper version of the libraries and other object files that PTRANSP requires. The GCNMP package and its interface code have been added to the SVN repository at PPPL.
International Nuclear Information System (INIS)
Simmons, C.S.; Cole, C.R.
1985-08-01
This document was written for the National Low-Level Waste Management Program to provide guidance for managers and site operators who need to select ground-water transport codes for assessing shallow-land burial site performance. The guidance given in this report also serves the needs of applications-oriented users who work under the direction of a manager or site operator. The guidelines are published in two volumes designed to support the needs of users having different technical backgrounds. An executive summary, published separately, gives managers and site operators an overview of the main guideline report. Volume 1, titled ''Guideline Approach,'' consists of Chapters 1 through 5 and a glossary. Chapters 2 through 5 provide the more detailed discussions about the code selection approach. This volume, Volume 2, consists of four appendices reporting on the technical evaluation test cases designed to help verify the accuracy of ground-water transport codes. 20 refs
International Nuclear Information System (INIS)
Weber, C.F.; Beahm, E.C.; Kress, T.S.; Daish, S.R.; Shockley, W.E.
1989-01-01
The ultimate aim of a description of iodine behavior in severe LWR accidents is a time-dependent accounting of iodine species released into containment and to the environment. Factors involved in the behavior of iodine can be conveniently divided into four general categories: (1) initial release into containment, (2) interaction of iodine species in containment not directly involving water pools, (3) interaction of iodine species in, or with, water pools, and (4) interaction with special systems such as ice condensers or gas treatment systems. To fill the large gaps in knowledge and to provide a means for assaying the iodine source term, this program has proceeded along two paths: (1) Experimental studies of the chemical behavior of iodine under containment conditions. (2) Development of TRENDS (Transport and Retention of Nuclides in Dominant Sequences), a computer code for modeling the behavior of iodine in containment and its release from containment. The main body of this report consists of a description of TRENDS. These two parts to the program are complementary in that models within TRENDS use data that were produced in the experimental program; therefore, these models are supported by experimental evidence that was obtained under conditions expected in severe accidents. 7 refs., 1 fig., 2 tabs
International Nuclear Information System (INIS)
Kirk, B.L.
1985-12-01
The ITS (Integrated Tiger Series) Monte Carlo code package developed at Sandia National Laboratories and distributed as CCC-467/ITS by the Radiation Shielding Information Center (RSIC) at Oak Ridge National Laboratory (ORNL) consists of eight codes - the standard codes, TIGER, CYLTRAN, ACCEPT; the P-codes, TIGERP, CYLTRANP, ACCEPTP; and the M-codes ACCEPTM, CYLTRANM. The codes have been adapted to run on the IBM 3081, VAX 11/780, CDC-7600, and Cray 1 with the use of the update emulator UPEML. This manual should serve as a guide to a user running the codes on IBM computers having 370 architecture. The cases listed were tested on the IBM 3033, under the MVS operating system using the VS Fortran Level 1.3.1 compiler
International Nuclear Information System (INIS)
Shin, Chang Hwan; Seo, Kyong Won; Chun, Tae Hyun; Kim, Kang Seog
2005-03-01
Code coupling activities have so far focused on coupling the neutronics modules with the CFD module. An interface module for the CFD-ACE/DeCART coupling was established as an alternative to the original STAR-CD/DeCART interface. The interface module for DeCART/CFD-ACE was validated by single-pin model. The optimized CFD mesh was decided through the calculation of multi-pin model. It was important to consider turbulent mixing of subchannels for calculation of fuel temperature. For the parallel calculation, the optimized decompose process was necessary to reduce the calculation costs and setting of the iteration and convergence criterion for each code was important, too
Energy Technology Data Exchange (ETDEWEB)
Ahnert, C.; Aragones, J. M.
1981-07-01
This Is a users manual of the neutron transport code TWOTRAN-TRACA, which is a version of the original TWOTRAN-GG from the Los Alamos Laboratory, with some modifications made at JEN. A detailed input data description is given as well as the new modifications developed at JEN. (Author) 8 refs.
International Nuclear Information System (INIS)
Koyama, Kinji; Taji, Yukichi; Miyasaka, Shun-ichi; Minami, Kazuyoshi.
1977-07-01
The modular code system RADHEAT is for producing coupled multigroup neutron and gamma-ray cross section sets, analyzing the neutron and gamma-ray transport, and calculating the energy deposition and atomic displacements due to these radiations in a nuclear reactor or shield. The basic neutron cross sections and secondary gamma-ray production data are taken from ENDF/B and POPOP4 libraries respectively. The system (1) generates multigroup neutron cross sections, energy deposition coefficients and atomic displacement factors due to neutron reactions, (2) generates multigroup gamma-ray cross sections and energy transfer coefficients, (3) generates secondary gamma-ray production cross sections, (4) combines these cross sections into the coupled set, (5) outputs and updates the multigroup cross section libraries in convenient formats for other transport codes, (6) analyzes the neutron and gamma-ray transport and calculates the energy deposition and the number density of atomic displacements in a medium, (7) collapses the cross sections to a broad-group structure, by option, using the weighting functions obtained by one-dimensional transport calculation, and (8) plots, by option, multigroup cross sections, and neutron and gamma-ray distributions. Definitions of the input data required in various options of the code system are also given. (auth.)
International Nuclear Information System (INIS)
Thiagu Supramaniam
2007-01-01
The aim of this research was to propose a new neutron collimator design for thermal neutron radiography facility using tangential beam port of PUSPATI TRIGA Mark II reactor, Malaysia Institute of Nuclear Technology Research (MINT). Best geometry and materials for neutron collimator were chosen in order to obtain a uniform beam with maximum thermal neutron flux, high L/ D ratio, high neutron to gamma ratio and low beam divergence with high resolution. Monte Carlo N-particle Transport Code version 5 (MCNP 5) was used to optimize six neutron collimator components such as beam port medium, neutron scatterer, neutron moderator, gamma filter, aperture and collimator wall. The reactor and tangential beam port setup in MCNP5 was plotted according to its actual sizes. A homogeneous reactor core was assumed and population control method of variance reduction technique was applied by using cell importance. The comparison between experimental results and simulated results of the thermal neutron flux measurement of the bare tangential beam port, shows that both graph obtained had similar pattern. This directly suggests the reliability of MCNP5 in order to obtained optimal neutron collimator parameters. The simulated results of the optimal neutron medium, shows that vacuum was the best medium to transport neutrons followed by helium gas and air. The optimized aperture component was boral with 3 cm thickness. The optimal aperture center hole diameter was 2 cm which produces 88 L/ D ratio. Simulation also shows that graphite neutron scatterer improves thermal neutron flux while reducing fast neutron flux. Neutron moderator was used to moderate fast and epithermal neutrons in the beam port. Paraffin wax with 90 cm thick was bound to be the best neutron moderator material which produces the highest thermal neutron flux at the image plane. Cylindrical shape high density polyethylene neutron collimator produces the highest thermal neutron flux at the image plane rather than divergent
Iwamoto, Yosuke
2018-03-01
In this study, the Monte Carlo displacement damage calculation method in the Particle and Heavy-Ion Transport code System (PHITS) was improved to calculate displacements per atom (DPA) values due to irradiation by electrons (or positrons) and gamma rays. For the damage due to electrons and gamma rays, PHITS simulates electromagnetic cascades using the Electron Gamma Shower version 5 (EGS5) algorithm and calculates DPA values using the recoil energies and the McKinley-Feshbach cross section. A comparison of DPA values calculated by PHITS and the Monte Carlo assisted Classical Method (MCCM) reveals that they were in good agreement for gamma-ray irradiations of silicon and iron at energies that were less than 10 MeV. Above 10 MeV, PHITS can calculate DPA values not only for electrons but also for charged particles produced by photonuclear reactions. In DPA depth distributions under electron and gamma-ray irradiations, build-up effects can be observed near the target's surface. For irradiation of 90-cm-thick carbon by protons with energies of more than 30 GeV, the ratio of the secondary electron DPA values to the total DPA values is more than 10% and increases with an increase in incident energy. In summary, PHITS can calculate DPA values for all particles and materials over a wide energy range between 1 keV and 1 TeV for electrons, gamma rays, and charged particles and between 10-5 eV and 1 TeV for neutrons.
KIM, Jong Woon; LEE, Young-Ouk
2017-09-01
As computing power gets better and better, computer codes that use a deterministic method seem to be less useful than those using the Monte Carlo method. In addition, users do not like to think about space, angles, and energy discretization for deterministic codes. However, a deterministic method is still powerful in that we can obtain a solution of the flux throughout the problem, particularly as when particles can barely penetrate, such as in a deep penetration problem with small detection volumes. Recently, a new state-of-the-art discrete-ordinates code, ATTILA, was developed and has been widely used in several applications. ATTILA provides the capabilities to solve geometrically complex 3-D transport problems by using an unstructured tetrahedral mesh. Since 2009, we have been developing our own code by benchmarking ATTILA. AETIUS is a discrete ordinates code that uses an unstructured tetrahedral mesh such as ATTILA. For pre- and post- processing, Gmsh is used to generate an unstructured tetrahedral mesh by importing a CAD file (*.step) and visualizing the calculation results of AETIUS. Using a CAD tool, the geometry can be modeled very easily. In this paper, we describe a brief overview of AETIUS and provide numerical results from both AETIUS and a Monte Carlo code, MCNP5, in a deep penetration problem with small detection volumes. The results demonstrate the effectiveness and efficiency of AETIUS for such calculations.
International Nuclear Information System (INIS)
Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa; Nakagawa, Masayuki
2005-06-01
In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two vectorized Monte Carlo codes MVP and GMVP have been developed at JAERI. MVP is based on the continuous energy model and GMVP is on the multigroup model. Compared with conventional scalar codes, these codes achieve higher computation speed by a factor of 10 or more on vector super-computers. Both codes have sufficient functions for production use by adopting accurate physics model, geometry description capability and variance reduction techniques. The first version of the codes was released in 1994. They have been extensively improved and new functions have been implemented. The major improvements and new functions are (1) capability to treat the scattering model expressed with File 6 of the ENDF-6 format, (2) time-dependent tallies, (3) reaction rate calculation with the pointwise response function, (4) flexible source specification, (5) continuous-energy calculation at arbitrary temperatures, (6) estimation of real variances in eigenvalue problems, (7) point detector and surface crossing estimators, (8) statistical geometry model, (9) function of reactor noise analysis (simulation of the Feynman-α experiment), (10) arbitrary shaped lattice boundary, (11) periodic boundary condition, (12) parallelization with standard libraries (MPI, PVM), (13) supporting many platforms, etc. This report describes the physical model, geometry description method used in the codes, new functions and how to use them. (author)
International Nuclear Information System (INIS)
Hykes, J. M.; Azmy, Y. Y.; Schunert, S.; King, S. H.; Klingensmith, J. J.
2009-01-01
The goal of this work is to determine the viability of modeling an important x-ray procedure, the computed tomography (CT) scan of a pregnant woman and her conceptus using a deterministic radiation transport program. A prior experimental study provides the deposited dose as measured in an anthropomorphic phantom, with detectors positioned in the estimated uterine location. In this paper, we first verify the discrete ordinates code TORT3.2 and a suitably constructed multigroup cross section library against the Monte Carlo code MCNP5. Using MCNP, we demonstrate that accounting for the transport of secondary electrons is unnecessary in tissue-equivalent material. After demonstrating proper verification, we proceed to validate the MCNP and TORT simulations against data measured for the CTDI FDA phantom. In the model, the computed edge-to-center dose ratio is within experimental uncertainty, while the computed exposures are less than 35% from the measured values. (authors)
International Nuclear Information System (INIS)
Hardin, Emmanuelle
1999-01-01
The study of cation interactions with solid materials is useful in order to define the chemistry interaction component of the MIMICC project (Multidimensional Instrumented Module for Investigations on chemistry-transport Coupled Codes). This project will validate the chemistry-transport coupled codes. Database have to be supplied on the cesium or ytterbium interactions with solid materials in suspension. The solid materials are: a strong cation exchange resin, a natural sand which presents small impurities, and a zirconium phosphate. The cation exchange resin is useful to check that the surface complexation theory can be applied on a pure cation exchanger. The sand is a natural material, and its isotherms will be interpreted using pure oxide-cation system data, such as pure silica-cation data. Then the study on the zirconium phosphate salt is interesting because of the increasing complexity in the processes (dissolution, sorption and co-precipitation). These data will enable to approach natural systems, constituted by several complex solids which can interfere on each other. These data can also be used for chemistry-transport coupled codes. Potentiometric titration, sorption isotherms, sorption kinetics, cation surface saturation curves are made, in order to obtain the different parameters relevant to the cation sorption at the solid surface, for each solid-electrolyte-cation system. The influence of different parameters such as ionic strength, pH, and electrolyte is estimated. All the experimental curves are fitted with FITEQL code based on the surface complexation theory using the constant capacitance model, in order to give a mechanistic interpretation of the ion retention phenomenon at the solid surface. The speciation curves of all systems are plotted, using the FITEQL code too. Systems with an increasing complexity are studied: dissolution, sorption and coprecipitation coexist in the cation-salt systems. Then the data obtained on each single solid, considered
Ghorai, S. K.
1983-01-01
The purpose of this project was to use a one-dimensional discrete coordinates transport code called ANISN in order to determine the energy-angle-spatial distribution of neutrons in a 6-feet cube rock box which houses a D-T neutron generator at its center. The project was two-fold. The first phase of the project involved adaptation of the ANISN code written for an IBM 360/75/91 computer to the UNIVAC system at JSC. The second phase of the project was to use the code with proper geometry, source function and rock material composition in order to determine the neutron flux distribution around the rock box when a 14.1 MeV neutron generator placed at its center is activated.
Energy Technology Data Exchange (ETDEWEB)
Parks, C.V.; Broadhead, B.L.; Hermann, O.W.; Tang, J.S.; Cramer, S.N.; Gauthey, J.C.; Kirk, B.L.; Roussin, R.W.
1988-07-01
This report provides a preliminary assessment of the computational tools and existing methods used to obtain radiation dose rates from shielded spent nuclear fuel and high-level radioactive waste (HLW). Particular emphasis is placed on analysis tools and techniques applicable to facilities/equipment designed for the transport or storage of spent nuclear fuel or HLW. Applications to cask transport, storage, and facility handling are considered. The report reviews the analytic techniques for generating appropriate radiation sources, evaluating the radiation transport through the shield, and calculating the dose at a desired point or surface exterior to the shield. Discrete ordinates, Monte Carlo, and point kernel methods for evaluating radiation transport are reviewed, along with existing codes and data that utilize these methods. A literature survey was employed to select a cadre of codes and data libraries to be reviewed. The selection process was based on specific criteria presented in the report. Separate summaries were written for several codes (or family of codes) that provided information on the method of solution, limitations and advantages, availability, data access, ease of use, and known accuracy. For each data library, the summary covers the source of the data, applicability of these data, and known verification efforts. Finally, the report discusses the overall status of spent fuel shielding analysis techniques and attempts to illustrate areas where inaccuracy and/or uncertainty exist. The report notes the advantages and limitations of several analysis procedures and illustrates the importance of using adequate cross-section data sets. Additional work is recommended to enable final selection/validation of analysis tools that will best meet the US Department of Energy's requirements for use in developing a viable HLW management system. 188 refs., 16 figs., 27 tabs.
International Nuclear Information System (INIS)
Marrel, A.
2008-01-01
In the studies of environmental transfer and risk assessment, numerical models are used to simulate, understand and predict the transfer of pollutant. These computer codes can depend on a high number of uncertain input parameters (geophysical variables, chemical parameters, etc.) and can be often too computer time expensive. To conduct uncertainty propagation studies and to measure the importance of each input on the response variability, the computer code has to be approximated by a meta model which is build on an acceptable number of simulations of the code and requires a negligible calculation time. We focused our research work on the use of Gaussian process meta model to make the sensitivity analysis of the code. We proposed a methodology with estimation and input selection procedures in order to build the meta model in the case of a high number of inputs and with few simulations available. Then, we compared two approaches to compute the sensitivity indices with the meta model and proposed an algorithm to build prediction intervals for these indices. Afterwards, we were interested in the choice of the code simulations. We studied the influence of different sampling strategies on the predictiveness of the Gaussian process meta model. Finally, we extended our statistical tools to a functional output of a computer code. We combined a decomposition on a wavelet basis with the Gaussian process modelling before computing the functional sensitivity indices. All the tools and statistical methodologies that we developed were applied to the real case of a complex hydrogeological computer code, simulating radionuclide transport in groundwater. (author) [fr
International Nuclear Information System (INIS)
Rattan, D.S.
1993-11-01
NSURE stands for Near-Surface Repository code. NSURE is a performance assessment code. developed for the safety assessment of near-surface disposal facilities for low-level radioactive waste (LLRW). Part one of this report documents the NSURE model, governing equations and formulation of the mathematical models, and their implementation under the SYVAC3 executive. The NSURE model simulates the release of nuclides from an engineered vault, their subsequent transport via the groundwater and surface water pathways tot he biosphere, and predicts the resulting dose rate to a critical individual. Part two of this report consists of a User's manual, describing simulation procedures, input data preparation, output and example test cases
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.
International Nuclear Information System (INIS)
Nagaya, Yasunobu; Okumura, Keisuke; Sakurai, Takeshi; Mori, Takamasa
2017-03-01
In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two Monte Carlo codes MVP (continuous-energy method) and GMVP (multigroup method) have been developed at Japan Atomic Energy Agency. The codes have adopted a vectorized algorithm and have been developed for vector-type supercomputers. They also support parallel processing with a standard parallelization library MPI and thus a speed-up of Monte Carlo calculations can be achieved on general computing platforms. The first and second versions of the codes were released in 1994 and 2005, respectively. They have been extensively improved and new capabilities have been implemented. The major improvements and new capabilities are as follows: (1) perturbation calculation for effective multiplication factor, (2) exact resonant elastic scattering model, (3) calculation of reactor kinetics parameters, (4) photo-nuclear model, (5) simulation of delayed neutrons, (6) generation of group constants. This report describes the physical model, geometry description method used in the codes, new capabilities and input instructions. (author)
Energy Technology Data Exchange (ETDEWEB)
Wilcox, T. P.
1973-09-20
The code ANISN-L solves the one-dimensional, multigroup, time-independent Boltzmann transport equation by the method of discrete ordinates. In problems involving a fissionable system, it can calculate the system multiplication or alpha. In such cases, it is also capable of determining isotopic concentrations, radii, zone widths, or buckling in order to achieve a given multiplication or alpha. The code may also calculate fluxes caused by a specified fixed source. Neutron, gamma, and coupled neutron--gamma problems may be solved in either the forward or adjoint (backward) modes. Cross sections describing upscatter, as well as the usual downscatter, may be employed. This report describes the use of ANISN-L; this is a revised version of ANISN which handles both large and small problems efficiently on CDC-7600 computers. (RWR)
Energy Technology Data Exchange (ETDEWEB)
Chang H. Oh; Eung S. Kim; Mike Patterson
2010-06-01
Abstract – A tritium permeation analyses code (TPAC) was developed by Idaho National Laboratory for the purpose of analyzing tritium distributions in very high temperature reactor (VHTR) systems, including integrated hydrogen production systems. A MATLAB SIMULINK software package was used in developing the code. The TPAC is based on the mass balance equations of tritium-containing species and various forms of hydrogen coupled with a variety of tritium sources, sinks, and permeation models. In the TPAC, ternary fission and neutron reactions with 6Li, 7Li 10B, and 3He were taken into considerations as tritium sources. Purification and leakage models were implemented as main tritium sinks. Permeation of tritium and H2 through pipes, vessels, and heat exchangers were considered as main tritium transport paths. In addition, electroyzer and isotope exchange models were developed for analyzing hydrogen production systems, including high temperature electrolysis and sulfur-iodine processes.
