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...
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)
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
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.
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.
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.
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
DEFF Research Database (Denmark)
Ortúzar, Juan de Dios; Cherchi, Elisabetta; Rizzi, Luis
2014-01-01
Transport is a large, multidisciplinary and fascinating field, encompassing vastly different areas of research. In fact transport interests span from not very well understood (in fieldwork) issues related with survey methods to highly complex questions associated with the dynamic equilibration...... of supply and demand in strategic planning contexts; the latter involving large zoning systems, huge multimodal networks and highly complex dynamic modelling approaches (Mahmassani, 2001). But questions also arise at a more macro level (and in a different time span) regarding the interaction of transport...... and land use, and also at the more micro level with the dynamics of road traffic and public transport modelling, an area which is particularly interesting due to its high complexity in less developed nations (de Cea et al., 2005). We do not have the expertise or the space to dwell on all these issues...
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.
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
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
Codes and standards research, development and demonstration roadmap
Energy Technology Data Exchange (ETDEWEB)
None, None
2008-07-22
C&S RD&D Roadmap - 2008: This Roadmap is a guide to the Research, Development & Demonstration activities that will provide data required for Standards Development Organizations (SDOs) to develop performance-based codes and standards for a commercial hydrogen fueled transportation sector in the U.S.
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.)
Transport research: Quo Vadis?
CSIR Research Space (South Africa)
Rust, FC
2008-07-01
Full Text Available It is well-recognised internationally that transport and transport infrastructure play a major role both in the stimulation of economic growth, creation of job opportunities and in poverty alleviation. This is of particular importance in South...
Student Dress Codes and Uniforms. Research Brief
Johnston, Howard
2009-01-01
According to an Education Commission of the States "Policy Report", research on the effects of dress code and school uniform policies is inconclusive and mixed. Some researchers find positive effects; others claim no effects or only perceived effects. While no state has legislatively mandated the wearing of school uniforms, 28 states and…
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
CSIR Research Space (South Africa)
Mokonyama, Mathetha T
2008-03-01
Full Text Available In South Africa, airport and airline services epitomise what many would like to see in everyday public transport. The CSIR investigates what it will take to provide a commercial public transport service in South Africa which resembles commercial air...
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
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
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
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
Morris, Suzanne E.
2010-01-01
This paper provides a review of institutional authorship policies as required by the "Australian Code for the Responsible Conduct of Research" (the "Code") (National Health and Medical Research Council (NHMRC), the Australian Research Council (ARC) & Universities Australia (UA) 2007), and assesses them for Code compliance.…
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)
Subsurface transport program: Research summary
International Nuclear Information System (INIS)
1987-01-01
DOE's research program in subsurface transport is designed to provide a base of fundamental scientific information so that the geochemical, hydrological, and biological mechanisms that contribute to the transport and long term fate of energy related contaminants in subsurface ecosystems can be understood. Understanding the physical and chemical mechanisms that control the transport of single and co-contaminants is the underlying concern of the program. Particular attention is given to interdisciplinary research and to geosphere-biosphere interactions. The scientific results of the program will contribute to resolving Departmental questions related to the disposal of energy-producing and defense wastes. The background papers prepared in support of this document contain additional information on the relevance of the research in the long term to energy-producing technologies. Detailed scientific plans and other research documents are available for high priority research areas, for example, in subsurface transport of organic chemicals and mixtures and in the microbiology of deep aquifers. 5 figs., 1 tab
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.
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,
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.
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.
Ethical conduct for research : a code of scientific ethics
Marcia Patton-Mallory; Kathleen Franzreb; Charles Carll; Richard Cline
2000-01-01
The USDA Forest Service recently developed and adopted a code of ethical conduct for scientific research and development. The code addresses issues related to research misconduct, such as fabrication, falsification, or plagiarism in proposing, performing, or reviewing research or in reporting research results, as well as issues related to professional misconduct, such...