International Nuclear Information System (INIS)
Simmons, C.S.; Cole, C.R.
1985-05-01
This document was written for the National Low-Level Waste Management Program to provide guidance for managers and site operators who need to select ground-water transport codes for assessing shallow-land burial site performance. The guidance given in this report also serves the needs of applications-oriented users who work under the direction of a manager or site operator. The guidelines are published in two volumes designed to support the needs of users having different technical backgrounds. An executive summary, published separately, gives managers and site operators an overview of the main guideline report. This volume includes specific recommendations for decision-making managers and site operators on how to use these guidelines. The more detailed discussions about the code selection approach are provided. 242 refs., 6 figs
Norman, Ryan B.; Badavi, Francis F.; Blattnig, Steve R.; Atwell, William
2011-01-01
A deterministic suite of radiation transport codes, developed at NASA Langley Research Center (LaRC), which describe the transport of electrons, photons, protons, and heavy ions in condensed media is used to simulate exposures from spectral distributions typical of electrons, protons and carbon-oxygen-sulfur (C-O-S) trapped heavy ions in the Jovian radiation environment. The particle transport suite consists of a coupled electron and photon deterministic transport algorithm (CEPTRN) and a coupled light particle and heavy ion deterministic transport algorithm (HZETRN). The primary purpose for the development of the transport suite is to provide a means for the spacecraft design community to rapidly perform numerous repetitive calculations essential for electron, proton and heavy ion radiation exposure assessments in complex space structures. In this paper, the radiation environment of the Galilean satellite Europa is used as a representative boundary condition to show the capabilities of the transport suite. While the transport suite can directly access the output electron spectra of the Jovian environment as generated by the Jet Propulsion Laboratory (JPL) Galileo Interim Radiation Electron (GIRE) model of 2003; for the sake of relevance to the upcoming Europa Jupiter System Mission (EJSM), the 105 days at Europa mission fluence energy spectra provided by JPL is used to produce the corresponding dose-depth curve in silicon behind an aluminum shield of 100 mils ( 0.7 g/sq cm). The transport suite can also accept ray-traced thickness files from a computer-aided design (CAD) package and calculate the total ionizing dose (TID) at a specific target point. In that regard, using a low-fidelity CAD model of the Galileo probe, the transport suite was verified by comparing with Monte Carlo (MC) simulations for orbits JOI--J35 of the Galileo extended mission (1996-2001). For the upcoming EJSM mission with a potential launch date of 2020, the transport suite is used to compute
International Nuclear Information System (INIS)
Alonso, A.; Gonzalez, C.
1991-01-01
This report refers to the work carried out within the Chair of Nuclear Technology, Polytechnical University of Madrid, in the frame of the project Modelling the Chemical Behaviour of Tellurium Species in the Reactor Pressure Vessel and the Reactor Cooling System under Severe Accident Conditions (Contract N. 3608-88-12 ELISPPC). It is related to the validation and improvement effort on the tellurium chemistry model in the RAFT code. The use of this code was decided during the Second Shared Cost Action Progress Meeting on Modelling and Code Development, as it seemed to be the best model among those available to the Chair of Nuclear Technology. The improvement effort consists in the inclusion of new tellurium species both in the gaseous and condensed phase; in the incorporation of a simple model for the interaction of tellurium with silver aerosols and finally in some other minor changes, such as the introduction of new variables to compute separately chemisorption and condensation of Te 2 onto walls, and the detection and correction of some Fortran errors found in the as received version of the code. The validation effort reported in this document includes the analysis of the MARVIKEN 4 and 7 experiments with the modified/improved versions of RAFT 1.0 and RAFT 1.1
JH, Summerfield; MW, Manley
2016-01-01
A simple simulation of chemical species movement is presented. The species traverse a Nafion membrane in a fuel cell. Three cells are examined: direct methanol, direct ethanol, and direct glucose. The species are tracked using excess proton concentration, electric field strength, and voltage. The Matlab computer code is provided.
76 FR 2744 - Disclosure of Code-Share Service by Air Carriers and Sellers of Air Transportation
2011-01-14
... global distribution systems, which may be assisting travel agents to establish airline ticket sales Web... intended to provide a reminder to ticket agents with respect to their code-share disclosure responsibility... carriers, foreign air carriers and ticket agents, added a new section 41712(c) that specifically requires...
International Nuclear Information System (INIS)
1997-01-01
The Network Code defines the rights and responsibilities of all users of the natural gas transportation system in the liberalised gas industry in the United Kingdom. This report describes the operation of the Code, what it means, how it works and its implications for the various participants in the industry. The topics covered are: development of the competitive gas market in the UK; key points in the Code; gas transportation charging; impact of the Code on producers upstream; impact on shippers; gas storage; supply point administration; impact of the Code on end users; the future. (20 tables; 33 figures) (UK)
International Nuclear Information System (INIS)
Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.
1977-11-01
The report documents the computer code block VENTURE designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P 1 ) in up to three-dimensional geometry. It uses and generates interface data files adopted in the cooperative effort sponsored by the Reactor Physics Branch of the Division of Reactor Research and Development of the Energy Research and Development Administration. Several different data handling procedures have been incorporated to provide considerable flexibility; it is possible to solve a wide variety of problems on a variety of computer configurations relatively efficiently
Energy Technology Data Exchange (ETDEWEB)
Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.
1977-11-01
The report documents the computer code block VENTURE designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P/sub 1/) in up to three-dimensional geometry. It uses and generates interface data files adopted in the cooperative effort sponsored by the Reactor Physics Branch of the Division of Reactor Research and Development of the Energy Research and Development Administration. Several different data handling procedures have been incorporated to provide considerable flexibility; it is possible to solve a wide variety of problems on a variety of computer configurations relatively efficiently.
International Nuclear Information System (INIS)
Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.
1975-10-01
The computer code block VENTURE, designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P 1 ) in up to three-dimensional geometry is described. A variety of types of problems may be solved: the usual eigenvalue problem, a direct criticality search on the buckling, on a reciprocal velocity absorber (prompt mode), or on nuclide concentrations, or an indirect criticality search on nuclide concentrations, or on dimensions. First-order perturbation analysis capability is available at the macroscopic cross section level
Energy Technology Data Exchange (ETDEWEB)
Chepe P, M. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico); Xolocostli M, J. V.; Gomez T, A. M. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E., E-mail: liaison.web@gmail.com [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. San Pedro Zacatenco, 07730 Ciudad de Mexico (Mexico)
2016-09-15
Due to the current computing power, the deterministic codes for analyzing nuclear reactors that have been used for several years are becoming more relevant, since much more precise solution techniques can be used; the last century would have been very difficult, since memory and processor capacities were very limited or had high prices on the components. In this work we analyze the effect of the anisotropic dispersion of the effective dispersion section, compared to the isotropic dispersion. The anisotropy implementation was carried out in the AZTRAN transport code, which is part of the AZTLAN platform for nuclear reactors analysis (in development). The AZTRAN code solves the Boltzmann transport equation in one, two and three dimensions at steady state, using the multi-group technique for energy discretization, the RTN-0 nodal method in spatial discretization and for angular discretization the discrete ordinates without considering anisotropy originally. The effect of the anisotropy dispersion on the effective multiplication factor and the axial and radial power on a fuel assembly BWR type are analyzed. (Author)
Verbeke, Jérôme M.; Petit, Odile; Chebboubi, Abdelhazize; Litaize, Olivier
2018-01-01
Fission modeling in general-purpose Monte Carlo transport codes often relies on average nuclear data provided by international evaluation libraries. As such, only average fission multiplicities are available and correlations between fission neutrons and photons are missing. Whereas uncorrelated fission physics is usually sufficient for standard reactor core and radiation shielding calculations, correlated fission secondaries are required for specialized nuclear instrumentation and detector modeling. For coincidence counting detector optimization for instance, precise simulation of fission neutrons and photons that remain correlated in time from birth to detection is essential. New developments were recently integrated into the Monte Carlo transport code TRIPOLI-4 to model fission physics more precisely, the purpose being to access event-by-event fission events from two different fission models: FREYA and FIFRELIN. TRIPOLI-4 simulations can now be performed, either by connecting via an API to the LLNL fission library including FREYA, or by reading external fission event data files produced by FIFRELIN beforehand. These new capabilities enable us to easily compare results from Monte Carlo transport calculations using the two fission models in a nuclear instrumentation application. In the first part of this paper, broad underlying principles of the two fission models are recalled. We then present experimental measurements of neutron angular correlations for 252Cf(sf) and 240Pu(sf). The correlations were measured for several neutron kinetic energy thresholds. In the latter part of the paper, simulation results are compared to experimental data. Spontaneous fissions in 252Cf and 240Pu are modeled by FREYA or FIFRELIN. Emitted neutrons and photons are subsequently transported to an array of scintillators by TRIPOLI-4 in analog mode to preserve their correlations. Angular correlations between fission neutrons obtained independently from these TRIPOLI-4 simulations, using
International Nuclear Information System (INIS)
Jimenez P, D. A.
2014-01-01
The accidents in Unit 2 of the Three Mile Island Nuclear Power Plant (NPP) in the United States (March 28 th , 1979), the one in Unit 4 of the NPP Chernobyl in Ukraine (April 26 th , 1986) and the explosions in some units of Fukushima NPP in Japan (March 11 th , 2011) boosted the investigations on severe accidents with core damage and, in particular, the threat to the ultimate barrier by an eventual explosion from uncontrolled Hydrogen combustion within the containment was considered of particular relevance. Research programs for analyzing Hydrogen behavior and control during this kind of accidents were early initiated by research and regulatory bodies. Assessment on Hydrogen behavior once it has been postulated to be released on the containment system can be divided into two phases, in the first one, transport and the concentrations of the gas mixtures and steam in each volume or area comprised between the structures of the containment are calculated, in the second one, the propagation of the detonation of the Hydrogen is calculated if there are the conditions to occur. Currently, there are computer programs that can be used in one, or both stages of computation, and they are based on one of the two solution methods in current use, one of them are integrated codes (e.g. MELCOR), which consists in assuming the containment as a network composed of hydraulic tanks or nodes on which the balance equations of mass and energy have to be solved, the network is connected by ducts or connections where the momentum balance equation arise. This methodology relies on the use of semi-empirical relationships and the criteria used to define a geometric pattern, are subjective. The second method, which is having relevance due to the large computing power of modern computers, is the numerical solution of the three-dimensional Navier-Stokes equations in complex geometries. This method of solution is known as Computational Fluid Dynamics (CFD), and offers the advantage of using a
Benchmark of PENELOPE code for low-energy photon transport: dose comparisons with MCNP4 and EGS4
International Nuclear Information System (INIS)
Ye, Sung-Joon; Brezovich, Ivan A; Pareek, Prem; Naqvi, Shahid A
2004-01-01
The expanding clinical use of low-energy photon emitting 125 I and 103 Pd seeds in recent years has led to renewed interest in their dosimetric properties. Numerous papers pointed out that higher accuracy could be obtained in Monte Carlo simulations by utilizing newer libraries for the low-energy photon cross-sections, such as XCOM and EPDL97. The recently developed PENELOPE 2001 Monte Carlo code is user friendly and incorporates photon cross-section data from the EPDL97. The code has been verified for clinical dosimetry of high-energy electron and photon beams, but has not yet been tested at low energies. In the present work, we have benchmarked the PENELOPE code for 10-150 keV photons. We computed radial dose distributions from 0 to 10 cm in water at photon energies of 10-150 keV using both PENELOPE and MCNP4C with either DLC-146 or DLC-200 cross-section libraries, assuming a point source located at the centre of a 30 cm diameter and 20 cm length cylinder. Throughout the energy range of simulated photons (except for 10 keV), PENELOPE agreed within statistical uncertainties (at worst ±5%) with MCNP/DLC-146 in the entire region of 1-10 cm and with published EGS4 data up to 5 cm. The dose at 1 cm (or dose rate constant) of PENELOPE agreed with MCNP/DLC-146 and EGS4 data within approximately ±2% in the range of 20-150 keV, while MCNP/DLC-200 produced values up to 9% lower in the range of 20-100 keV than PENELOPE or the other codes. However, the differences among the four datasets became negligible above 100 keV
Directory of Open Access Journals (Sweden)
François Van Dorpe
2007-01-01
Full Text Available The results of four gas tracer experiments of atmospheric dispersion on a regional scale are used for the benchmarking of two atmospheric dispersion modeling codes, MINERVE-SPRAY (CEA, and NOSTRADAMUS (IBRAE. The main topic of this comparison is to estimate the Lagrangian code capability to predict the radionuclide atmospheric transfer on a large field, in the case of risk assessment of nuclear power plant for example. For the four experiments, the results of calculations show a rather good agreement between the two codes, and the order of magnitude of the concentrations measured on the soil is predicted. Simulation is best for sampling points located ten kilometers from the source, while we note a divergence for more distant points results (difference in concentrations by a factor 2 to 5. This divergence may be explained by the fact that, for these four experiments, only one weather station (near the point source was used on a field of 10 000 km2, generating the simulation of a uniform wind field throughout the calculation domain.
International Nuclear Information System (INIS)
Roussin, R.W.
1993-01-01
From the very early days in its history Radiation Shielding Information Center (RSIC) has been involved with high energy radiation transport. The National Aeronautics and Space Administration was an early sponsor of RSIC until the completion of the Apollo Moon Exploration Program. In addition, the intranuclear cascade work of Bertini at Oak Ridge National Laboratory provided valuable resources which were made available through RSIC. Over the years, RSIC has had interactions with many of the developers of high energy radiation transport computing technology and data libraries and has been able to collect and disseminate this technology. The current status of this technology will be reviewed and prospects for new advancements will be examined
Wareing, Todd A.; Failla, Gregory; Horton, John L.; Eifel, Patricia J.; Mourtada, Firas
2009-01-01
A patient dose distribution was calculated by a 3D multi‐group SN particle transport code for intracavitary brachytherapy of the cervix uteri and compared to previously published Monte Carlo results. A Cs‐137 LDR intracavitary brachytherapy CT data set was chosen from our clinical database. MCNPX version 2.5.c, was used to calculate the dose distribution. A 3D multi‐group SN particle transport code, Attila version 6.1.1 was used to simulate the same patient. Each patient applicator was built in SolidWorks, a mechanical design package, and then assembled with a coordinate transformation and rotation for the patient. The SolidWorks exported applicator geometry was imported into Attila for calculation. Dose matrices were overlaid on the patient CT data set. Dose volume histograms and point doses were compared. The MCNPX calculation required 14.8 hours, whereas the Attila calculation required 22.2 minutes on a 1.8 GHz AMD Opteron CPU. Agreement between Attila and MCNPX dose calculations at the ICRU 38 points was within ±3%. Calculated doses to the 2 cc and 5 cc volumes of highest dose differed by not more than ±1.1% between the two codes. Dose and DVH overlays agreed well qualitatively. Attila can calculate dose accurately and efficiently for this Cs‐137 CT‐based patient geometry. Our data showed that a three‐group cross‐section set is adequate for Cs‐137 computations. Future work is aimed at implementing an optimized version of Attila for radiotherapy calculations. PACS number: 87.53.Jw
Fowler, S. J.; Driesner, T.; Hingerl, F. F.; Kulik, D. A.; Wagner, T.
2011-12-01
We apply a new, C++-based computational model for hydrothermal fluid-rock interaction and scale formation in geothermal reservoirs. The model couples the Complex System Modelling Platform (CSMP++) code for fluid flow in porous and fractured media (Matthai et al., 2007) with the Gibbs energy minimization numerical kernel GEMS3K of the GEM-Selektor (GEMS3) geochemical modelling package (Kulik et al., 2010) in a modular fashion. CSMP++ includes interfaces to commercial file formats, accommodating complex geometry construction using CAD (Rhinoceros) and meshing (ANSYS) software. The CSMP++ approach employs finite element-finite volume spatial discretization, implicit or explicit time discretization, and operator splitting. GEMS3K can calculate complex fluid-mineral equilibria based on a variety of equation of state and activity models. A selection of multi-electrolyte aqueous solution models, such as extended Debye-Huckel, Pitzer (Harvie et al., 1984), EUNIQUAC (Thomsen et al., 1996), and the new ELVIS model (Hingerl et al., this conference), makes it well-suited for application to a wide range of geothermal conditions. An advantage of the GEMS3K solver is simultaneous consideration of complex solid solutions (e.g., clay minerals), gases, fluids, and aqueous solutions. Each coupled simulation results in a thermodynamically-based description of the geochemical and physical state of a hydrothermal system evolving along a complex P-T-X path. The code design allows efficient, flexible incorporation of numerical and thermodynamic database improvements. We demonstrate the coupled code workflow and applicability to compositionally and physically complex natural systems relevant to enhanced geothermal systems, where temporally and spatially varying chemical interactions may take place within diverse lithologies of varying geometry. Engesgaard, P. & Kipp, K. L. (1992). Water Res. Res. 28: 2829-2843. Harvie, C. E.; Møller, N. & Weare, J. H. (1984). Geochim. Cosmochim. Acta 48
International Nuclear Information System (INIS)
Anon.
1998-01-01
Here is the decree of the thirtieth of July 1998 relative to road transportation, to trade and brokerage of wastes. It requires to firms which carry out a road transportation as well as to traders and to brokers of wastes to declare their operations to the prefect. The declaration has to be renewed every five years. (O.M.)
Energy Technology Data Exchange (ETDEWEB)
Gomez T, A. M.; Xolocostli M, V. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Lopez M, R.; Filio L, C.; Mugica R, C. A. [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico); Royl, P., E-mail: armando.gomez@inin.gob.mx [Karlsruhe Institute of Technology, Consultor, Hermann-von-Helmholtz-Platz, D-76344 Eggenstein -Leopoldshafen, Karlsruhe (Germany)
2013-10-15
The scenario of electric power total loss in the nuclear power plant of Laguna Verde (NPP-L V) has been analyzed using the code MELCOR previously, until reaching fault conditions of the primary container. A mitigation measure to avoid the loss of the primary contention is the realization of a venting toward the secondary contention (reactor building), however this measure bears the potential explosions occurrence risk when the hydrogen accumulated in the primary container with the oxygen of the reactor building atmosphere reacting. In this work a scenario has been supposed that considers the mentioned venting when the pressure of 4.5 kg/cm{sup 2} is reached in the primary container. The information for the hydrogen like an entrance fact is obtained of the MELCOR results and the hydrogen transport in both contentions is analyzed with the code CFD GASFLOW that allows predicting the detailed distribution of the hydrogen volumetric concentration and the possible detonation of flammability conditions in the reactor building. The results show that the venting will produce detonation conditions in the venting level (level 33) and flammability in the level of the recharge floor. The methodology here described constitutes the base of a detailed calculation system of this type of phenomena that can use to make safety evaluations in the NPP-L V on scenarios that include gases transport. (Author)
International Nuclear Information System (INIS)
Beckert, C.