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
Transportation Research Analysis Computing Center (TRACC)
Federal Laboratory Consortium — Argonne National Laboratory initiated a multi-year program with the US Department of Transportation (USDOT) in October 2006, to establish the Transportation Research...
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
What to do with a Dead Research Code
Nemiroff, Robert J.
2016-01-01
The project has ended -- should all of the computer codes that enabled the project be deleted? No. Like research papers, research codes typically carry valuable information past project end dates. Several possible end states to the life of research codes are reviewed. Historically, codes are typically left dormant on an increasingly obscure local disk directory until forgotten. These codes will likely become any or all of: lost, impossible to compile and run, difficult to decipher, and likely deleted when the code's proprietor moves on or dies. It is argued here, though, that it would be better for both code authors and astronomy generally if project codes were archived after use in some way. Archiving is advantageous for code authors because archived codes might increase the author's ADS citable publications, while astronomy as a science gains transparency and reproducibility. Paper-specific codes should be included in the publication of the journal papers they support, just like figures and tables. General codes that support multiple papers, possibly written by multiple authors, including their supporting websites, should be registered with a code registry such as the Astrophysics Source Code Library (ASCL). Codes developed on GitHub can be archived with a third party service such as, currently, BackHub. An important code version might be uploaded to a web archiving service like, currently, Zenodo or Figshare, so that this version receives a Digital Object Identifier (DOI), enabling it to found at a stable address into the future. Similar archiving services that are not DOI-dependent include perma.cc and the Internet Archive Wayback Machine at archive.org. Perhaps most simply, copies of important codes with lasting value might be kept on a cloud service like, for example, Google Drive, while activating Google's Inactive Account Manager.
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)
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
Research on pre-processing of QR Code
Sun, Haixing; Xia, Haojie; Dong, Ning
2013-10-01
QR code encodes many kinds of information because of its advantages: large storage capacity, high reliability, full arrange of utter-high-speed reading, small printing size and high-efficient representation of Chinese characters, etc. In order to obtain the clearer binarization image from complex background, and improve the recognition rate of QR code, this paper researches on pre-processing methods of QR code (Quick Response Code), and shows algorithms and results of image pre-processing for QR code recognition. Improve the conventional method by changing the Souvola's adaptive text recognition method. Additionally, introduce the QR code Extraction which adapts to different image size, flexible image correction approach, and improve the efficiency and accuracy of QR code image processing.
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)
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)
Collaboration between physical activity researchers and transport planners
DEFF Research Database (Denmark)
Crist, Katie; Bolling, Khalisa; Schipperijn, Jasper
2018-01-01
sources. Informants believed that research collaboration could increase capacity by providing unbiased data and access to students to assist with targeted research. Collaboration could also increase the relevance of academic research in applied settings. Identified barriers included: setting up contracts...... the benefits and barriers of researcher and transport agency collaboration, and 4) identify the facilitators to collaboration for these groups. Semi-structured interviews were conducted with 17 transport modeling, planning or engineering professionals, transport agency directors, and academics with relevant...... expertise in health or transport planning. A thematic analysis was conducted following structural coding by two researchers. The analysis revealed that geographic and physical activity data that are current, local, objective and specific to individual AT trips would improve upon currently available data...
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
Research and application of nanotechnology in transportation
CSIR Research Space (South Africa)
Steyn, WJvdM
2008-07-01
Full Text Available of many of the developments in the nanotechnology field in the area of transportation engineering is growing. Current research in this area focus on the development of improved materials for construction of transportation facilities, characterisation...
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)
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)
Light ion beam transport research at NRL
International Nuclear Information System (INIS)
Hinshelwood, D.D.; Boller, J.R.; Cooperstein, G.