2007-01-01
Conventionally the data preparation of the neutron cross sections for reactor-core calculations pursues with 2D cell codes. Aim of this thesis was, to develop a 3D cell code, to study with this code 3D effects, and to evaluate the necessarity of a 3D data preparation of the neutron cross sections. For the calculation of the neutron transport the method of the first-collision probabilities, which are calculated with the ray-tracing method, was chosen. The mathematical algorithms were implemented in the 2D/3D cell code TransRay. For the geometry part of the program the geometry module of a Monte Carlo code was used.The ray tracing in 3D was parallelized because of the high computational time. The program TransRay was verified on 2D test problems. For a reference pressured-water reactor following 3D problems were studied: A partly immersed control rod and void (vacuum or steam) around a fuel rod as model of a steam void. All problems were for comparison calculated also with the programs HELIOS(2D) and MCNP(3D). The dependence of the multiplication factor and the averaged two-group cross section on the immersion depth of the control rod respectively of the height of the steam void were studied. The 3D-calculated two-group cross sections were compared with three conventional approximations: Linear interpolation, interpolation with flux weighting, and homogenization, At the 3D problem of the control rod it was shown that the interpolation with flux weighting is a good approximation. Therefore here a 3D data preparation is not necessary. At the test case of the single control rod, which is surrounded by the void, the three approximation for the two-group cross sections were proved as unsufficient. Therefore a 3D data preparation is necessary. The single fuel-rod cell with void can be considered as the limiting case of a reactor, in which a phase interface has been formed [de
Rodriguez, M.; Brualla, L.
2018-04-01
Monte Carlo simulation of radiation transport is computationally demanding to obtain reasonably low statistical uncertainties of the estimated quantities. Therefore, it can benefit in a large extent from high-performance computing. This work is aimed at assessing the performance of the first generation of the many-integrated core architecture (MIC) Xeon Phi coprocessor with respect to that of a CPU consisting of a double 12-core Xeon processor in Monte Carlo simulation of coupled electron-photonshowers. The comparison was made twofold, first, through a suite of basic tests including parallel versions of the random number generators Mersenne Twister and a modified implementation of RANECU. These tests were addressed to establish a baseline comparison between both devices. Secondly, through the p DPM code developed in this work. p DPM is a parallel version of the Dose Planning Method (DPM) program for fast Monte Carlo simulation of radiation transport in voxelized geometries. A variety of techniques addressed to obtain a large scalability on the Xeon Phi were implemented in p DPM. Maximum scalabilities of 84 . 2 × and 107 . 5 × were obtained in the Xeon Phi for simulations of electron and photon beams, respectively. Nevertheless, in none of the tests involving radiation transport the Xeon Phi performed better than the CPU. The disadvantage of the Xeon Phi with respect to the CPU owes to the low performance of the single core of the former. A single core of the Xeon Phi was more than 10 times less efficient than a single core of the CPU for all radiation transport simulations.
Energy Technology Data Exchange (ETDEWEB)
Hilbert, F. [Nuclear Cargo and Service GmbH, Hanau (Germany); Morishima, M.; Tamaki, H. [Mitsubishi Heavy Industries, Kobe (Japan)
2004-07-01
For the shielding analysis of transport and storage casks for spent fuel the use of computer codes is state of the art. However, in most applications the computer models used for the analysis are simplified to circular geometries to save modelling effort and calculation time. Furthermore, the active zone of the fuel is modelled as homogeneous zone with a uniform average burn-up. In the first part of the present paper it is shown that an exact model is feasible and the effect of the geometrical shape on the dose rates is illustrated. The second part of the paper shows the comparison of the dose rates calculated with 5 different fuel models. Finally, the accuracy of the calculations is discussed.
Energy Technology Data Exchange (ETDEWEB)
Lee, Ki Bog; Kim, Yeong Il; Kim, Kang Seok; Kim, Sang Ji; Kim, Young Gyun; Song, Hoon; Lee, Dong Uk; Lee, Byoung Oon; Jang, Jin Wook; Lim, Hyun Jin; Kim, Hak Sung
2004-05-01
In this report, the results of KALIMER (Korea Advanced LIquid MEtal Reactor) core design calculated by the K-CORE computing system are compared and analyzed with those of MCDEP calculation. The effective multiplication factor, flux distribution, fission power distribution and the number densities of the important nuclides effected from the depletion calculation for the R-Z model and Hex-Z model of KALIMER core are compared. It is confirmed that the results of K-CORE system compared with those of MCDEP based on the Monte Carlo transport theory method agree well within 700 pcm for the effective multiplication factor estimation and also within 2% in the driver fuel region, within 10% in the radial blanket region for the reaction rate and the fission power density. Thus, the K-CORE system for the core design of KALIMER by treating the lumped fission product and mainly important nuclides can be used as a core design tool keeping the necessary accuracy.
National Research Council Canada - National Science Library
Allshouse, Michael; Armstrong, Frederick Henry; Burns, Stephen; Courts, Michael; Denn, Douglas; Fortunato, Paul; Gettings, Daniel; Hansen, David; Hoffman, Douglas; Jones, Robert
2007-01-01
.... The ability of the global transportation industry to rapidly move passengers and products from one corner of the globe to another continues to amaze even those wise to the dynamics of such operations...
Energy Technology Data Exchange (ETDEWEB)
Onishi, Yasuo; Yokuda, Satoru T.; Kurikami, Hiroshi
2014-03-28
The accident at the Fukushima Daiichi Nuclear Power Plant in March 2011 caused widespread environmental contamination. Although decontamination activities have been performed in residential areas of the Fukushima area, decontamination of forests, rivers, and reservoirs is still controversial because of the economical, ecological, and technical difficulties. Thus, an evaluation of contaminant transport in such an environment is important for safety assessment and for implementation of possible countermeasures to reduce radiation exposure to the public. The investigation revealed that heavy rainfall events play a significant role in transporting radioactive cesium deposited on the land surface, via soil erosion and sediment transport in rivers. Therefore, we simulated the sediment and cesium transport in the Ukedo River and its tributaries in Fukushima Prefecture, including the Ogaki Dam Reservoir, and the Ogi Dam Reservoir of the Oginosawa River in Fukushima Prefecture during and after a heavy rainfall event by using the TODAM (Time-dependent, One-dimensional Degradation And Migration) code. The main outcomes are the following: • Suspended sand is mostly deposited on the river bottom. Suspended silt and clay, on the other hand, are hardly deposited in the Ukedo River and its tributaries except in the Ogaki Dam Reservoir in the Ukedo River even in low river discharge conditions. • Cesium migrates mainly during high river discharge periods during heavy rainfall events. Silt and clay play more important roles in cesium transport to the sea than sand does. • The simulation results explain variations in the field data on cesium distributions in the river. Additional field data currently being collected and further modeling with these data may shed more light on the cesium distribution variations. • Effects of 40-hour heavy rainfall events on clay and cesium transport continue for more than a month. This is because these reservoirs slow down the storm-induced high
International Nuclear Information System (INIS)
Aquino Branco, O.E. de
1995-01-01
This paper presents a methodology for the evaluation of the maximum radionuclides concentrations in groundwaters, due to the transport/migration of decay chains. Analytical solution of the equations system is difficult, even if only three elements of the decay chain are considered. Therefore, a numerical solution is most convenient. An application of the MICRAS code, developed to assess maximum concentrations of each radionuclide, starting with the initial concentrations, is presented. The maximum concentration profile for 226 Ra, calculated using MICRAS, is compared with the results obtained through an analytical and a numerical model. The fitness of results is considered good. Simplified models, like the one represented by the application of MICRAS, are largely employed in the section in the selection and characterization of sites for radioactive wastes repositories and in studies of safety evaluation for the same purpose. A detailed analysis of the transport/migration of contaminants in aquifers requires a large quantify of data from the site and from the installation as well, which makes this analysis expensive and inviable during the preliminary phases of the studies. (author). 6 refs, 1 fig, 1 tab
International Nuclear Information System (INIS)
Riggle, K.J.; Roddy, J.W.
1989-02-01
This work is part of the ongoing Systems Modeling Program at Oak Ridge National Laboratory, which is assisting the DOE Office of Civilian Radioactive Waste Management in selecting appropriate computer codes for the process of licensing a high-level radioactive waste repository or a monitored retrievable storage facility. A preliminary study of codes for predicting dose to man following airborne releases of radionuclides is described. These codes use models for estimating atmospheric dispersion of activity and deposition onto the ground surface, exposures via external irradiation, inhalation of airborne activity, and ingestion following transport through terrestrial food chains, and the dose per unit exposure for each exposure mode. A set of criteria is given for use in choosing codes for further examination. From a list of over 150 computer codes, five were selected for review
Maina, Fadji Hassane; Ackerer, Philippe; Younes, Anis; Guadagnini, Alberto; Berkowitz, Brian
2017-06-07
We present a combined experimental and numerical modeling study that addresses two principal questions: (i) is any particular Eulerian-based method used to solve the classical advection-dispersion equation (ADE) clearly superior (relative to the others), in terms of yielding solutions that reproduce BTCs of the kind that are typically sampled at the outlet of a laboratory cell? and (ii) in the presence of matches of comparable quality against such BTCs, do any of these methods render different (or similar) numerical BTCs at locations within the domain? To address these questions, we obtained measurements from carefully controlled laboratory experiments, and employ them as a reference against which numerical results are benchmarked and compared. The experiments measure solute transport breakthrough curves (BTCs) through a square domain containing various configurations of coarse, medium, and fine quartz sand. The approaches to solve the ADE involve Eulerian-Lagrangian and Eulerian (finite volume, finite elements, mixed and discontinuous finite elements) numerical methods. Model calibration is not examined; permeability and porosity of each sand were determined previously through separate, standard laboratory tests, while dispersivities are assigned values proportional to mean grain size. We find that the spatial discretization of the flow field is of critical importance, due to the non-uniformity of the domain. Although simulated BTCs at the system outlet are observed to be very similar for these various numerical methods, computed local (point-wise, inside the domain) BTCs can be very different. We find that none of the numerical methods is able to fully reproduce the measured BTCs. The impact of model parameter uncertainty on the calculated BTCs is characterized through a set of numerical Monte Carlo simulations; in cases where the impact is significant, assessment of simulation matches to the experimental data can be ambiguous. Copyright © 2017 Elsevier B.V. All
Energy Technology Data Exchange (ETDEWEB)
Nakka, B.W.; Chan, T.
1994-12-01
A deterministic particle-tracking code (TRACK3D) has been developed to compute convective flow paths of conservative (nonreactive) contaminants through porous geological media. TRACK3D requires the groundwater velocity distribution, which, in our applications, results from flow simulations using AECL`s MOTIF code. The MOTIF finite-element code solves the transient and steady-state coupled equations of groundwater flow, solute transport and heat transport in fractured/porous media. With few modifications, TRACK3D can be used to analyse the velocity distributions calculated by other finite-element or finite-difference flow codes. This report describes the assumptions, limitations, organization, operation and applications of the TRACK3D code, and provides a comprehensive user`s manual.
Picot-Colbeaux, Géraldine; Devau, Nicolas; Thiéry, Dominique; Pettenati, Marie; Surdyk, Nicolas; Parmentier, Marc; Amraoui, Nadia; Crastes de Paulet, François; André, Laurent
2016-04-01
Chalk aquifer is the main water resource for domestic water supply in many parts in northern France. In same basin, groundwater is frequently affected by quality problems concerning nitrates. Often close to or above the drinking water standards, nitrate concentration in groundwater is mainly due to historical agriculture practices, combined with leakage and aquifer recharge through the vadose zone. The complexity of processes occurring into such an environment leads to take into account a lot of knowledge on agronomy, geochemistry and hydrogeology in order to understand, model and predict the spatiotemporal evolution of nitrate content and provide a decision support tool for the water producers and stakeholders. To succeed in this challenge, conceptual and numerical models representing accurately the Chalk aquifer specificity need to be developed. A multidisciplinary approach is developed to simulate storage and transport from the ground surface until groundwater. This involves a new agronomic module "NITRATE" (NItrogen TRansfer for Arable soil to groundwaTEr), a soil-crop model allowing to calculate nitrogen mass balance in arable soil, and the "PHREEQC" numerical code for geochemical calculations, both coupled with the 3D transient groundwater numerical code "MARTHE". Otherwise, new development achieved on MARTHE code allows the use of dual porosity and permeability calculations needed in the fissured Chalk aquifer context. This method concerning the integration of existing multi-disciplinary tools is a real challenge to reduce the number of parameters by selecting the relevant equations and simplifying the equations without altering the signal. The robustness and the validity of these numerical developments are tested step by step with several simulations constrained by climate forcing, land use and nitrogen inputs over several decades. In the first time, simulations are performed in a 1D vertical unsaturated soil column for representing experimental nitrates
Energy Technology Data Exchange (ETDEWEB)
Cramer, S.N.
1984-01-01
The MORSE code is a large general-use multigroup Monte Carlo code system. Although no claims can be made regarding its superiority in either theoretical details or Monte Carlo techniques, MORSE has been, since its inception at ORNL in the late 1960s, the most widely used Monte Carlo radiation transport code. The principal reason for this popularity is that MORSE is relatively easy to use, independent of any installation or distribution center, and it can be easily customized to fit almost any specific need. Features of the MORSE code are described.
International Nuclear Information System (INIS)
Cramer, S.N.
1984-01-01
The MORSE code is a large general-use multigroup Monte Carlo code system. Although no claims can be made regarding its superiority in either theoretical details or Monte Carlo techniques, MORSE has been, since its inception at ORNL in the late 1960s, the most widely used Monte Carlo radiation transport code. The principal reason for this popularity is that MORSE is relatively easy to use, independent of any installation or distribution center, and it can be easily customized to fit almost any specific need. Features of the MORSE code are described
Beam-dynamics codes used at DARHT
Energy Technology Data Exchange (ETDEWEB)
Ekdahl, Jr., Carl August [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2017-02-01
Several beam simulation codes are used to help gain a better understanding of beam dynamics in the DARHT LIAs. The most notable of these fall into the following categories: for beam production – Tricomp Trak orbit tracking code, LSP Particle in cell (PIC) code, for beam transport and acceleration – XTR static envelope and centroid code, LAMDA time-resolved envelope and centroid code, LSP-Slice PIC code, for coasting-beam transport to target – LAMDA time-resolved envelope code, LSP-Slice PIC code. These codes are also being used to inform the design of Scorpius.
International Nuclear Information System (INIS)
Peek, J.M.; Halbleib, J.A.
1983-01-01
The inner-shell ionization data for electron-target collisions now in use in the TIGER and TIGERP electron-transport codes are extracted and compared with other data for these processes. The TIGER cross sections for K-shell ionization by electron collisions are found to be seriously in error for large-Z targets and incident electron energies greater than 1 MeV. A series of TIGER and TIGERP runs were carried out with and without improved K-shell electron ionization cross section data replacing that now in use. The relative importance of electron-impact and photon ionization of the various subshells was also extracted from these runs. In general, photon ionization dominated in the examples studied so the sensitivity of many predicted properties to errors in the electron-impact subshell ionization data was not large. However, some differences were found and, as all possible applications were not covered in this study, it is recommended that these electron-impact data now in TIGER and TIGERP be replaced. Cross section data for the processes under study are reviewed and those that are most suitable for this application are identified. 19 references, 9 figures, 2 tables
Directory of Open Access Journals (Sweden)
Fabio Burderi
2007-05-01
Full Text Available Motivated by the study of decipherability conditions for codes weaker than Unique Decipherability (UD, we introduce the notion of coding partition. Such a notion generalizes that of UD code and, for codes that are not UD, allows to recover the ``unique decipherability" at the level of the classes of the partition. By tacking into account the natural order between the partitions, we define the characteristic partition of a code X as the finest coding partition of X. This leads to introduce the canonical decomposition of a code in at most one unambiguouscomponent and other (if any totally ambiguouscomponents. In the case the code is finite, we give an algorithm for computing its canonical partition. This, in particular, allows to decide whether a given partition of a finite code X is a coding partition. This last problem is then approached in the case the code is a rational set. We prove its decidability under the hypothesis that the partition contains a finite number of classes and each class is a rational set. Moreover we conjecture that the canonical partition satisfies such a hypothesis. Finally we consider also some relationships between coding partitions and varieties of codes.
International Nuclear Information System (INIS)
Alali, Abdullah
2014-01-01
The one-group interfacial area transport equation has been coupled to a wall heat flux partitioning model in the framework of two-phase Eulerian approach using the OpenFOAM CFD code for better prediction of subcooled boiling phenomena which is essential for safety analysis of nuclear reactors. The interfacial area transport equation has been modified to include the effect of bubble nucleation at the wall and condensation by subcooled liquid in the bulk that governs the non-uniform bubble size distribution.
International Nuclear Information System (INIS)
Dershowitz, W; Herbert, A.; Long, J.
1989-03-01
The hydrology of the SCV site will be modelled utilizing discrete fracture flow models. These models are complex, and can not be fully cerified by comparison to analytical solutions. The best approach for verification of these codes is therefore cross-verification between different codes. This is complicated by the variation in assumptions and solution techniques utilized in different codes. Cross-verification procedures are defined which allow comparison of the codes developed by Harwell Laboratory, Lawrence Berkeley Laboratory, and Golder Associates Inc. Six cross-verification datasets are defined for deterministic and stochastic verification of geometric and flow features of the codes. Additional datasets for verification of transport features will be documented in a future report. (13 figs., 7 tabs., 10 refs.) (authors)
DEFF Research Database (Denmark)
Ejsing-Duun, Stine; Hansbøl, Mikala
Sammenfatning af de mest væsentlige pointer fra hovedrapporten: Dokumentation og evaluering af Coding Class......Sammenfatning af de mest væsentlige pointer fra hovedrapporten: Dokumentation og evaluering af Coding Class...
Energy Technology Data Exchange (ETDEWEB)
Moridis, G.; Pruess, K.
1995-04-01
This report discusses the details of modifications made to the TOUGH2 family of codes to complement its direct solver which significantly increases the size of problems solved by the TOUGH2 code. With this modification, the TOUGH2 system is being tested in multiphase, multicomponent fluid and heat flow problems related to vadose zone hydrology, nuclear waste disposal, and environmental remediation.
Ampomah, Osei Yaw; Avetisyan, Anna; Hansen, Espen; Svenson, Johan; Huser, Thomas; Jensen, John Beck; Bhuvaneswari, T V
2013-09-01
The thu operon (thuEFGKAB) in Sinorhizobium meliloti codes for transport and utilization functions of the disaccharide trehalose. Sequenced genomes of members of the Rhizobiaceae reveal that some rhizobia and Agrobacterium possess the entire thu operon in similar organizations and that Mesorhizobium loti MAFF303099 lacks the transport (thuEFGK) genes. In this study, we show that this operon is dedicated to the transport and assimilation of maltitol and isomers of sucrose (leucrose, palatinose, and trehalulose) in addition to trehalulose, not only in S. meliloti but also in Agrobacterium tumefaciens. By using genetic complementation, we show that the thuAB genes of S. meliloti, M. loti, and A. tumefaciens are functionally equivalent. Further, we provide both genetic and biochemical evidence to show that these bacteria assimilate these disaccharides by converting them to their respective 3-keto derivatives and that the thuAB genes code for this ketodisaccharide-forming enzyme(s). Formation of 3-ketotrehalose in real time in live S. meliloti is shown through Raman spectroscopy. The presence of an additional ketodisaccharide-forming pathway(s) in A. tumefaciens is also indicated. To our knowledge, this is the first report to identify the genes that code for the conversion of disaccharides to their 3-ketodisaccharide derivatives in any organism.