1996-01-01
Transport of light ion beams through low-pressure background gas is under investigation at NRL in support of the light-ion ICF program at Sandia National Laboratories. Scaling experiments and the field solver/orbit code ATHETA have been used to design and construct a focusing, extraction applied-B diode for transport experiments. An active anode source has been developed to provide a high proton fraction in the ion beam and a fast ion turn-on time. A very sensitive Zeeman diagnostic is being developed to determine the net current distribution in the beam/transport system. Both analytical and numerical techniques using several codes are being applied to transport modeling, leading to the capability of full system studies. (author). 1 tab., 5 figs., 10 refs
Code of Conduct on the Safety of Research Reactors
International Nuclear Information System (INIS)
2006-09-01
The Board of Governors of the International Atomic Energy Agency (IAEA) adopted the Code of Conduct on the Safety of Research Reactors on 8 March 2004. The Board's action was the culmination of several years of work to develop the Code and obtain a consensus on its provisions. The process leading to the Code began in 1998, when the International Nuclear Safety Advisory Group (INSAG) informed the Director General of concerns about the safety of research reactors. In 2000, INSAG recommended that the Secretariat begin developing an international protocol or a similar legal instrument to address those concerns. In September 2000, in resolution GC(44)/RES/14, the General Conference requested the Secretariat ''within its available resources, to continue work on exploring options to strengthen the international nuclear safety arrangements for civil research reactors, taking due account of input from INSAG and the views of other relevant bodies''. A working group convened by the Secretariat pursuant to that request recommended that ''the Agency consider establishing an international action plan for research reactors'' and that the action plan include preparation of a Code of Conduct ''that would clearly establish the desirable attributes for management of research reactor safety''. In September 2001, the Board requested that the Secretariat develop and implement, in conjunction with Member States, an international research reactor safety enhancement plan which included preparation of a Code of Conduct on the Safety of Research Reactors. Subsequently, in resolution GC(45)/RES/10.A, the General Conference endorsed the Board's request. Pursuant to that request, a Code of Conduct on the Safety of Research Reactors was drafted at two meetings of an Open-ended Working Group of Legal and Technical Experts. This draft Code of Conduct was circulated to all Member States for comment. On the basis of the responses received, a revised draft of the Code was prepared by the Secretariat
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
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.)
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.
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)
Solute carrier transporters: Pharmacogenomics research ...
African Journals Online (AJOL)
Aghogho
2010-12-27
-binding cassette) transporters, which include MDR1, a protein that pumps xenobiotics from cells, and the SLC (solute carrier) trans- porters, which take up neurotransmitters, nutrients, heavy metals, and other substrates into ...
Computer codes for problems of isotope and radiation research
International Nuclear Information System (INIS)
Remer, M.
1986-12-01
A survey is given of computer codes for problems in isotope and radiation research. Altogether 44 codes are described as titles with abstracts. 17 of them are in the INIS scope and are processed individually. The subjects are indicated in the chapter headings: 1) analysis of tracer experiments, 2) spectrum calculations, 3) calculations of ion and electron trajectories, 4) evaluation of gamma irradiation plants, and 5) general software
Utilization of MCNP code in the research and design for China advanced research reactor
International Nuclear Information System (INIS)
Shen Feng
2006-01-01
MCNP, which is the internationalized neutronics code, is used for nuclear research and design in China Advanced Research Reactor (CARR). MCNP is an important neutronics code in the research and design for CARR since many calculation tasks could be undertaken by it. Many nuclear parameters on reactor core, the design and optimization research for many reactor utilizations, much verification for other nuclear calculation code and so on are conducted with help of MCNP. (author)
Solute carrier transporters: Pharmacogenomics research ...
African Journals Online (AJOL)
This paper reviews the solute carrier transporters and highlights the fact that there is much to be learnt from characterizing human genomic variation in South Africa and sub-Saharan Africa, especially with regards to health applications. Genomic diversity in this region is indeed relatively under-studied despite being home to ...
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.
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)
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
2010 Transportation Research Board Environment and Energy Research Conference
2010-05-01
The Transportation Research Boards (TRB) 2010 Environment and Energy Workshop: Better Delivery of Better Solutions, which will be held June 6-10, 2010 in Raleigh, North Carolina, will commence with a session to discuss research needs in the worksh...
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).