Geochemical computer codes. A review
International Nuclear Information System (INIS)
Andersson, K.
1987-01-01
In this report a review of available codes is performed and some code intercomparisons are also discussed. The number of codes treating natural waters (groundwater, lake water, sea water) is large. Most geochemical computer codes treat equilibrium conditions, although some codes with kinetic capability are available. A geochemical equilibrium model consists of a computer code, solving a set of equations by some numerical method and a data base, consisting of thermodynamic data required for the calculations. There are some codes which treat coupled geochemical and transport modeling. Some of these codes solve the equilibrium and transport equations simultaneously while other solve the equations separately from each other. The coupled codes require a large computer capacity and have thus as yet limited use. Three code intercomparisons have been found in literature. It may be concluded that there are many codes available for geochemical calculations but most of them require a user that us quite familiar with the code. The user also has to know the geochemical system in order to judge the reliability of the results. A high quality data base is necessary to obtain a reliable result. The best results may be expected for the major species of natural waters. For more complicated problems, including trace elements, precipitation/dissolution, adsorption, etc., the results seem to be less reliable. (With 44 refs.) (author)
Monte Carlo code development in Los Alamos
International Nuclear Information System (INIS)
Carter, L.L.; Cashwell, E.D.; Everett, C.J.; Forest, C.A.; Schrandt, R.G.; Taylor, W.M.; Thompson, W.L.; Turner, G.D.
1974-01-01
The present status of Monte Carlo code development at Los Alamos Scientific Laboratory is discussed. A brief summary is given of several of the most important neutron, photon, and electron transport codes. 17 references. (U.S.)
Induction technology optimization code
International Nuclear Information System (INIS)
Caporaso, G.J.; Brooks, A.L.; Kirbie, H.C.
1992-01-01
A code has been developed to evaluate relative costs of induction accelerator driver systems for relativistic klystrons. The code incorporates beam generation, transport and pulsed power system constraints to provide an integrated design tool. The code generates an injector/accelerator combination which satisfies the top level requirements and all system constraints once a small number of design choices have been specified (rise time of the injector voltage and aspect ratio of the ferrite induction cores, for example). The code calculates dimensions of accelerator mechanical assemblies and values of all electrical components. Cost factors for machined parts, raw materials and components are applied to yield a total system cost. These costs are then plotted as a function of the two design choices to enable selection of an optimum design based on various criteria. (Author) 11 refs., 3 figs
Energy Technology Data Exchange (ETDEWEB)
Vilches, M. [Servicio de Fisica y Proteccion Radiologica, Hospital Regional Universitario ' Virgen de las Nieves' , Avda. de las Fuerzas Armadas, 2, E-18014 Granada (Spain)], E-mail: mvilches@ugr.es; Garcia-Pareja, S. [Servicio de Radiofisica Hospitalaria, Hospital Regional Universitario ' Carlos Haya' , Avda. Carlos Haya, s/n, E-29010 Malaga (Spain); Guerrero, R. [Servicio de Radiofisica, Hospital Universitario ' San Cecilio' , Avda. Dr. Oloriz, 16, E-18012 Granada (Spain); Anguiano, M.; Lallena, A.M. [Departamento de Fisica Atomica, Molecular y Nuclear, Universidad de Granada, E-18071 Granada (Spain)
2007-09-21
When a therapeutic electron linear accelerator is simulated using a Monte Carlo (MC) code, the tuning of the initial spectra and the renormalization of dose (e.g., to maximum axial dose) constitute a common practice. As a result, very similar depth dose curves are obtained for different MC codes. However, if renormalization is turned off, the results obtained with the various codes disagree noticeably. The aim of this work is to investigate in detail the reasons of this disagreement. We have found that the observed differences are due to non-negligible differences in the angular scattering of the electron beam in very thin slabs of dense material (primary foil) and thick slabs of very low density material (air). To gain insight, the effects of the angular scattering models considered in various MC codes on the dose distribution in a water phantom are discussed using very simple geometrical configurations for the LINAC. The MC codes PENELOPE 2003, PENELOPE 2005, GEANT4, GEANT3, EGSnrc and MCNPX have been used.
Error correcting coding for OTN
DEFF Research Database (Denmark)
Justesen, Jørn; Larsen, Knud J.; Pedersen, Lars A.
2010-01-01
Forward error correction codes for 100 Gb/s optical transmission are currently receiving much attention from transport network operators and technology providers. We discuss the performance of hard decision decoding using product type codes that cover a single OTN frame or a small number...
Indian Academy of Sciences (India)
message symbols downstream, network coding achieves vast performance gains by permitting intermediate nodes to carry out algebraic oper- ations on the incoming data. In this article we present a tutorial introduction to network coding as well as an application to the e±cient operation of distributed data-storage networks.
International Nuclear Information System (INIS)
Gosmain, Cecile-Aline
2011-01-01
In the framework of French research program on Generation IV sodium cooled fast reactor, one possible option consists in burning minor actinides in this kind of Advanced Sodium Technological Reactor. Two types of transmutation mode are studied in the world : the homogeneous mode of transmutation where actinides are scattered with very low enrichment ratio in fissile assemblies and the heterogeneous mode where fissile core is surrounded by blanket assemblies filled with minor actinides with ratio of incorporated actinides up to 20%. Depending on which element is considered to be burnt and on its content, these minor actinides contents imply constraints on assemblies' transportation between Nuclear Power Plants and fuel cycle facilities. In this study, we present some academic studies in order to identify some key constraints linked to the residual power and neutron/gamma load of such kind of blanket assemblies. To simplify the approach, we considered a modeling of a 'model cask' dedicated to the transportation of a unique irradiated blanket assembly loaded with 20% of Americium and basically inspired from an existent cask designed initially for the damaged fissile Superphenix assembly transport. Thermal calculations performed with EDF-SYRTHES code have shown that due to thermal limitations on cladding temperature, the decay time to be considered before transportation is 20 years. This study is based on explicit 3D representations of the cask and the contained blanket assembly with the Monte Carlo code TRIPOLI/JEFF3.1.1 library and concludes that after such a decay time, the transportation of a unique Americium radial blanket is feasible only if the design of our model cask is modified in order to comply with the dose limitation criterion. (author)
Introduction of SCIENCE code package
International Nuclear Information System (INIS)
Lu Haoliang; Li Jinggang; Zhu Ya'nan; Bai Ning
2012-01-01
The SCIENCE code package is a set of neutronics tools based on 2D assembly calculations and 3D core calculations. It is made up of APOLLO2F, SMART and SQUALE and used to perform the nuclear design and loading pattern analysis for the reactors on operation or under construction of China Guangdong Nuclear Power Group. The purpose of paper is to briefly present the physical and numerical models used in each computation codes of the SCIENCE code pack age, including the description of the general structure of the code package, the coupling relationship of APOLLO2-F transport lattice code and SMART core nodal code, and the SQUALE code used for processing the core maps. (authors)
DEFF Research Database (Denmark)
Ejsing-Duun, Stine; Hansbøl, Mikala
Denne rapport rummer evaluering og dokumentation af Coding Class projektet1. Coding Class projektet blev igangsat i skoleåret 2016/2017 af IT-Branchen i samarbejde med en række medlemsvirksomheder, Københavns kommune, Vejle Kommune, Styrelsen for IT- og Læring (STIL) og den frivillige forening......, design tænkning og design-pædagogik, Stine Ejsing-Duun fra Forskningslab: It og Læringsdesign (ILD-LAB) ved Institut for kommunikation og psykologi, Aalborg Universitet i København. Vi har fulgt og gennemført evaluering og dokumentation af Coding Class projektet i perioden november 2016 til maj 2017....... Coding Class projektet er et pilotprojekt, hvor en række skoler i København og Vejle kommuner har igangsat undervisningsaktiviteter med fokus på kodning og programmering i skolen. Evalueringen og dokumentationen af projektet omfatter kvalitative nedslag i udvalgte undervisningsinterventioner i efteråret...
Directory of Open Access Journals (Sweden)
Anthony McCosker
2014-03-01
Full Text Available As well as introducing the Coding Labour section, the authors explore the diffusion of code across the material contexts of everyday life, through the objects and tools of mediation, the systems and practices of cultural production and organisational management, and in the material conditions of labour. Taking code beyond computation and software, their specific focus is on the increasingly familiar connections between code and labour with a focus on the codification and modulation of affect through technologies and practices of management within the contemporary work organisation. In the grey literature of spreadsheets, minutes, workload models, email and the like they identify a violence of forms through which workplace affect, in its constant flux of crisis and ‘prodromal’ modes, is regulated and governed.
International Nuclear Information System (INIS)
Buechse, H.; Langowski, A.; Lein, M.; Nagel, R.; Schmidt, H.; Stammel, M.
1995-01-01
The report gives the results of investigations on the validation of computer codes used to prove nuclear safety during transport and storage of spent VVER - fuel of NPP Greifswald and Rheinsberg. Characteristics of typical spent fuel (nuclide concentration, neutron source strength, gamma spectrum, decay heat) - calculated with several codes - and dose rates (e.g. in the surrounding of a loaded spent fuel cask) - based on the different source terms - are presented. Differences and their possible reasons are discussed. The results show that despite the differences in the source terms all relevant health physics requirements are met for all cases of source term. The validation of the criticality code OMEGA was established by calculation of appr. 200 critical experiments of LWR fuel, including VVER fuel rod arrangements. The mean error of the effective multiplication factor k eff is -0,01 compared to the experiment for this area of applicability. Thus, the OMEGA error of 2% assumed in earlier works has turned out to be sufficiently conservative. (orig.) [de
Energy Technology Data Exchange (ETDEWEB)
Ravishankar, C., Hughes Network Systems, Germantown, MD
1998-05-08
Speech is the predominant means of communication between human beings and since the invention of the telephone by Alexander Graham Bell in 1876, speech services have remained to be the core service in almost all telecommunication systems. Original analog methods of telephony had the disadvantage of speech signal getting corrupted by noise, cross-talk and distortion Long haul transmissions which use repeaters to compensate for the loss in signal strength on transmission links also increase the associated noise and distortion. On the other hand digital transmission is relatively immune to noise, cross-talk and distortion primarily because of the capability to faithfully regenerate digital signal at each repeater purely based on a binary decision. Hence end-to-end performance of the digital link essentially becomes independent of the length and operating frequency bands of the link Hence from a transmission point of view digital transmission has been the preferred approach due to its higher immunity to noise. The need to carry digital speech became extremely important from a service provision point of view as well. Modem requirements have introduced the need for robust, flexible and secure services that can carry a multitude of signal types (such as voice, data and video) without a fundamental change in infrastructure. Such a requirement could not have been easily met without the advent of digital transmission systems, thereby requiring speech to be coded digitally. The term Speech Coding is often referred to techniques that represent or code speech signals either directly as a waveform or as a set of parameters by analyzing the speech signal. In either case, the codes are transmitted to the distant end where speech is reconstructed or synthesized using the received set of codes. A more generic term that is applicable to these techniques that is often interchangeably used with speech coding is the term voice coding. This term is more generic in the sense that the
DEFF Research Database (Denmark)
Cox, Geoff
Speaking Code begins by invoking the “Hello World” convention used by programmers when learning a new language, helping to establish the interplay of text and code that runs through the book. Interweaving the voice of critical writing from the humanities with the tradition of computing and softwa...... expression in the public realm. The book’s line of argument defends language against its invasion by economics, arguing that speech continues to underscore the human condition, however paradoxical this may seem in an era of pervasive computing....
Kovtun, Oleg; Sakrikar, Dhananjay; Tomlinson, Ian D; Chang, Jerry C; Arzeta-Ferrer, Xochitl; Blakely, Randy D; Rosenthal, Sandra J
2015-04-15
The presynaptic, cocaine- and amphetamine-sensitive dopamine (DA) transporter (DAT, SLC6A3) controls the intensity and duration of synaptic dopamine signals by rapid clearance of DA back into presynaptic nerve terminals. Abnormalities in DAT-mediated DA clearance have been linked to a variety of neuropsychiatric disorders, including addiction, autism, and attention deficit/hyperactivity disorder (ADHD). Membrane trafficking of DAT appears to be an important, albeit incompletely understood, post-translational regulatory mechanism; its dysregulation has been recently proposed as a potential risk determinant of these disorders. In this study, we demonstrate a link between an ADHD-associated DAT mutation (Arg615Cys, R615C) and variation on DAT transporter cell surface dynamics, a combination only previously studied with ensemble biochemical and optical approaches that featured limited spatiotemporal resolution. Here, we utilize high-affinity, DAT-specific antagonist-conjugated quantum dot (QD) probes to establish the dynamic mobility of wild-type and mutant DATs at the plasma membrane of living cells. Single DAT-QD complex trajectory analysis revealed that the DAT 615C variant exhibited increased membrane mobility relative to DAT 615R, with diffusion rates comparable to those observed after lipid raft disruption. This phenomenon was accompanied by a loss of transporter mobilization triggered by amphetamine, a common component of ADHD medications. Together, our data provides the first dynamic imaging of single DAT proteins, providing new insights into the relationship between surface dynamics and trafficking of both wild-type and disease-associated transporters. Our approach should be generalizable to future studies that explore the possibilities of perturbed surface DAT dynamics that may arise as a consequence of genetic alterations, regulatory changes, and drug use that contribute to the etiology or treatment of neuropsychiatric disorders.
International Nuclear Information System (INIS)
Pulkkanen, V.-M.; Nordman, H.
2010-03-01
Traditional radionuclide transport models overestimate significantly some phenomena, or completely ignore them. This motivates the development of new more precise models. As a result, this work is a description of commissioning of a new KBS-3V near-field radionuclide transport model, which has been done with a commercial software called GoldSim. According to earlier models, GoldSim model uses rz coordinates, but the solubilities of radionuclides have been treated more precisely. To begin with, the physical phenomena concerning near-field transport have been introduced according to GoldSim way of thinking. Also, the computational methods of GoldSim have been introduced and compared to methods used earlier. The actual verification of GoldSim model has been carried out by comparing the GoldSim results from simple cases to the corresponding results obtained with REPCOM, a software developed by VTT and used in several safety assessments. The results agree well. Finally, a few complicated cases were studied. In these cases, the REPCOM's limitations in handling of some phenomena become evident. The differences in the results are caused especially by the extension of the solubility limit to the whole computational domain, and the element-wise treatment of the solubilities which was used instead of nuclide-wise treatment. This work has been carried out as a special assignment to the former laboratory of Advanced Energy Systems in Helsinki University of Technology. The work was done at VTT. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Pulkkanen, V.-M.; Nordman, H. (VTT Technical Research Centre, Espoo (Finland))
2010-03-15
Traditional radionuclide transport models overestimate significantly some phenomena, or completely ignore them. This motivates the development of new more precise models. As a result, this work is a description of commissioning of a new KBS-3V near-field radionuclide transport model, which has been done with a commercial software called GoldSim. According to earlier models, GoldSim model uses rz coordinates, but the solubilities of radionuclides have been treated more precisely. To begin with, the physical phenomena concerning near-field transport have been introduced according to GoldSim way of thinking. Also, the computational methods of GoldSim have been introduced and compared to methods used earlier. The actual verification of GoldSim model has been carried out by comparing the GoldSim results from simple cases to the corresponding results obtained with REPCOM, a software developed by VTT and used in several safety assessments. The results agree well. Finally, a few complicated cases were studied. In these cases, the REPCOM's limitations in handling of some phenomena become evident. The differences in the results are caused especially by the extension of the solubility limit to the whole computational domain, and the element-wise treatment of the solubilities which was used instead of nuclide-wise treatment. This work has been carried out as a special assignment to the former laboratory of Advanced Energy Systems in Helsinki University of Technology. The work was done at VTT. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Delbecq, J.M
1999-07-01
The Aster code is a 2D or 3D finite-element calculation code for structures developed by the R and D direction of Electricite de France (EdF). This dossier presents a complete overview of the characteristics and uses of the Aster code: introduction of version 4; the context of Aster (organisation of the code development, versions, systems and interfaces, development tools, quality assurance, independent validation); static mechanics (linear thermo-elasticity, Euler buckling, cables, Zarka-Casier method); non-linear mechanics (materials behaviour, big deformations, specific loads, unloading and loss of load proportionality indicators, global algorithm, contact and friction); rupture mechanics (G energy restitution level, restitution level in thermo-elasto-plasticity, 3D local energy restitution level, KI and KII stress intensity factors, calculation of limit loads for structures), specific treatments (fatigue, rupture, wear, error estimation); meshes and models (mesh generation, modeling, loads and boundary conditions, links between different modeling processes, resolution of linear systems, display of results etc..); vibration mechanics (modal and harmonic analysis, dynamics with shocks, direct transient dynamics, seismic analysis and aleatory dynamics, non-linear dynamics, dynamical sub-structuring); fluid-structure interactions (internal acoustics, mass, rigidity and damping); linear and non-linear thermal analysis; steels and metal industry (structure transformations); coupled problems (internal chaining, internal thermo-hydro-mechanical coupling, chaining with other codes); products and services. (J.S.)
Energy Technology Data Exchange (ETDEWEB)
Lindemuth, I.R.
1979-02-28
This report describes ANIMAL, a two-dimensional Eulerian magnetohydrodynamic computer code. ANIMAL's physical model also appears. Formulated are temporal and spatial finite-difference equations in a manner that facilitates implementation of the algorithm. Outlined are the functions of the algorithm's FORTRAN subroutines and variables.
Indian Academy of Sciences (India)
Home; Journals; Resonance – Journal of Science Education; Volume 15; Issue 7. Network Coding. K V Rashmi Nihar B Shah P Vijay Kumar. General Article Volume 15 Issue 7 July 2010 pp 604-621. Fulltext. Click here to view fulltext PDF. Permanent link: https://www.ias.ac.in/article/fulltext/reso/015/07/0604-0621 ...
International Nuclear Information System (INIS)
Lindemuth, I.R.
1979-01-01
This report describes ANIMAL, a two-dimensional Eulerian magnetohydrodynamic computer code. ANIMAL's physical model also appears. Formulated are temporal and spatial finite-difference equations in a manner that facilitates implementation of the algorithm. Outlined are the functions of the algorithm's FORTRAN subroutines and variables
Indian Academy of Sciences (India)
Codes and Channels. A noisy communication channel is illustrated in Fig- ... nication channel. Suppose we want to transmit a message over the unreliable communication channel so that even if the channel corrupts some of the bits we are able to recover ..... is d-regular, meaning thereby that every vertex has de- gree d.
Indian Academy of Sciences (India)
Home; Journals; Resonance – Journal of Science Education; Volume 10; Issue 1. Expander Codes - The Sipser–Spielman Construction. Priti Shankar. General Article Volume 10 ... Author Affiliations. Priti Shankar1. Department of Computer Science and Automation, Indian Institute of Science Bangalore 560 012, India.
Indian Academy of Sciences (India)
Network coding is a technique to increase the amount of information °ow in a network by mak- ing the key observation that information °ow is fundamentally different from commodity °ow. Whereas, under traditional methods of opera- tion of data networks, intermediate nodes are restricted to simply forwarding their incoming.
International Nuclear Information System (INIS)
2000-01-01
This literature survey was undertaken jointly by the International Maritime Organization (IMO) and the International Atomic Energy Agency (IAEA) as a step in addressing the subject of environmental impact of accidents involving materials subject to the IMO's Code for the Safe Carriage of Irradiated Nuclear Fuel, Plutonium and High-Level Radioactive Wastes in Flasks on Board Ships, also known as the INF Code. The results of the survey are provided in two volumes: the first one containing the description of the search and search results with the list of generated publication titles, and the second volume containing the abstracts of those publications deemed relevant for the purposes of the literature survey. Literature published between 1980 and mid-1999 was reviewed by two independent consultants who generated publication titles by performing searches of appropriate databases, and selected the abstracts of relevant publications for inclusion in this survey. The IAEA operates INIS, the world's leading computerised bibliographical information system on the peaceful uses of nuclear energy. The acronym INIS stands for International Nuclear Information System. INIS Members are responsible for determining the relevant nuclear literature produced within their borders or organizational confines, and then preparing the associated input in accordance with INIS rules. INIS records are included in other major databases such as the Energy, Science and Technology database of the DIALOG service. Because it is the INIS Members, rather than the IAEA Secretariat, who are responsible for its contents, it was considered appropriate that INIS be the primary source of information for this literature review. Selected unpublished reports were also reviewed, e.g. Draft Proceedings of the Special Consultative Meeting of Entities involved in the maritime transport of materials covered by the INF Code (SCM 5), March 1996. Many of the formal papers at SCM 5 were included in the literature
Energy Technology Data Exchange (ETDEWEB)
Chepe P, M. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico); Xolocostli M, J. V.; Gomez T, A. M. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E., E-mail: liaison.web@gmail.com [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, 07738 Ciudad de Mexico (Mexico)
2015-09-15
The deterministic transport codes for analysis of nuclear reactors have been used for several years already, these codes have evolved in terms of the methodology used and the degree of accuracy, because at the present time has more computer power. In this paper, the transport code used considers the classical technique of multi-group for discretization energy, for space discretization uses the nodal methods, while for the angular discretization the discrete ordinates method is used; so that presents the development and implementation of a set of numerical quadratures of SQ{sub N} type symmetrical with the same weight for each angular direction and these are compared with the quadratures of EQ{sub N} type. The two sets of numerical quadratures were implemented in the program AZTRAN to a problem with isotropic medium in XYZ geometry, in steady state using the nodal method RTN-0 (Raviart-Thomas-Nedelec). The analyzed results correspond to the effective multiplication factor k{sub eff} and neutron angular flux with approximations from S{sub 4} to S{sub 16}. (Author)
International Nuclear Information System (INIS)
Altomare, S.; Minton, G.
1975-02-01
PANDA is a new two-group one-dimensional (slab/cylinder) neutron diffusion code designed to replace and extend the FAB series. PANDA allows for the nonlinear effects of xenon, enthalpy and Doppler. Fuel depletion is allowed. PANDA has a completely general search facility which will seek criticality, maximize reactivity, or minimize peaking. Any single parameter may be varied in a search. PANDA is written in FORTRAN IV, and as such is nearly machine independent. However, PANDA has been written with the present limitations of the Westinghouse CDC-6600 system in mind. Most computation loops are very short, and the code is less than half the useful 6600 memory size so that two jobs can reside in the core at once. (auth)
Energy Technology Data Exchange (ETDEWEB)
Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.; Stuart, D.S.; Thompson, S.L. [Sandia National Labs., Albuquerque, NM (United States); Hodge, S.A.; Hyman, C.R.; Sanders, R.L. [Oak Ridge National Lab., TN (United States)
1995-03-01
MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.
International Nuclear Information System (INIS)
Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.; Stuart, D.S.; Thompson, S.L.; Hodge, S.A.; Hyman, C.R.; Sanders, R.L.
1995-03-01
MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR's phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package
Toride, N.; Matsuoka, K.
2017-12-01
In order to predict the fate and transport of nitrogen in a reduced paddy field as a result of decomposition of organic matter, we implemented within the PHREEQC program a modified coupled carbon and nitrogen cycling model based on the LEACHM code. SOM decay processes from organic carbon (Org-C) to biomass carbon (Bio-C), humus carbon (Hum-C), and carbon dioxide (CO2) were described using first-order kinetics. Bio-C was recycled into the organic pool. When oxygen was available in an aerobic condition, O2 was used to produce CO2 as an electron accepter. When O2 availability is low, other electron acceptors such as NO3-, Mn4+, Fe3+, SO42-, were used depending on the redox potential. Decomposition of Org-N was related to the carbon cycle using the C/N ratio. Mineralization and immobilization were determined based on available NH4-N and the nitrogen demand for the formation of biomass and humus. Although nitrification was independently described with the first-order decay process, denitrification was linked with the SOM decay since NO3- was an electron accepter for the CO2 production. Proton reactions were coupled with the nitrification from NH4+ to NO3-, and the ammonium generation from NH3 to NH4+. Furthermore, cation and anion exchange reactions were included with the permanent negative charges and the pH dependent variable charges. The carbon and nitrogen cycling model described with PHREEQC was linked with HYDRUS-1D using the HP1 code. Various nitrogen and carbon transport scenarios were demonstrated for the application of organic matter to a saturated paddy soil.
Sommariva, C.; Nardon, E.; Beyer, P.; Hoelzl, M.; Huijsmans, G. T. A.; van Vugt, D.; Contributors, JET
2018-01-01
In order to contribute to the understanding of runaway electron generation mechanisms during tokamak disruptions, a test particle tracker is introduced in the JOREK 3D non-linear MHD code, able to compute both full and guiding center relativistic orbits. Tests of the module show good conservation of the invariants of motion and consistency between full orbit and guiding center solutions. A first application is presented where test electron confinement properties are investigated in a massive gas injection-triggered disruption simulation in JET-like geometry. It is found that electron populations initialised before the thermal quench (TQ) are typically not fully deconfined in spite of the global stochasticity of the magnetic field during the TQ. The fraction of ‘survivors’ decreases from a few tens down to a few tenths of percent as the electron energy varies from 1 keV to 10 MeV. The underlying mechanism for electron ‘survival’ is the prompt reformation of closed magnetic surfaces at the plasma core and, to a smaller extent, the subsequent reappearance of a magnetic surface at the edge. It is also found that electrons are less deconfined at 10 MeV than at 1 MeV, which appears consistent with a phase averaging effect due to orbit shifts at high energy.
Montgomery, O. L.
1977-01-01
Procedures developed for digitizing the transportation arteries, airports, and dock facilities of Alabama and placing them in a computerized format compatible with the Alabama Resource Information System are described. The time required to digitize by the following methods: (a) manual, (b) Telereadex 29 with film reading and digitizing system, and (c) digitizing tablets was evaluated. A method for digitizing and storing information from the U. T. M. grid cell base which was compatible with the system was developed and tested. The highways, navigable waterways, railroads, airports, and docks in the study area were digitized and the data stored. The manual method of digitizing was shown to be best for small amounts of data, while the graphic input from the digitizing tablets would be the best approach for entering the large amounts of data required for an entire state.
International Nuclear Information System (INIS)
Pellacani, F.; Macian, R.; Chiva, S.; Pena, C.
2011-01-01
In this paper upward, isothermal and turbulent bubbly flow in tubes is numerically modeled by using ANSYS CFX 12.1 with the aim of creating a basis for the reliable simulation of the flow along a vertical channel in a nuclear reactor as long term goal. Two approaches based on the mono-dispersed model and on the one-group Interfacial Area Transport Equation (IATE) model are used in order to maintain the computational effort as low as possible. This work represents the necessary step to implement a two-group interfacial area transport equation that will be able to dynamically represent the changes in interfacial structure in the transition region from bubbly to slug flow. The drag coefficient is calculated using the Grace model and the interfacial non-drag forces are also included. The Antal model is used for the calculation of the wall lubrication force coefficient. The lift force coefficient is obtained from the Tomiyama model. The turbulent dispersion force is taken into account and is modeled using the FAD (Favre averaged drag) approach, while the turbulence transfer is simulated with the Sato's model. The liquid velocity is in the range between 0.5 and 2 m/s and the average void fraction varies between 5 and 15%.The source and sink terms for break-up and coalescence needed for the calculation of the implemented Interfacial Area Density are those proposed by Yao and Morel. The model has been checked using experimental results by Mendez. Radial profile distributions of void fraction, interfacial area density and bubble mean diameter are shown at the axial position equivalent to z/D=56. The results obtained by the simulations have a good agreement with the experimental data but show also the need of a better study of the coalescence and breakup phenomena to develop more accurate interaction models. (author)
Energy Technology Data Exchange (ETDEWEB)
Beckert, C.
2007-12-19
Conventionally the data preparation of the neutron cross sections for reactor-core calculations pursues with 2D cell codes. Aim of this thesis was, to develop a 3D cell code, to study with this code 3D effects, and to evaluate the necessarity of a 3D data preparation of the neutron cross sections. For the calculation of the neutron transport the method of the first-collision probabilities, which are calculated with the ray-tracing method, was chosen. The mathematical algorithms were implemented in the 2D/3D cell code TransRay. For the geometry part of the program the geometry module of a Monte Carlo code was used.The ray tracing in 3D was parallelized because of the high computational time. The program TransRay was verified on 2D test problems. For a reference pressured-water reactor following 3D problems were studied: A partly immersed control rod and void (vacuum or steam) around a fuel rod as model of a steam void. All problems were for comparison calculated also with the programs HELIOS(2D) and MCNP(3D). The dependence of the multiplication factor and the averaged two-group cross section on the immersion depth of the control rod respectively of the height of the steam void were studied. The 3D-calculated two-group cross sections were compared with three conventional approximations: Linear interpolation, interpolation with flux weighting, and homogenization, At the 3D problem of the control rod it was shown that the interpolation with flux weighting is a good approximation. Therefore here a 3D data preparation is not necessary. At the test case of the single control rod, which is surrounded by the void, the three approximation for the two-group cross sections were proved as unsufficient. Therefore a 3D data preparation is necessary. The single fuel-rod cell with void can be considered as the limiting case of a reactor, in which a phase interface has been formed. [German] Standardmaessig erfolgt die Datenaufbereitung der Neutronenwirkungsquerschnitte fuer
From concatenated codes to graph codes
DEFF Research Database (Denmark)
Justesen, Jørn; Høholdt, Tom
2004-01-01
We consider codes based on simple bipartite expander graphs. These codes may be seen as the first step leading from product type concatenated codes to more complex graph codes. We emphasize constructions of specific codes of realistic lengths, and study the details of decoding by message passing...
Energy Technology Data Exchange (ETDEWEB)
Jimenez P, D. A.
2014-07-01
The accidents in Unit 2 of the Three Mile Island Nuclear Power Plant (NPP) in the United States (March 28{sup th}, 1979), the one in Unit 4 of the NPP Chernobyl in Ukraine (April 26{sup th}, 1986) and the explosions in some units of Fukushima NPP in Japan (March 11{sup th}, 2011) boosted the investigations on severe accidents with core damage and, in particular, the threat to the ultimate barrier by an eventual explosion from uncontrolled Hydrogen combustion within the containment was considered of particular relevance. Research programs for analyzing Hydrogen behavior and control during this kind of accidents were early initiated by research and regulatory bodies. Assessment on Hydrogen behavior once it has been postulated to be released on the containment system can be divided into two phases, in the first one, transport and the concentrations of the gas mixtures and steam in each volume or area comprised between the structures of the containment are calculated, in the second one, the propagation of the detonation of the Hydrogen is calculated if there are the conditions to occur. Currently, there are computer programs that can be used in one, or both stages of computation, and they are based on one of the two solution methods in current use, one of them are integrated codes (e.g. MELCOR), which consists in assuming the containment as a network composed of hydraulic tanks or nodes on which the balance equations of mass and energy have to be solved, the network is connected by ducts or connections where the momentum balance equation arise. This methodology relies on the use of semi-empirical relationships and the criteria used to define a geometric pattern, are subjective. The second method, which is having relevance due to the large computing power of modern computers, is the numerical solution of the three-dimensional Navier-Stokes equations in complex geometries. This method of solution is known as Computational Fluid Dynamics (CFD), and offers the advantage of
Codes & standards research, development & demonstration Roadmap
Energy Technology Data Exchange (ETDEWEB)
None, None
2008-07-22
This Roadmap is a guide to the Research, Development & Demonstration activities that will provide data required for SDOs to develop performance-based codes and standards for a commercial hydrogen fueled transportation sector in the U.S.
Automatic coding method of the ACR Code
International Nuclear Information System (INIS)
Park, Kwi Ae; Ihm, Jong Sool; Ahn, Woo Hyun; Baik, Seung Kook; Choi, Han Yong; Kim, Bong Gi
1993-01-01
The authors developed a computer program for automatic coding of ACR(American College of Radiology) code. The automatic coding of the ACR code is essential for computerization of the data in the department of radiology. This program was written in foxbase language and has been used for automatic coding of diagnosis in the Department of Radiology, Wallace Memorial Baptist since May 1992. The ACR dictionary files consisted of 11 files, one for the organ code and the others for the pathology code. The organ code was obtained by typing organ name or code number itself among the upper and lower level codes of the selected one that were simultaneous displayed on the screen. According to the first number of the selected organ code, the corresponding pathology code file was chosen automatically. By the similar fashion of organ code selection, the proper pathologic dode was obtained. An example of obtained ACR code is '131.3661'. This procedure was reproducible regardless of the number of fields of data. Because this program was written in 'User's Defined Function' from, decoding of the stored ACR code was achieved by this same program and incorporation of this program into program in to another data processing was possible. This program had merits of simple operation, accurate and detail coding, and easy adjustment for another program. Therefore, this program can be used for automation of routine work in the department of radiology
Fundamentals of convolutional coding
Johannesson, Rolf
2015-01-01
Fundamentals of Convolutional Coding, Second Edition, regarded as a bible of convolutional coding brings you a clear and comprehensive discussion of the basic principles of this field * Two new chapters on low-density parity-check (LDPC) convolutional codes and iterative coding * Viterbi, BCJR, BEAST, list, and sequential decoding of convolutional codes * Distance properties of convolutional codes * Includes a downloadable solutions manual
Energy Technology Data Exchange (ETDEWEB)
Nourtier-Mazauric, E.
2003-03-15
This thesis presents a thermodynamic and kinetic model of interactions between a fluid and ideal solid solutions represented by several end-members. The reaction between a solid solution and the aqueous solution results from the competition between the stoichiometric dissolution of the initial solid solution and the co-precipitation of the least soluble solid solution in the fluid at considered time. This model was implemented in ARCHIMEDE, a computer code of reactive transport in porous media, then applied to various examples. In the case of binary solid solutions, a graphical method allowed to determine the compositions of the precipitating solid solutions, with the aid of the end-member chemical potentials. The obtained program could be used to notably model the diagenesis of clayey or carbonated oil reservoirs, or the ground pollutant dispersion. (author)
SNSPH: A Parallel 3-D Smoothed Particle Radiation Hydrodynamics Code
Fryer, C. L.; Rockefeller, G.; Warren, M. S.
2005-01-01
We provide a description of the SNSPH code--a parallel 3-dimensional radiation hydrodynamics code implementing treecode gravity, smooth particle hydrodynamics, and flux-limited diffusion transport schemes. We provide descriptions of the physics and parallelization techniques for this code. We present performance results on a suite of code tests (both standard and new), showing the versatility of such a code, but focusing on what we believe are important aspects of modeling core-collapse super...
CHARTB multigroup transport package
International Nuclear Information System (INIS)
Baker, L.
1979-03-01
The physics and numerical implementation of the radiation transport routine used in the CHARTB MHD code are discussed. It is a one-dimensional (Cartesian, cylindrical, and spherical symmetry), multigroup,, diffusion approximation. Tests and applications will be discussed as well
Structure and Operation of the ITS Code System
Halbleib, J.
The TIGER series of time-independent coupled electron-photon Monte Carlo transport codes is a group of multimaterial and multidimensional codes designed to provide a state-of-the-art description of the production and transport of the electron-photon cascade by combining microscopic photon transport with a macroscopic random walk1 for electron transport. Major contributors to its evolution are listed in Table 10.1.
New Mexico Univ., Albuquerque. American Indian Law Center.
The Model Children's Code was developed to provide a legally correct model code that American Indian tribes can use to enact children's codes that fulfill their legal, cultural and economic needs. Code sections cover the court system, jurisdiction, juvenile offender procedures, minor-in-need-of-care, and termination. Almost every Code section is…
DEFF Research Database (Denmark)
Høholdt, Tom; Beelen, Peter; Ghorpade, Sudhir Ramakant
2010-01-01
We consider a new class of linear codes, called affine Grassmann codes. These can be viewed as a variant of generalized Reed-Muller codes and are closely related to Grassmann codes.We determine the length, dimension, and the minimum distance of any affine Grassmann code. Moreover, we show that af...
Computation of the bounce-average code
International Nuclear Information System (INIS)
Cutler, T.A.; Pearlstein, L.D.; Rensink, M.E.
1977-01-01
The bounce-average computer code simulates the two-dimensional velocity transport of ions in a mirror machine. The code evaluates and bounce-averages the collision operator and sources along the field line. A self-consistent equilibrium magnetic field is also computed using the long-thin approximation. Optionally included are terms that maintain μ, J invariance as the magnetic field changes in time. The assumptions and analysis that form the foundation of the bounce-average code are described. When references can be cited, the required results are merely stated and explained briefly. A listing of the code is appended
49 CFR 236.727 - Circuit, track; coded.
2010-10-01
... 49 Transportation 4 2010-10-01 2010-10-01 false Circuit, track; coded. 236.727 Section 236.727 Transportation Other Regulations Relating to Transportation (Continued) FEDERAL RAILROAD ADMINISTRATION, DEPARTMENT OF TRANSPORTATION RULES, STANDARDS, AND INSTRUCTIONS GOVERNING THE INSTALLATION, INSPECTION...
Code system BCG for gamma-ray skyshine calculation
International Nuclear Information System (INIS)
Ryufuku, Hiroshi; Numakunai, Takao; Miyasaka, Shun-ichi; Minami, Kazuyoshi.
1979-03-01
A code system BCG has been developed for calculating conveniently and efficiently gamma-ray skyshine doses using the transport calculation codes ANISN and DOT and the point-kernel calculation codes G-33 and SPAN. To simplify the input forms to the system, the forms for these codes are unified, twelve geometric patterns are introduced to give material regions, and standard data are available as a library. To treat complex arrangements of source and shield, it is further possible to use successively the code such that the results from one code may be used as input data to the same or other code. (author)
Generalized concatenated quantum codes
International Nuclear Information System (INIS)
Grassl, Markus; Shor, Peter; Smith, Graeme; Smolin, John; Zeng Bei
2009-01-01
We discuss the concept of generalized concatenated quantum codes. This generalized concatenation method provides a systematical way for constructing good quantum codes, both stabilizer codes and nonadditive codes. Using this method, we construct families of single-error-correcting nonadditive quantum codes, in both binary and nonbinary cases, which not only outperform any stabilizer codes for finite block length but also asymptotically meet the quantum Hamming bound for large block length.