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
Guidelines for an environmental code of ethics for research institutions
International Nuclear Information System (INIS)
Gardusi, Claudia; Aquino, Afonso Rodrigues de
2009-01-01
The purpose of this work is to reflect about actions that may contribute to the creation of mechanisms to protect the environment in the development of research projects at research institutions, specifically the Nuclear and Energy Research Institute - IPEN. A brief review of part of the ethical values applied to the process of scientific development during the old, medieval and modern periods is presented, showing the split of the nature ethical principles. It is also reported an overview of the creation of codes of ethics applied to research institutions. Moreover, criteria are presented to settle guidelines to protect the environment during the development of research projects. (author)
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
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
Transportation Research & Analysis Computing Center
Federal Laboratory Consortium — The technical objectives of the TRACC project included the establishment of a high performance computing center for use by USDOT research teams, including those from...
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.
Grounded Theorising Applied to IS Research - Developing a Coding Strategy
Directory of Open Access Journals (Sweden)
Bruce Rowlands
2005-05-01
Full Text Available This paper provides an example of developing a coding strategy to build theory of the roles of methods in IS development. The research seeks to identify and understand how system development methods are used in an IS department within a large Australian bank. The paper details a theoretical framework, particulars of data collection, and documents an early phase of analysis – data reduction and the generation of an initial coding scheme. Guided by a framework to study the use of methods, the analysis demonstrates the framework’s plausibility in order to develop theoretical relationships with which to develop a grounded theory.
The MIMIC Code Repository: enabling reproducibility in critical care research.
Johnson, Alistair Ew; Stone, David J; Celi, Leo A; Pollard, Tom J
2018-01-01
Lack of reproducibility in medical studies is a barrier to the generation of a robust knowledge base to support clinical decision-making. In this paper we outline the Medical Information Mart for Intensive Care (MIMIC) Code Repository, a centralized code base for generating reproducible studies on an openly available critical care dataset. Code is provided to load the data into a relational structure, create extractions of the data, and reproduce entire analysis plans including research studies. Concepts extracted include severity of illness scores, comorbid status, administrative definitions of sepsis, physiologic criteria for sepsis, organ failure scores, treatment administration, and more. Executable documents are used for tutorials and reproduce published studies end-to-end, providing a template for future researchers to replicate. The repository's issue tracker enables community discussion about the data and concepts, allowing users to collaboratively improve the resource. The centralized repository provides a platform for users of the data to interact directly with the data generators, facilitating greater understanding of the data. It also provides a location for the community to collaborate on necessary concepts for research progress and share them with a larger audience. Consistent application of the same code for underlying concepts is a key step in ensuring that research studies on the MIMIC database are comparable and reproducible. By providing open source code alongside the freely accessible MIMIC-III database, we enable end-to-end reproducible analysis of electronic health records. © The Author 2017. Published by Oxford University Press on behalf of the American Medical Informatics Association.
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.
Ogunrin, Olubunmi A; Daniel, Folasade; Ansa, Victor
2016-12-01
Responsibility for protection of research participants from harm and exploitation rests on Research Ethics Committees and principal investigators. The Nigerian National Code of Health Research Ethics defines responsibilities of stakeholders in research so its knowledge among researchers will likely aid ethical conduct of research. The levels of awareness and knowledge of the Code among biomedical researchers in southern Nigerian research institutions was assessed. Four institutions were selected using a stratified random sampling technique. Research participants were selected by purposive sampling and completed a pre-tested structured questionnaire. A total of 102 biomedical researchers completed the questionnaires. Thirty percent of the participants were aware of the National Code though 64% had attended at least one training seminar in research ethics. Twenty-five percent had a fairly acceptable knowledge (scores 50%-74%) and 10% had excellent knowledge of the code (score ≥75%). Ninety-five percent expressed intentions to learn more about the National Code and agreed that it is highly relevant to the ethical conduct of research. Awareness and knowledge of the Code were found to be very limited among biomedical researchers in southern Nigeria. There is need to improve awareness and knowledge through ethics seminars and training. Use of existing Nigeria-specific online training resources is also encouraged.
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.
Directory of Open Access Journals (Sweden)
David A Springate
Full Text Available Lists of clinical codes are the foundation for research undertaken using electronic medical records (EMRs. If clinical code lists are not available, reviewers are unable to determine the validity of research, full study replication is impossible, researchers are unable to make effective comparisons between studies, and the construction of new code lists is subject to much duplication of effort. Despite this, the publication of clinical codes is rarely if ever a requirement for obtaining grants, validating protocols, or publishing research. In a representative sample of 450 EMR primary research articles indexed on PubMed, we found that only 19 (5.1% were accompanied by a full set of published clinical codes and 32 (8.6% stated that code lists were available on request. To help address these problems, we have built an online repository where researchers using EMRs can upload and download lists of clinical codes. The repository will enable clinical researchers to better validate EMR studies, build on previous code lists and compare disease definitions across studies. It will also assist health informaticians in replicating database studies, tracking changes in disease definitions or clinical coding practice through time and sharing clinical code information across platforms and data sources as research objects.
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)
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
New developments in transportation for research reactors
Energy Technology Data Exchange (ETDEWEB)
Mondanel, J.L. [Transnucleaire, F-75008 Paris (France)
1998-07-01
For more than 30 years, Transnucleaire has been performing safely a large number of national and international transports of radioactive material. Transnucleaire has also designed and supplied numerous packagings for all types of nuclear fuel cycle radioactive materials: for front-end and back-end products and for power and research reactors. Since the last meeting held in Bruges, Transnucleaire has been continuously involved in transportation activities for fresh and irradiated materials for research reactors. We are pleased to take the opportunity in this meeting to share with reactor operators, official bodies and other partners, the on-going developments in transportation and associated services. Special attention will be paid to the starting of transports of MTR spent fuel elements to the La Hague reprocessing plant where COGEMA offers reprocessing services on a long-term basis to reactors operators. Detailed information is provided on regulatory issues, which may affect transport activities: evolution of the regulations, real experiences of recent transportation and development of new packaging designs. Options and solutions will be proposed by Transnucleaire to improve the situation for continuation of national and international transports at an acceptable price whilst maintaining an ultimate level of safety (author)
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...
Transport safety research abstracts. No. 1
International Nuclear Information System (INIS)
1991-07-01
The Transport Safety Research Abstracts is a collection of reports from Member States of the International Atomic Energy Agency, and other international organizations on research in progress or just completed in the area of safe transport of radioactive material. The main aim of TSRA is to draw attention to work that is about to be published, thus enabling interested parties to obtain further information through direct correspondence with the investigators. Information contained in this issue covers work being undertaken in 6 Member States and contracted by 1 international organization; it is hoped with succeeding issues that TSRA will be able to widen this base. TSRA is modelled after other IAEA publications describing work in progress in other programme areas, namely Health Physics Research Abstracts (No. 14 was published in 1989), Waste Management Research Abstracts (No. 20 was published in 1990), and Nuclear Safety Research Abstracts (No. 2 was published in 1990)
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)
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
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
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.
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)
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...
Li, Chuan-qi; Yang, Meng-jie; Luo, De-jun; Lu, Ye; Kong, Yi-pu; Zhang, Dong-chuang
2014-09-01
A new kind of variable-length codes with good correlation properties for the multirate asynchronous optical code division multiple access (OCDMA) multimedia networks is proposed, called non-repetition interval (NRI) codes. The NRI codes can be constructed by structuring the interval-sets with no repetition, and the code length depends on the number of users and the code weight. According to the structural characteristics of NRI codes, the formula of bit error rate (BER) is derived. Compared with other variable-length codes, the NRI codes have lower BER. A multirate OCDMA multimedia simulation system is designed and built, the longer codes are assigned to the users who need slow speed, while the shorter codes are assigned to the users who need high speed. It can be obtained by analyzing the eye diagram that the user with slower speed has lower BER, and the conclusion is the same as the actual demand in multimedia data transport.