DEFF Research Database (Denmark)
Sørensen, Jesper Hemming; Koike-Akino, Toshiaki; Orlik, Philip
2012-01-01
This paper proposes a concept called rateless feedback coding. We redesign the existing LT and Raptor codes, by introducing new degree distributions for the case when a few feedback opportunities are available. We show that incorporating feedback to LT codes can significantly decrease both...... the coding overhead and the encoding/decoding complexity. Moreover, we show that, at the price of a slight increase in the coding overhead, linear complexity is achieved with Raptor feedback coding....
Shavers, M. R.; Atwell, W.; Cucinotta, F. A.; Badhwar, G. D. (Technical Monitor)
1999-01-01
cell killing from GCR, including patterns of cell killing from single particle tracks. can provide useful information on expected differences between proton and HZE tracks and clinical experiences with photon irradiation. To model effects on cells in the brain, it is important that transport models accurately describe changes in the GCR due to interactions in the cranium and proximate tissues. We describe calculations of the attenuated GCR particle fluxes at three dose-points in the brain and associated patterns of cell killing using biophysical models. The effects of the brain self-shielding and bone-tissue interface of the skull in modulating the GCR environment are considered. For each brain dose-point, the mass distribution in the surrounding 4(pi) solid angle is characterized using the CAM model to trace 512 rays. The CAM model describes the self-shielding by converting the tissue distribution to mass-equivalent aluminum, and nominal values of spacecraft shielding is considered. Particle transport is performed with the proton, neutron, and heavy-ion transport code HZETRN with the nuclear fragmentation model QMSFRG. The distribution of cells killed along the path of individual GCR ions is modeled using in vitro cell inactivation data for cells with varying sensitivity. Monte Carlo simulations of arrays of inactivated cells are considered for protons and heavy ions and used to describe the absolute number of cell killing events of various magnitude in the brain from the GCR. Included are simulations of positions of inactivated cells from stopping heavy ions and nuclear stars produced by high-energy ions most importantly, protons and neutrons.
Abraham, Nikhil
2015-01-01
Hands-on exercises help you learn to code like a pro No coding experience is required for Coding For Dummies,your one-stop guide to building a foundation of knowledge inwriting computer code for web, application, and softwaredevelopment. It doesn't matter if you've dabbled in coding or neverwritten a line of code, this book guides you through the basics.Using foundational web development languages like HTML, CSS, andJavaScript, it explains in plain English how coding works and whyit's needed. Online exercises developed by Codecademy, a leading online codetraining site, help hone coding skill
Gao, Wen
2015-01-01
This comprehensive and accessible text/reference presents an overview of the state of the art in video coding technology. Specifically, the book introduces the tools of the AVS2 standard, describing how AVS2 can help to achieve a significant improvement in coding efficiency for future video networks and applications by incorporating smarter coding tools such as scene video coding. Topics and features: introduces the basic concepts in video coding, and presents a short history of video coding technology and standards; reviews the coding framework, main coding tools, and syntax structure of AV
Computer Code for Nanostructure Simulation
Filikhin, Igor; Vlahovic, Branislav
2009-01-01
Due to their small size, nanostructures can have stress and thermal gradients that are larger than any macroscopic analogue. These gradients can lead to specific regions that are susceptible to failure via processes such as plastic deformation by dislocation emission, chemical debonding, and interfacial alloying. A program has been developed that rigorously simulates and predicts optoelectronic properties of nanostructures of virtually any geometrical complexity and material composition. It can be used in simulations of energy level structure, wave functions, density of states of spatially configured phonon-coupled electrons, excitons in quantum dots, quantum rings, quantum ring complexes, and more. The code can be used to calculate stress distributions and thermal transport properties for a variety of nanostructures and interfaces, transport and scattering at nanoscale interfaces and surfaces under various stress states, and alloy compositional gradients. The code allows users to perform modeling of charge transport processes through quantum-dot (QD) arrays as functions of inter-dot distance, array order versus disorder, QD orientation, shape, size, and chemical composition for applications in photovoltaics and physical properties of QD-based biochemical sensors. The code can be used to study the hot exciton formation/relation dynamics in arrays of QDs of different shapes and sizes at different temperatures. It also can be used to understand the relation among the deposition parameters and inherent stresses, strain deformation, heat flow, and failure of nanostructures.
Peripheral Codes in ASTRA for the TJ-II
International Nuclear Information System (INIS)
Lopez-Bruna, D.; Reynolds, J. M.; Cappa, A.; Martinell, J.; Garcia, J.; Gutierrez-Tapia, C.
2010-01-01
The study of data from the TJ-II device is often done with transport calculations based on the ASTRA transport system. However, complicated independent codes are used to obtain fundamental ingredients in these calculations, such as the particle and/or energy sources. These codes are accessible from ASTRA through the procedures explained in this report. (Author) 37 refs.
Computer code ANISN multiplying media and shielding calculation II. Code description (input/output)
International Nuclear Information System (INIS)
Maiorino, J.R.
1990-01-01
The user manual of the ANISN computer code describing input and output subroutines is presented. ANISN code was developed to solve one-dimensional transport equation for neutron or gamma rays in slab, sphere or cylinder geometry with general anisotropic scattering. The solution technique is the discrete ordinate method. (M.C.K.)
Contaminant transport at a waste residue deposit
DEFF Research Database (Denmark)
Engesgaard, Peter Knudegaard; Traberg, Rikke
1996-01-01
Contaminant transport in an aquifer at an incinerator waste residue deposit in Denmark is simulated. A two-dimensional, geochemical transport code is developed for this purpose and tested by comparison to results from another code, The code is applied to a column experiment and to the field site...
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)
REVISED STREAM CODE AND WASP5 BENCHMARK
International Nuclear Information System (INIS)
Chen, K
2005-01-01
STREAM is an emergency response code that predicts downstream pollutant concentrations for releases from the SRS area to the Savannah River. The STREAM code uses an algebraic equation to approximate the solution of the one dimensional advective transport differential equation. This approach generates spurious oscillations in the concentration profile when modeling long duration releases. To improve the capability of the STREAM code to model long-term releases, its calculation module was replaced by the WASP5 code. WASP5 is a US EPA water quality analysis program that simulates one-dimensional pollutant transport through surface water. Test cases were performed to compare the revised version of STREAM with the existing version. For continuous releases, results predicted by the revised STREAM code agree with physical expectations. The WASP5 code was benchmarked with the US EPA 1990 and 1991 dye tracer studies, in which the transport of the dye was measured from its release at the New Savannah Bluff Lock and Dam downstream to Savannah. The peak concentrations predicted by the WASP5 agreed with the measurements within ±20.0%. The transport times of the dye concentration peak predicted by the WASP5 agreed with the measurements within ±3.6%. These benchmarking results demonstrate that STREAM should be capable of accurately modeling releases from SRS outfalls
Contaminant transport at a waste residue deposit
DEFF Research Database (Denmark)
Engesgaard, Peter Knudegaard; Traberg, Rikke
1996-01-01
Contaminant transport in an aquifer at an incinerator waste residue deposit in Denmark is simulated. A two-dimensional, geochemical transport code is developed for this purpose and tested by comparison to results from another code, The code is applied to a column experiment and to the field site...... wells. Contamination was first controlled by transport, and later by transport and ion exchange. In both the column and field site simulations the code is used to identify the controlling transport processes, physical or geochemical (ion exchange and mineral precipitation), and to estimate the Involved...
Locally orderless registration code
DEFF Research Database (Denmark)
2012-01-01
This is code for the TPAMI paper "Locally Orderless Registration". The code requires intel threadding building blocks installed and is provided for 64 bit on mac, linux and windows.......This is code for the TPAMI paper "Locally Orderless Registration". The code requires intel threadding building blocks installed and is provided for 64 bit on mac, linux and windows....
Indian Academy of Sciences (India)
sound quality is, in essence, obtained by accurate waveform coding and decoding of the audio signals. In addition, the coded audio information is protected against disc errors by the use of a Cross Interleaved Reed-Solomon Code (CIRC). Reed-. Solomon codes were discovered by Irving Reed and Gus Solomon in 1960.
Adamson , Brian; Adjih , Cédric; Bilbao , Josu; Firoiu , Victor; Fitzek , Frank; Samah , Ghanem ,; Lochin , Emmanuel; Masucci , Antonia; Montpetit , Marie-Jose; Pedersen , Morten V.; Peralta , Goiuri; Roca , Vincent; Paresh , Saxena; Sivakumar , Senthil
2017-01-01
Internet Research Task Force - Working document of the Network Coding Research Group (NWCRG), draft-irtf-nwcrg-network-coding-taxonomy-05 (work in progress), https://datatracker.ietf.org/doc/draft-irtf-nwcrg-network-coding-taxonomy/; This document summarizes a recommended terminology for Network Coding concepts and constructs. It provides a comprehensive set of terms with unique names in order to avoid ambiguities in future Network Coding IRTF and IETF documents. This document is intended to ...
Crompton, Helen; LaFrance, Jason; van 't Hooft, Mark
2012-01-01
A QR (quick-response) code is a two-dimensional scannable code, similar in function to a traditional bar code that one might find on a product at the supermarket. The main difference between the two is that, while a traditional bar code can hold a maximum of only 20 digits, a QR code can hold up to 7,089 characters, so it can contain much more…
International Nuclear Information System (INIS)
Lenain, Roland
2015-01-01
This thesis is devoted to the implementation of a domain decomposition method applied to the neutron transport equation. The objective of this work is to access high-fidelity deterministic solutions to properly handle heterogeneities located in nuclear reactor cores, for problems' size ranging from color-sets of assemblies to large reactor cores configurations in 2D and 3D. The innovative algorithm developed during the thesis intends to optimize the use of parallelism and memory. The approach also aims to minimize the influence of the parallel implementation on the performances. These goals match the needs of APOLLO3 project, developed at CEA and supported by EDF and AREVA, which must be a portable code (no optimization on a specific architecture) in order to achieve best estimate modeling with resources ranging from personal computer to compute cluster available for engineers analyses. The proposed algorithm is a Parallel Multigroup-Block Jacobi one. Each sub-domain is considered as a multi-group fixed-source problem with volume-sources (fission) and surface-sources (interface flux between the sub-domains). The multi-group problem is solved in each sub-domain and a single communication of the interface flux is required at each power iteration. The spectral radius of the resolution algorithm is made similar to the one of a classical resolution algorithm with a nonlinear diffusion acceleration method: the well-known Coarse Mesh Finite Difference. In this way an ideal scalability is achievable when the calculation is parallelized. The memory organization, taking advantage of shared memory parallelism, optimizes the resources by avoiding redundant copies of the data shared between the sub-domains. Distributed memory architectures are made available by a hybrid parallel method that combines both paradigms of shared memory parallelism and distributed memory parallelism. For large problems, these architectures provide a greater number of processors and the amount of
International Nuclear Information System (INIS)
Anwarul Islam, M.; Akramuzzaman, M.M.; Zakaria, G.A.
2012-01-01
Manufacturing of miniaturized high activity 192 Ir sources have been made a market preference in modern brachytherapy. The smaller dimensions of the sources are flexible for smaller diameter of the applicators and it is also suitable for interstitial implants. Presently, miniaturized 60 Co HDR sources have been made available with identical dimensions to those of 192 Ir sources. 60 Co sources have an advantage of longer half life while comparing with 192 Ir source. High dose rate brachytherapy sources with longer half life are logically pragmatic solution for developing country in economic point of view. This study is aimed to compare the TG-43U1 dosimetric parameters for new BEBIG 60 Co HDR and new microSelectron 192 Ir HDR sources. Dosimetric parameters are calculated using EGSnrc-based Monte Carlo simulation code accordance with the AAPM TG-43 formalism for microSelectron HDR 192 Ir v2 and new BEBIG 60 Co HDR sources. Air-kerma strength per unit source activity, calculated in dry air are 9.698x10 -8 ± 0.55% U Bq -1 and 3.039x10 -7 ± 0.41% U Bq -1 for the above mentioned two sources, respectively. The calculated dose rate constants per unit air-kerma strength in water medium are 1.116±0.12% cGy h -1 U -1 and 1.097±0.12% cGy h -1 U -1 , respectively, for the two sources. The values of radial dose function for distances up to 1 cm and more than 22 cm for BEBIG 60 Co HDR source are higher than that of other source. The anisotropic values are sharply increased to the longitudinal sides of the BEBIG 60 Co source and the rise is comparatively sharper than that of the other source. Tissue dependence of the absorbed dose has been investigated with vacuum phantom for breast, compact bone, blood, lung, thyroid, soft tissue, testis, and muscle. No significant variation is noted at 5 cm of radial distance in this regard while comparing the two sources except for lung tissues. The true dose rates are calculated with considering photon as well as electron transport using
Efficient Coding of Information: Huffman Coding -RE ...
Indian Academy of Sciences (India)
1. Introduction. Shannon's landmark paper 'A Mathematical Theory of. Communication' [1] laid the foundation for communica- ... coding theory, codes over graphs and iterative techniques, and informa- tion theory. .... An important consequence of independence is that if. {Xb X2 , . Xn} are independent random variables, each.
Turbo Codes Extended with Outer BCH Code
DEFF Research Database (Denmark)
Andersen, Jakob Dahl
1996-01-01
The "error floor" observed in several simulations with the turbo codes is verified by calculation of an upper bound to the bit error rate for the ensemble of all interleavers. Also an easy way to calculate the weight enumerator used in this bound is presented. An extended coding scheme is proposed...
EGS4 code system and its application
International Nuclear Information System (INIS)
Shin, Chang Ho; Kim, Jong Kyung
1998-01-01
The EGS4 code system is a powerful and user-friend software package permitting state-of-the-art Monte Carlo solution of time-independent coupled electron/photon transport problems, with or without presence of macroscopic electric and magnetic fields. The EGS4 code system consists of EGS4, PEGS4, and USER code. The EGS4 code is designed to simulate electromagnetic cascades in various geometries and at energies up to a few thousand GeV and down to cut-off kinetic energies of 10 and 1 keV for electrons and photons, respectively. The radiation transport of electrons or photons can be simulated in any elements, compound, or mixture. The PEGS4 code, data preparation package, creates data to be used by the EGS4 code, using cross section tables for elements 1 through 100. USER code should be written. This consists of a MAIN program and the subroutines HOWFAR and AUSGAB, the latter two determining the geometry and output (scoring), respectively. The EGS4 code system has been written in a MORTRAN language, extended FORTRAN language. The EGS4 code system has been used in a wide range of applications, such as beam target design, accelerator shielding analysis, gas bremsstrahlung analysis, nuclear data evaluation, and so on. An example is calculation of photonuclear reaction (γ, n) yield and produced neutron energy distribution. In this work, the routine for photonuclear reaction yield and neutron energy distribution calculation was developed using the EGS4 code system. The photonuclear reaction yield was obtained by the convolution of the photonuclear reaction cross section and photon differential track length. The photonuclear reaction cross section was evaluated from Lorentz formula. Benchmark calculation was performed to compare our results with Hansen's those. The results obtained from the EGS4 code system and Hansen's those are in good agreement
The Los Alamos accelerator code group
International Nuclear Information System (INIS)
Krawczyk, F.L.; Billen, J.H.; Ryne, R.D.; Takeda, Harunori; Young, L.M.
1995-01-01
The Los Alamos Accelerator Code Group (LAACG) is a national resource for members of the accelerator community who use and/or develop software for the design and analysis of particle accelerators, beam transport systems, light sources, storage rings, and components of these systems. Below the authors describe the LAACG's activities in high performance computing, maintenance and enhancement of POISSON/SUPERFISH and related codes and the dissemination of information on the INTERNET
The Los Alamos accelerator code group
Energy Technology Data Exchange (ETDEWEB)
Krawczyk, F.L.; Billen, J.H.; Ryne, R.D.; Takeda, Harunori; Young, L.M.
1995-05-01
The Los Alamos Accelerator Code Group (LAACG) is a national resource for members of the accelerator community who use and/or develop software for the design and analysis of particle accelerators, beam transport systems, light sources, storage rings, and components of these systems. Below the authors describe the LAACG`s activities in high performance computing, maintenance and enhancement of POISSON/SUPERFISH and related codes and the dissemination of information on the INTERNET.
Remote-Handled Transuranic Content Codes
International Nuclear Information System (INIS)
2001-01-01
The Remote-Handled Transuranic (RH-TRU) Content Codes (RH-TRUCON) document represents the development of a uniform content code system for RH-TRU waste to be transported in the 72-Bcask. It will be used to convert existing waste form numbers, content codes, and site-specific identification codes into a system that is uniform across the U.S. Department of Energy (DOE) sites.The existing waste codes at the sites can be grouped under uniform content codes without any lossof waste characterization information. The RH-TRUCON document provides an all-encompassing description for each content code and compiles this information for all DOE sites. Compliance with waste generation, processing, and certification procedures at the sites (outlined in this document foreach content code) ensures that prohibited waste forms are not present in the waste. The content code gives an overall description of the RH-TRU waste material in terms of processes and packaging, as well as the generation location. This helps to provide cradle-to-grave traceability of the waste material so that the various actions required to assess its qualification as payload for the 72-B cask can be performed. The content codes also impose restrictions and requirements on the manner in which a payload can be assembled. The RH-TRU Waste Authorized Methods for Payload Control (RH-TRAMPAC), Appendix 1.3.7 of the 72-B Cask Safety Analysis Report (SAR), describes the current governing procedures applicable for the qualification of waste as payload for the 72-B cask. The logic for this classification is presented in the 72-B Cask SAR. Together, these documents (RH-TRUCON, RH-TRAMPAC, and relevant sections of the 72-B Cask SAR) present the foundation and justification for classifying RH-TRU waste into content codes. Only content codes described in thisdocument can be considered for transport in the 72-B cask. Revisions to this document will be madeas additional waste qualifies for transport. Each content code uniquely
International Nuclear Information System (INIS)
Sacramento, A.M. do.
1989-01-01
This user's manual contains all the necessary information concerning the use of SEVERO code. This computer code is related to the statistics of extremes = extreme winds, extreme precipitation and flooding hazard risk analysis. (A.C.A.S.)
International Nuclear Information System (INIS)
Anon.
1977-01-01
Progress in model and code development for reactor physics calculations is summarized. The codes included CINDER-10, PHROG, RAFFLE GAPP, DCFMR, RELAP/4, PARET, and KENO. Kinetics models for the PBF were developed
Djordjevic, Ivan; Vasic, Bane
2010-01-01
This unique book provides a coherent and comprehensive introduction to the fundamentals of optical communications, signal processing and coding for optical channels. It is the first to integrate the fundamentals of coding theory and optical communication.
Zhang, Linfan; Zheng, Shuang
2015-01-01
Quick Response code opens possibility to convey data in a unique way yet insufficient prevention and protection might lead into QR code being exploited on behalf of attackers. This thesis starts by presenting a general introduction of background and stating two problems regarding QR code security, which followed by a comprehensive research on both QR code itself and related issues. From the research a solution taking advantages of cloud and cryptography together with an implementation come af...
Xu, Mingliang; Su, Hao; Li, Yafei; Li, Xi; Liao, Jing; Niu, Jianwei; Lv, Pei; Zhou, Bing
2018-01-01
With the continued proliferation of smart mobile devices, Quick Response (QR) code has become one of the most-used types of two-dimensional code in the world. Aiming at beautifying the appearance of QR codes, existing works have developed a series of techniques to make the QR code more visual-pleasant. However, these works still leave much to be desired, such as visual diversity, aesthetic quality, flexibility, universal property, and robustness. To address these issues, in this paper, we pro...
ARC Code TI: CODE Software Framework
National Aeronautics and Space Administration — CODE is a software framework for control and observation in distributed environments. The basic functionality of the framework allows a user to observe a distributed...
ARC Code TI: ROC Curve Code Augmentation
National Aeronautics and Space Administration — ROC (Receiver Operating Characteristic) curve Code Augmentation was written by Rodney Martin and John Stutz at NASA Ames Research Center and is a modification of ROC...
A. van Deursen (Arie); L.M.F. Moonen (Leon); A. van den Bergh; G. Kok
2001-01-01
textabstractTwo key aspects of extreme programming (XP) are unit testing and merciless refactoring. Given the fact that the ideal test code / production code ratio approaches 1:1, it is not surprising that unit tests are being refactored. We found that refactoring test code is different from
Error Correcting Codes -34 ...
Indian Academy of Sciences (India)
the reading of data from memory the receiving process. Protecting data in computer memories was one of the earliest applications of Hamming codes. We now describe the clever scheme invented by Hamming in 1948. To keep things simple, we describe the binary length 7 Hamming code. Encoding in the Hamming Code.
Clinton, Janeen S.
This activity packet offers simple directions for setting up a Morse Code system appropriate to interfacing with any of several personal computer systems. Worksheets are also included to facilitate teaching Morse Code to persons with visual or other disabilities including blindness, as it is argued that the code is best learned auditorily. (PB)
Energy Technology Data Exchange (ETDEWEB)
Picini, P.; Caropreso, G.; Antonini, A.; Galuppi, G.; Sbrana, M. [ENEA, Centro Ricerche Casaccia, Rome (Italy). Dip. Ambiente; Bardone, G. [ENEA, Centro Ricerche Casaccia, Rome (Italy). Dip. Energia; Malvestuto, V.; Ricotta, A. [CNR, Rome (Italy). Ist. di Fisica dell`Atmosfera
1996-04-01
In the Part I of this paper a mathematical model to calculate the sampling and transport efficiencies (both in laminar and turbulent condition) of any sampling and transport system decomposable in several cylindrical elemental component is presented. In the Part II an experimental facility built in Casaccia ENEA laboratory is described and the measures carried out to validate the model are reported.
MUXS: a code to generate multigroup cross sections for sputtering calculations
Energy Technology Data Exchange (ETDEWEB)
Hoffman, T.J.; Robinson, M.T.; Dodds, H.L. Jr.
1982-10-01
This report documents MUXS, a computer code to generate multigroup cross sections for charged particle transport problems. Cross sections generated by MUXS can be used in many multigroup transport codes, with minor modifications to these codes, to calculate sputtering yields, reflection coefficients, penetration distances, etc.
Research on Primary Shielding Calculation Source Generation Codes
Zheng, Zheng; Mei, Qiliang; Li, Hui; Shangguan, Danhua; Zhang, Guangchun
2017-09-01
Primary Shielding Calculation (PSC) plays an important role in reactor shielding design and analysis. In order to facilitate PSC, a source generation code is developed to generate cumulative distribution functions (CDF) for the source particle sample code of the J Monte Carlo Transport (JMCT) code, and a source particle sample code is deveoped to sample source particle directions, types, coordinates, energy and weights from the CDFs. A source generation code is developed to transform three dimensional (3D) power distributions in xyz geometry to source distributions in r θ z geometry for the J Discrete Ordinate Transport (JSNT) code. Validation on PSC model of Qinshan No.1 nuclear power plant (NPP), CAP1400 and CAP1700 reactors are performed. Numerical results show that the theoretical model and the codes are both correct.
Preliminary design studies for the DESCARTES and CIDER codes
International Nuclear Information System (INIS)
Eslinger, P.W.; Miley, T.B.; Ouderkirk, S.J.; Nichols, W.E.
1992-12-01
The Hanford Environmental Dose Reconstruction (HEDR) project is developing several computer codes to model the release and transport of radionuclides into the environment. This preliminary design addresses two of these codes: Dynamic Estimates of Concentrations and Radionuclides in Terrestrial Environments (DESCARTES) and Calculation of Individual Doses from Environmental Radionuclides (CIDER). The DESCARTES code will be used to estimate the concentration of radionuclides in environmental pathways, given the output of the air transport code HATCHET. The CIDER code will use information provided by DESCARTES to estimate the dose received by an individual. This document reports on preliminary design work performed by the code development team to determine if the requirements could be met for Descartes and CIDER. The document contains three major sections: (i) a data flow diagram and discussion for DESCARTES, (ii) a data flow diagram and discussion for CIDER, and (iii) a series of brief statements regarding the design approach required to address each code requirement
The TESS [Tandem Experiment Simulation Studies] computer code user's manual
International Nuclear Information System (INIS)
Procassini, R.J.
1990-01-01
TESS (Tandem Experiment Simulation Studies) is a one-dimensional, bounded particle-in-cell (PIC) simulation code designed to investigate the confinement and transport of plasma in a magnetic mirror device, including tandem mirror configurations. Mirror plasmas may be modeled in a system which includes an applied magnetic field and/or a self-consistent or applied electrostatic potential. The PIC code TESS is similar to the PIC code DIPSI (Direct Implicit Plasma Surface Interactions) which is designed to study plasma transport to and interaction with a solid surface. The codes TESS and DIPSI are direct descendants of the PIC code ES1 that was created by A. B. Langdon. This document provides the user with a brief description of the methods used in the code and a tutorial on the use of the code. 10 refs., 2 tabs
Software Certification - Coding, Code, and Coders
Havelund, Klaus; Holzmann, Gerard J.
2011-01-01
We describe a certification approach for software development that has been adopted at our organization. JPL develops robotic spacecraft for the exploration of the solar system. The flight software that controls these spacecraft is considered to be mission critical. We argue that the goal of a software certification process cannot be the development of "perfect" software, i.e., software that can be formally proven to be correct under all imaginable and unimaginable circumstances. More realistically, the goal is to guarantee a software development process that is conducted by knowledgeable engineers, who follow generally accepted procedures to control known risks, while meeting agreed upon standards of workmanship. We target three specific issues that must be addressed in such a certification procedure: the coding process, the code that is developed, and the skills of the coders. The coding process is driven by standards (e.g., a coding standard) and tools. The code is mechanically checked against the standard with the help of state-of-the-art static source code analyzers. The coders, finally, are certified in on-site training courses that include formal exams.
DEFF Research Database (Denmark)
Soon, Winnie
2014-01-01
, Twitter and Facebook). The focus is not to investigate the functionalities and efficiencies of the code, but to study and interpret the program level of code in order to trace the use of various technological methods such as third-party libraries and platforms’ interfaces. These are important...... to understand the socio-technical side of a changing network environment. Through the study of code, including but not limited to source code, technical specifications and other materials in relation to the artwork production, I would like to explore the materiality of code that goes beyond technical...
Rice, R. F.; Lee, J. J.
1986-01-01
Scheme for coding facsimile messages promises to reduce data transmission requirements to one-tenth current level. Coding scheme paves way for true electronic mail in which handwritten, typed, or printed messages or diagrams sent virtually instantaneously - between buildings or between continents. Scheme, called Universal System for Efficient Electronic Mail (USEEM), uses unsupervised character recognition and adaptive noiseless coding of text. Image quality of resulting delivered messages improved over messages transmitted by conventional coding. Coding scheme compatible with direct-entry electronic mail as well as facsimile reproduction. Text transmitted in this scheme automatically translated to word-processor form.
International Nuclear Information System (INIS)
Jow, Hong-Nian; Murfin, W.B.; Johnson, J.D.
1993-11-01
This report describes the source term estimation codes, XSORs. The codes are written for three pressurized water reactors (Surry, Sequoyah, and Zion) and two boiling water reactors (Peach Bottom and Grand Gulf). The ensemble of codes has been named ''XSOR''. The purpose of XSOR codes is to estimate the source terms which would be released to the atmosphere in severe accidents. A source term includes the release fractions of several radionuclide groups, the timing and duration of releases, the rates of energy release, and the elevation of releases. The codes have been developed by Sandia National Laboratories for the US Nuclear Regulatory Commission (NRC) in support of the NUREG-1150 program. The XSOR codes are fast running parametric codes and are used as surrogates for detailed mechanistic codes. The XSOR codes also provide the capability to explore the phenomena and their uncertainty which are not currently modeled by the mechanistic codes. The uncertainty distributions of input parameters may be used by an. XSOR code to estimate the uncertainty of source terms
Energy Technology Data Exchange (ETDEWEB)
2014-05-14
DLLExternalCode is the a general dynamic-link library (DLL) interface for linking GoldSim (www.goldsim.com) with external codes. The overall concept is to use GoldSim as top level modeling software with interfaces to external codes for specific calculations. The DLLExternalCode DLL that performs the linking function is designed to take a list of code inputs from GoldSim, create an input file for the external application, run the external code, and return a list of outputs, read from files created by the external application, back to GoldSim. Instructions for creating the input file, running the external code, and reading the output are contained in an instructions file that is read and interpreted by the DLL.
Review of SKB's Code Documentation and Testing
International Nuclear Information System (INIS)
Hicks, T.W.
2005-01-01
SKB is in the process of developing the SR-Can safety assessment for a KBS 3 repository. The assessment will be based on quantitative analyses using a range of computational codes aimed at developing an understanding of how the repository system will evolve. Clear and comprehensive code documentation and testing will engender confidence in the results of the safety assessment calculations. This report presents the results of a review undertaken on behalf of SKI aimed at providing an understanding of how codes used in the SR 97 safety assessment and those planned for use in the SR-Can safety assessment have been documented and tested. Having identified the codes us ed by SKB, several codes were selected for review. Consideration was given to codes used directly in SKB's safety assessment calculations as well as to some of the less visible codes that are important in quantifying the different repository barrier safety functions. SKB's documentation and testing of the following codes were reviewed: COMP23 - a near-field radionuclide transport model developed by SKB for use in safety assessment calculations. FARF31 - a far-field radionuclide transport model developed by SKB for use in safety assessment calculations. PROPER - SKB's harness for executing probabilistic radionuclide transport calculations using COMP23 and FARF31. The integrated analytical radionuclide transport model that SKB has developed to run in parallel with COMP23 and FARF31. CONNECTFLOW - a discrete fracture network model/continuum model developed by Serco Assurance (based on the coupling of NAMMU and NAPSAC), which SKB is using to combine hydrogeological modelling on the site and regional scales in place of the HYDRASTAR code. DarcyTools - a discrete fracture network model coupled to a continuum model, recently developed by SKB for hydrogeological modelling, also in place of HYDRASTAR. ABAQUS - a finite element material model developed by ABAQUS, Inc, which is used by SKB to model repository buffer
Multiple application coded switch development report
International Nuclear Information System (INIS)
Bernal, E.L.; Kestly, J.D.
1979-03-01
The development of the Multiple Application Coded Switch (MACS) and its related controller are documented; the functional and electrical characteristics are described; the interface requirements defined, and a troubleshooting guide provided. The system was designed for the Safe Secure Trailer System used for secure transportation of nuclear material
LATTICE: an interactive lattice computer code
International Nuclear Information System (INIS)
Staples, J.
1976-10-01
LATTICE is a computer code which enables an interactive user to calculate the functions of a synchrotron lattice. This program satisfies the requirements at LBL for a simple interactive lattice program by borrowing ideas from both TRANSPORT and SYNCH. A fitting routine is included
49 CFR 171.25 - Additional requirements for the use of the IMDG Code.
2010-10-01
... material poisonous (toxic) by inhalation, the T Codes specified in Column 13 of the Dangerous Goods List in... 49 Transportation 2 2010-10-01 2010-10-01 false Additional requirements for the use of the IMDG Code. 171.25 Section 171.25 Transportation Other Regulations Relating to Transportation PIPELINE AND...
SWAAM code development, verification and application to steam generator design
International Nuclear Information System (INIS)
Shin, Y.W.; Valentin, R.A.
1990-01-01
This paper describes the family of SWAAM codes developed by Argonne National Laboratory to analyze the effects of sodium/water reactions on LMR steam generators. The SWAAM codes were developed as design tools for analyzing various phenomena related to steam generator leaks and to predict the resulting thermal and hydraulic effects on the steam generator and the intermediate heat transport system (IHTS). The theoretical foundations and numerical treatments on which the codes are based are discussed, followed by a description of code capabilities and limitations, verification of the codes by comparison with experiment, and applications to steam generator and IHTS design. (author). 25 refs, 14 figs
Vectorization and parallelization of a production reactor assembly code
International Nuclear Information System (INIS)
Vujic, J.L.; Martin, W.R.
1991-01-01
In order to efficiently use new features of supercomputers, production codes, usually written 10 - 20 years ago, must be tailored for modern computer architectures. We have chosen to optimize the CPM-2 code, a production reactor assembly code based on the collision probability transport method. Substantial speedups in the execution times were obtained with the parallel/vector version of the CPM-2 code. In addition, we have developed a new transfer probability method, which removes some of the modelling limitations of the collision probability method encoded in the CPM-2 code, and can fully utilize parallel/vector architecture of a multiprocessor IBM 3090. (author)
Energy Technology Data Exchange (ETDEWEB)
Washington TRU Solutions
2007-07-01
The Remote-Handled Transuranic (RH-TRU) Content Codes (RH-TRUCON) document describes the inventory of RH-TRU waste within the transportation parameters specified by the Remote-Handled Transuranic Waste Authorized Methods for Payload Control (RH-TRAMPAC).1 The RH-TRAMPAC defines the allowable payload for the RH-TRU 72-B. This document is a catalog of RH-TRU 72-B authorized contents by site. A content code is defined by the following components: • A two-letter site abbreviation that designates the physical location of the generated/stored waste (e.g., ID for Idaho National Laboratory [INL]). The site-specific letter designations for each of the sites are provided in Table 1. • A three-digit code that designates the physical and chemical form of the waste (e.g., content code 317 denotes TRU Metal Waste). For RH-TRU waste to be transported in the RH-TRU 72-B, the first number of this three-digit code is “3.” The second and third numbers of the three-digit code describe the physical and chemical form of the waste. Table 2 provides a brief description of each generic code. Content codes are further defined as subcodes by an alpha trailer after the three-digit code to allow segregation of wastes that differ in one or more parameter(s). For example, the alpha trailers of the subcodes ID 322A and ID 322B may be used to differentiate between waste packaging configurations. As detailed in the RH-TRAMPAC, compliance with flammable gas limits may be demonstrated through the evaluation of compliance with either a decay heat limit or flammable gas generation rate (FGGR) limit per container specified in approved content codes. As applicable, if a container meets the watt*year criteria specified by the RH-TRAMPAC, the decay heat limits based on the dose-dependent G value may be used as specified in an approved content code. If a site implements the administrative controls outlined in the RH-TRAMPAC and Appendix 2.4 of the RH-TRU Payload Appendices, the decay heat or FGGR
Network coding at different layers in wireless networks
2016-01-01
This book focuses on how to apply network coding at different layers in wireless networks – including MAC, routing, and TCP – with special focus on cognitive radio networks. It discusses how to select parameters in network coding (e.g., coding field, number of packets involved, and redundant information ration) in order to be suitable for the varying wireless environments. The book explores how to deploy network coding in MAC to improve network performance and examines joint network coding with opportunistic routing to improve the successful rate of routing. In regards to TCP and network coding, the text considers transport layer protocol working with network coding to overcome the transmission error rate, particularly with how to use the ACK feedback of TCP to enhance the efficiency of network coding. The book pertains to researchers and postgraduate students, especially whose interests are in opportunistic routing and TCP in cognitive radio networks.
DEFF Research Database (Denmark)
Andersen, Christian Ulrik
2007-01-01
Computer art is often associated with computer-generated expressions (digitally manipulated audio/images in music, video, stage design, media facades, etc.). In recent computer art, however, the code-text itself – not the generated output – has become the artwork (Perl Poetry, ASCII Art, obfuscated...... code, etc.). The presentation relates this artistic fascination of code to a media critique expressed by Florian Cramer, claiming that the graphical interface represents a media separation (of text/code and image) causing alienation to the computer’s materiality. Cramer is thus the voice of a new ‘code...... avant-garde’. In line with Cramer, the artists Alex McLean and Adrian Ward (aka Slub) declare: “art-oriented programming needs to acknowledge the conditions of its own making – its poesis.” By analysing the Live Coding performances of Slub (where they program computer music live), the presentation...
Special issue on network coding
Monteiro, Francisco A.; Burr, Alister; Chatzigeorgiou, Ioannis; Hollanti, Camilla; Krikidis, Ioannis; Seferoglu, Hulya; Skachek, Vitaly
2017-12-01
Future networks are expected to depart from traditional routing schemes in order to embrace network coding (NC)-based schemes. These have created a lot of interest both in academia and industry in recent years. Under the NC paradigm, symbols are transported through the network by combining several information streams originating from the same or different sources. This special issue contains thirteen papers, some dealing with design aspects of NC and related concepts (e.g., fountain codes) and some showcasing the application of NC to new services and technologies, such as data multi-view streaming of video or underwater sensor networks. One can find papers that show how NC turns data transmission more robust to packet losses, faster to decode, and more resilient to network changes, such as dynamic topologies and different user options, and how NC can improve the overall throughput. This issue also includes papers showing that NC principles can be used at different layers of the networks (including the physical layer) and how the same fundamental principles can lead to new distributed storage systems. Some of the papers in this issue have a theoretical nature, including code design, while others describe hardware testbeds and prototypes.
High-fidelity plasma codes for burn physics
Energy Technology Data Exchange (ETDEWEB)
Cooley, James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Graziani, Frank [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Marinak, Marty [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Murillo, Michael [Michigan State Univ., East Lansing, MI (United States)
2016-10-19
Accurate predictions of equation of state (EOS), ionic and electronic transport properties are of critical importance for high-energy-density plasma science. Transport coefficients inform radiation-hydrodynamic codes and impact diagnostic interpretation, which in turn impacts our understanding of the development of instabilities, the overall energy balance of burning plasmas, and the efficacy of self-heating from charged-particle stopping. Important processes include thermal and electrical conduction, electron-ion coupling, inter-diffusion, ion viscosity, and charged particle stopping. However, uncertainties in these coefficients are not well established. Fundamental plasma science codes, also called high-fidelity plasma codes, are a relatively recent computational tool that augments both experimental data and theoretical foundations of transport coefficients. This paper addresses the current status of HFPC codes and their future development, and the potential impact they play in improving the predictive capability of the multi-physics hydrodynamic codes used in HED design.
Energy Technology Data Exchange (ETDEWEB)
Visser, B. [Stork Product Eng., Amsterdam (Netherlands)
1996-09-01
To support the discussion on aeroelastic codes, a description of the code FLEXLAST was given and experiences within benchmarks and measurement programmes were summarized. The code FLEXLAST has been developed since 1982 at Stork Product Engineering (SPE). Since 1992 FLEXLAST has been used by Dutch industries for wind turbine and rotor design. Based on the comparison with measurements, it can be concluded that the main shortcomings of wind turbine modelling lie in the field of aerodynamics, wind field and wake modelling. (au)
DEFF Research Database (Denmark)
Bombin Palomo, Hector
2015-01-01
Color codes are topological stabilizer codes with unusual transversality properties. Here I show that their group of transversal gates is optimal and only depends on the spatial dimension, not the local geometry. I also introduce a generalized, subsystem version of color codes. In 3D they allow t...... the transversal implementation of a universal set of gates by gauge fixing, while error-dectecting measurements involve only four or six qubits....
Bravyi, Sergey
Combining protection from noise and computational universality is one of the biggest challenges in the fault-tolerant quantum computing. Topological stabilizer codes such as the 2D surface code can tolerate a high level of noise but implementing logical gates, especially non-Clifford ones, requires a prohibitively large overhead due to the need of state distillation. In this talk I will describe a new family of 2D quantum error correcting codes that enable a transversal implementation of all logical gates required for the universal quantum computing. Transversal logical gates (TLG) are encoded operations that can be realized by applying some single-qubit rotation to each physical qubit. TLG are highly desirable since they introduce no overhead and do not spread errors. It has been known before that a quantum code can have only a finite number of TLGs which rules out computational universality. Our scheme circumvents this no-go result by combining TLGs of two different quantum codes using the gauge-fixing method pioneered by Paetznick and Reichardt. The first code, closely related to the 2D color code, enables a transversal implementation of all single-qubit Clifford gates such as the Hadamard gate and the π / 2 phase shift. The second code that we call a doubled color code provides a transversal T-gate, where T is the π / 4 phase shift. The Clifford+T gate set is known to be computationally universal. The two codes can be laid out on the honeycomb lattice with two qubits per site such that the code conversion requires parity measurements for six-qubit Pauli operators supported on faces of the lattice. I will also describe numerical simulations of logical Clifford+T circuits encoded by the distance-3 doubled color code. Based on a joint work with Andrew Cross.
Phonological coding during reading
Leinenger, Mallorie
2014-01-01
The exact role that phonological coding (the recoding of written, orthographic information into a sound based code) plays during silent reading has been extensively studied for more than a century. Despite the large body of research surrounding the topic, varying theories as to the time course and function of this recoding still exist. The present review synthesizes this body of research, addressing the topics of time course and function in tandem. The varying theories surrounding the function of phonological coding (e.g., that phonological codes aid lexical access, that phonological codes aid comprehension and bolster short-term memory, or that phonological codes are largely epiphenomenal in skilled readers) are first outlined, and the time courses that each maps onto (e.g., that phonological codes come online early (pre-lexical) or that phonological codes come online late (post-lexical)) are discussed. Next the research relevant to each of these proposed functions is reviewed, discussing the varying methodologies that have been used to investigate phonological coding (e.g., response time methods, reading while eyetracking or recording EEG and MEG, concurrent articulation) and highlighting the advantages and limitations of each with respect to the study of phonological coding. In response to the view that phonological coding is largely epiphenomenal in skilled readers, research on the use of phonological codes in prelingually, profoundly deaf readers is reviewed. Finally, implications for current models of word identification (activation-verification model (Van Order, 1987), dual-route model (e.g., Coltheart, Rastle, Perry, Langdon, & Ziegler, 2001), parallel distributed processing model (Seidenberg & McClelland, 1989)) are discussed. PMID:25150679
1988-01-01
American Bar Codes, Inc. developed special bar code labels for inventory control of space shuttle parts and other space system components. ABC labels are made in a company-developed anodizing aluminum process and consecutively marketed with bar code symbology and human readable numbers. They offer extreme abrasion resistance and indefinite resistance to ultraviolet radiation, capable of withstanding 700 degree temperatures without deterioration and up to 1400 degrees with special designs. They offer high resistance to salt spray, cleaning fluids and mild acids. ABC is now producing these bar code labels commercially or industrial customers who also need labels to resist harsh environments.
Waters, Joe
2012-01-01
Find out how to effectively create, use, and track QR codes QR (Quick Response) codes are popping up everywhere, and businesses are reaping the rewards. Get in on the action with the no-nonsense advice in this streamlined, portable guide. You'll find out how to get started, plan your strategy, and actually create the codes. Then you'll learn to link codes to mobile-friendly content, track your results, and develop ways to give your customers value that will keep them coming back. It's all presented in the straightforward style you've come to know and love, with a dash of humor thrown
International Nuclear Information System (INIS)
Reid, R.L.; Barrett, R.J.; Brown, T.G.
1985-03-01
The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged
National Aeronautics and Space Administration — ACCEPT consists of an overall software infrastructure framework and two main software components. The software infrastructure framework consists of code written to...
Forms of Approximate Radiation Transport
Brunner, G
2002-01-01
Photon radiation transport is described by the Boltzmann equation. Because this equation is difficult to solve, many different approximate forms have been implemented in computer codes. Several of the most common approximations are reviewed, and test problems illustrate the characteristics of each of the approximations. This document is designed as a tutorial so that code users can make an educated choice about which form of approximate radiation transport to use for their particular simulation.
Advanced thermionic reactor systems design code
International Nuclear Information System (INIS)
Lewis, B.R.; Pawlowski, R.A.; Greek, K.J.; Klein, A.C.
1991-01-01
An overall systems design code is under development to model an advanced in-core thermionic nuclear reactor system for space applications at power levels of 10 to 50 kWe. The design code is written in an object-oriented programming environment that allows the use of a series of design modules, each of which is responsible for the determination of specific system parameters. The code modules include a neutronics and core criticality module, a core thermal hydraulics module, a thermionic fuel element performance module, a radiation shielding module, a module for waste heat transfer and rejection, and modules for power conditioning and control. The neutronics and core criticality module determines critical core size, core lifetime, and shutdown margins using the criticality calculation capability of the Monte Carlo Neutron and Photon Transport Code System (MCNP). The remaining modules utilize results of the MCNP analysis along with FORTRAN programming to predict the overall system performance
International Nuclear Information System (INIS)
Kimura, Hideo; Takahashi, Tomoyuki; Shima, Shigeki; Matsuzuru, Hideo
1992-11-01
The computer code system GSRW (Generic Safety assessment code for geologic disposal of Radioactive Waste) was developed as in interim version of safety assessment methodology for geologic disposal of high-level radioactive waste. Scenarios used here are based on normal evolution scenarios which assume that the performance of a disposal system is not affected by probabilistic events. The code consists of three parts. The first part evaluates a source term from a disposal facility which consists mainly of a vitrified waste, a metallic container and a buffer zone. Two kinds of source term models are provided: Model 1 which simulate the dissolution of silicate component of glass and the diffusive transport of radionuclides in the buffere zone, and Model 2 which assumes that the concentration of a radionuclide is limited by the solubility of its specific chemical form at the interface between the buffer and a vitrified wastes. The second part analyses the transport of radionuclides in the geosphere, which is based on analytical solutions or numerical solutions of a mass transport equation involving the advection, dispersion, linear sorption and decay chain. The third part assesses the transport of radionuclides in the biosphere and the resulting radiological consequences to the man, which is based on a dynamic compartment model for the biosphere and a dose factor method for dose calculations. This report describes mathematical models used, the structure of the code system, and user information and instructions for execution of the code. (author)
Carlet, Claude; Mesnager, Sihem; Tang, Chunming; Qi, Yanfeng
2017-01-01
Linear complementary pairs (LCP) of codes play an important role in armoring implementations against side-channel attacks and fault injection attacks. One of the most common ways to construct LCP of codes is to use Euclidean linear complementary dual (LCD) codes. In this paper, we first introduce the concept of linear codes with $\\sigma$ complementary dual ($\\sigma$-LCD), which includes known Euclidean LCD codes, Hermitian LCD codes, and Galois LCD codes. As Euclidean LCD codes, $\\sigma$-LCD ...
International Nuclear Information System (INIS)
Wilson, J.L.; RamaRao, B.S.; McNeish, J.A.
1986-11-01
GRASP (GRound-Water Adjunct Senstivity Program) computes measures of the behavior of a ground-water system and the system's performance for waste isolation, and estimates the sensitivities of these measures to system parameters. The computed measures are referred to as ''performance measures'' and include weighted squared deviations of computed and observed pressures or heads, local Darcy velocity components and magnitudes, boundary fluxes, and travel distance and time along travel paths. The sensitivities are computed by the adjoint method and are exact derivatives of the performance measures with respect to the parameters for the modeled system, taken about the assumed parameter values. GRASP presumes steady-state, saturated grondwater flow, and post-processes the results of a multidimensional (1-D, 2-D, 3-D) finite-difference flow code. This document describes the mathematical basis for the model, the algorithms and solution techniques used, and the computer code design. The implementation of GRASP is verified with simple one- and two-dimensional flow problems, for which analytical expressions of performance measures and sensitivities are derived. The linkage between GRASP and multidimensional finite-difference flow codes is described. This document also contains a detailed user's manual. The use of GRASP to evaluate nuclear waste disposal issues has been emphasized throughout the report. The performance measures and their sensitivities can be employed to assist in directing data collection programs, expedite model calibration, and objectively determine the sensitivity of projected system performance to parameters
SRAC2006: A comprehensive neutronics calculation code system
International Nuclear Information System (INIS)
Okumura, Keisuke; Kugo, Teruhiko; Kaneko, Kunio; Tsuchihashi, Keichiro
2007-02-01
The SRAC is a code system applicable to neutronics analysis of a variety of reactor types. Since the publication of the second version of the users manual (JAERI-1302) in 1986 for the SRAC system, a number of additions and modifications to the functions and the library data have been made to establish a comprehensive neutronics code system. The current system includes major neutron data libraries (JENDL-3.3, JENDL-3.2, ENDF/B-VII, ENDF/B-VI.8, JEFF-3.1, JEF-2.2, etc.), and integrates five elementary codes for neutron transport and diffusion calculation; PIJ based on the collision probability method applicable to 16 kind of lattice models, S N transport codes ANISN(1D) and TWOTRN(2D), diffusion codes TUD(1D) and CITATION(multi-D). The system also includes an auxiliary code COREBN for multi-dimensional core burn-up calculation. (author)
International Nuclear Information System (INIS)
Lacoste, Andre-Claude
2005-01-01
This guide notably contains various forms associated with the declaration of significant events, and explanations to fill them in: significant event declaration form for a basic nuclear installation, significant event declaration form for radioactive material transport, significant event report for a basic nuclear installation, significant event report for radioactive material transport, declaration criteria for significant events related to the safety of non-PWR basic nuclear installations, declaration criteria for significant events related to PWR safety, significant events declared further to events resulting in group 1 unavailability and non-compliance with technical operating specifications, declaration criteria for significant events concerning radiation protection for basic nuclear installations, declaration criteria for significant events concerning environmental protection, applicable to basic nuclear installations, and declaration criteria for significant events concerning radioactive material transport
Burnup code for fuel assembly by Monte Carlo code. MKENO-BURN
International Nuclear Information System (INIS)
Naito, Yoshitaka; Suyama, Kenya; Masukawa, Fumihiro; Matsumoto, Kiyoshi; Kurosawa, Masayoshi; Kaneko, Toshiyuki.
1996-12-01
The evaluation of neutron spectrum is so important for burnup calculation of the heterogeneous geometry like recent BWR fuel assembly. MKENO-BURN is a multi dimensional burnup code that based on the three dimensional monte carlo neutron transport code 'MULTI-KENO' and the routine for the burnup calculation of the one dimensional burnup code 'UNITBURN'. MKENO-BURN analyzes the burnup problem of arbitrary regions after evaluating the neutron spectrum and making one group cross section in three dimensional geometry with MULTI-KENO. It enables us to do three dimensional burnup calculation. This report consists of general description of MKENO-BURN and the input data. (author)
Burnup code for fuel assembly by Monte Carlo code. MKENO-BURN
Energy Technology Data Exchange (ETDEWEB)
Naito, Yoshitaka; Suyama, Kenya; Masukawa, Fumihiro; Matsumoto, Kiyoshi; Kurosawa, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Toshiyuki
1996-12-01
The evaluation of neutron spectrum is so important for burnup calculation of the heterogeneous geometry like recent BWR fuel assembly. MKENO-BURN is a multi dimensional burnup code that based on the three dimensional monte carlo neutron transport code `MULTI-KENO` and the routine for the burnup calculation of the one dimensional burnup code `UNITBURN`. MKENO-BURN analyzes the burnup problem of arbitrary regions after evaluating the neutron spectrum and making one group cross section in three dimensional geometry with MULTI-KENO. It enables us to do three dimensional burnup calculation. This report consists of general description of MKENO-BURN and the input data. (author)
Remote-Handled Transuranic Content Codes
Energy Technology Data Exchange (ETDEWEB)
Washington TRU Solutions
2006-12-01
The Remote-Handled Transuranic (RH-TRU) Content Codes (RH-TRUCON) document describes the inventory of RH-TRU waste within the transportation parameters specified by the Remote-Handled Transuranic Waste Authorized Methods for Payload Control (RH-TRAMPAC).1 The RH-TRAMPAC defines the allowable payload for the RH-TRU 72-B. This document is a catalog of RH-TRU 72-B authorized contents by site. A content code is defined by the following components: • A two-letter site abbreviation that designates the physical location of the generated/stored waste (e.g., ID for Idaho National Laboratory [INL]). The site-specific letter designations for each of the sites are provided in Table 1. • A three-digit code that designates the physical and chemical form of the waste (e.g., content code 317 denotes TRU Metal Waste). For RH-TRU waste to be transported in the RH-TRU 72-B, the first number of this three-digit code is “3.” The second and third numbers of the three-digit code describe the physical and chemical form of the waste. Table 2 provides a brief description of each generic code. Content codes are further defined as subcodes by an alpha trailer after the three-digit code to allow segregation of wastes that differ in one or more parameter(s). For example, the alpha trailers of the subcodes ID 322A and ID 322B may be used to differentiate between waste packaging configurations. As detailed in the RH-TRAMPAC, compliance with flammable gas limits may be demonstrated through the evaluation of compliance with either a decay heat limit or flammable gas generation rate (FGGR) limit per container specified in approved content codes. As applicable, if a container meets the watt*year criteria specified by the RH-TRAMPAC, the decay heat limits based on the dose-dependent G value may be used as specified in an approved content code. If a site implements the administrative controls outlined in the RH-TRAMPAC and Appendix 2.4 of the RH-TRU Payload Appendices, the decay heat or FGGR
MELCOR Accident Consequence Code System (MACCS)
International Nuclear Information System (INIS)
Jow, H.N.; Sprung, J.L.; Ritchie, L.T.; Rollstin, J.A.; Chanin, D.I.
1990-02-01
This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management. 59 refs., 14 figs., 15 tabs
MELCOR Accident Consequence Code System (MACCS)
Energy Technology Data Exchange (ETDEWEB)
Jow, H.N.; Sprung, J.L.; Ritchie, L.T. (Sandia National Labs., Albuquerque, NM (USA)); Rollstin, J.A. (GRAM, Inc., Albuquerque, NM (USA)); Chanin, D.I. (Technadyne Engineering Consultants, Inc., Albuquerque, NM (USA))
1990-02-01
This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management. 59 refs., 14 figs., 15 tabs.
MELCOR Accident Consequence Code System (MACCS)
International Nuclear Information System (INIS)
Rollstin, J.A.; Chanin, D.I.; Jow, H.N.
1990-02-01
This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projections, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management
Energy Technology Data Exchange (ETDEWEB)
NONE
2003-07-01
These scientific days had for objective to draw up the situation of calculation codes of radiation transport, of sources estimation, of radiation doses managements and to draw the future perspectives. (N.C.)
An Overview of the Monte Carlo Methods, Codes, & Applications Group
Energy Technology Data Exchange (ETDEWEB)
Trahan, Travis John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2016-08-30
This report sketches the work of the Group to deliver first-principle Monte Carlo methods, production quality codes, and radiation transport-based computational and experimental assessments using the codes MCNP and MCATK for such applications as criticality safety, non-proliferation, nuclear energy, nuclear threat reduction and response, radiation detection and measurement, radiation health protection, and stockpile stewardship.
Integrated code development for studying laser driven plasmas
Energy Technology Data Exchange (ETDEWEB)
Takabe, Hideaki; Nagatomo, Hideo; Sunahara, Atsusi; Ohnishi, Naofumi; Naruo, Syuji; Mima, Kunioki [Osaka Univ., Suita (Japan). Inst. of Laser Engineering
1998-03-01
Present status and plan for developing an integrated implosion code are briefly explained by focusing on motivation, numerical scheme and issues to be developed more. Highly nonlinear stage of Rayleigh-Taylor instability of ablation front by laser irradiation has been simulated so as to be compared with model experiments. Improvement in transport and rezoning/remapping algorithms in ILESTA code is described. (author)
49 CFR 19.42 - Codes of conduct.
2010-10-01
... 49 Transportation 1 2010-10-01 2010-10-01 false Codes of conduct. 19.42 Section 19.42... Requirements Procurement Standards § 19.42 Codes of conduct. The recipient shall maintain written standards of conduct governing the performance of its employees engaged in the award and administration of contracts...
DEFF Research Database (Denmark)
Steensig, Jakob; Heinemann, Trine
2015-01-01
We welcome Tanya Stivers’s discussion (Stivers, 2015/this issue) of coding social interaction and find that her descriptions of the processes of coding open up important avenues for discussion, among other things of the precise ad hoc considerations that researchers need to bear in mind, both when...
Error Correcting Codes -34 ...
Indian Academy of Sciences (India)
Science, Bangalore. Her interests are in. Theoretical Computer. Science. SERIES I ARTICLE. Error Correcting Codes. 2. The Hamming Codes. Priti Shankar. In the first article of this series we showed how redundancy introduced into a message transmitted over a noisy channel could improve the reliability of transmission. In.
Indian Academy of Sciences (India)
set up a well defined goal - that of achieving a per- formance bound set by the noisy channel coding theo- rem, proved in the paper. Whereas the goal appeared elusive twenty five years ago, today, there are practi- cal codes and decoding algorithms that come close to achieving it. It is interesting to note that all known.
Indian Academy of Sciences (India)
Home; Journals; Resonance – Journal of Science Education; Volume 2; Issue 3. Error Correcting Codes - Reed Solomon Codes. Priti Shankar. Series Article Volume 2 Issue 3 March 1997 pp 33-47. Fulltext. Click here to view fulltext PDF. Permanent link: http://www.ias.ac.in/article/fulltext/reso/002/03/0033-0047 ...
Napier, Rebecca H; Bruelheide, Lori S; Demann, Eric T K; Haug, Richard H
2008-07-01
The purpose of this article is to highlight the importance of understanding various numeric and alpha-numeric codes for accurately billing dental and medically related services to private pay or third-party insurance carriers. In the United States, common dental terminology (CDT) codes are most commonly used by dentists to submit claims, whereas current procedural terminology (CPT) and International Classification of Diseases, Ninth Revision, Clinical Modification (ICD.9.CM) codes are more commonly used by physicians to bill for their services. The CPT and ICD.9.CM coding systems complement each other in that CPT codes provide the procedure and service information and ICD.9.CM codes provide the reason or rationale for a particular procedure or service. These codes are more commonly used for "medical necessity" determinations, and general dentists and specialists who routinely perform care, including trauma-related care, biopsies, and dental treatment as a result of or in anticipation of a cancer-related treatment, are likely to use these codes. Claim submissions for care provided can be completed electronically or by means of paper forms.