Surface transportation : research funding, federal role, and emerging issues
1996-09-01
Report provides information on the public and private funding for surface : transportation research, the transportation community's views on the federal : role for such research and the Department of Transportation's ability to fulfill : that role, a...
Radiation protection code of practice in academic and research institutes
International Nuclear Information System (INIS)
Abdalla, A. A. M.
2010-05-01
The main aim of this study was to establish a code of practice on radiation protection for safe control of radiation sources used in academic and research institutes, another aim of this study was to assess the current situation of radiation protection in some of the academic and research institutes.To achieve the aims of this study, a draft of a code of practice has been developed which is based on international and local relevant recommendation. The developed code includes the following main issues: regulatory responsibilities, radiation protection program and design of radiation installations. The second aim had been accomplished by conducting inspection visits to five (A, B, C, D and E) academic and to four (F, G, H and I ) research institutes. Eight of such institutes are located in Khartoum State and the ninth one is in Madani city (Aljazeera State). The inspection activities have been carried out using a standard inspection check list developed by the regulatory authority of the Sudan. The inspection missions to the above mentioned institutes involved also evaluation of radiation levels around the premises and storage areas of radiation sources. The dose rate measurement around radiation sources locations were found to be quite low. This mainly is due to the fact that the activities of most radionuclides that are used in these institutes are quite low ( in the range of micro curies). Also ,most the x-ray machines that were found in use for scientific academic and research purposes work at low k Vp of maximum 60 k Vp. None of the radiation workers in the inspected institutes has a personal radiation monitoring device, therefor staff dose levels have not been assessed. However it was noted that in most of the academic/ research studies radiation workers are only exposed to very low levels of radiation and for a very short time that dose not exceed 1 minute, therefore the expected occupational exposure of the staff is very low. Radiation measurement in public
Underworld - Bringing a Research Code to the Classroom
Moresi, L. N.; Mansour, J.; Giordani, J.; Farrington, R.; Kaluza, O.; Quenette, S.; Woodcock, R.; Squire, G.
2017-12-01
While there are many reasons to celebrate the passing of punch card programming and flickering green screens,the loss of the sense of wonder at the very existence of computers and the calculations they make possible shouldnot be numbered among them. Computers have become so familiar that students are often unaware that formal and careful design of algorithms andtheir implementations remains a valuable and important skill that has to be learned and practiced to achieveexpertise and genuine understanding. In teaching geodynamics and geophysics at undergraduate level, we aimed to be able to bring our researchtools into the classroom - even when those tools are advanced, parallel research codes that we typically deploy on hundredsor thousands of processors, and we wanted to teach not just the physical concepts that are modelled by these codes but asense of familiarity with computational modelling and the ability to discriminate a reliable model from a poor one. The underworld code (www.underworldcode.org) was developed for modelling plate-scale fluid mechanics and studyingproblems in lithosphere dynamics. Though specialised for this task, underworld has a straightforwardpython user interface that allows it to run within the environment of jupyter notebooks on a laptop (at modest resolution, of course).The python interface was developed for adaptability in addressing new research problems, but also lends itself to integration intoa python-driven learning environment. To manage the heavy demands of installing and running underworld in a teaching laboratory, we have developed a workflow in whichwe install docker containers in the cloud which support a number of students to run their own environment independently. We share ourexperience blending notebooks and static webpages into a single web environment, and we explain how we designed our graphics andanalysis tools to allow notebook "scripts" to be queued and run on a supercomputer.
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
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)
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
Directory of Open Access Journals (Sweden)
Dubravka Komić
Full Text Available Professional codes of ethics are social contracts among members of a professional group, which aim to instigate, encourage and nurture ethical behaviour and prevent professional misconduct, including research and publication. Despite the existence of codes of ethics, research misconduct remains a serious problem. A survey of codes of ethics from 795 professional organizations from the Illinois Institute of Technology's Codes of Ethics Collection showed that 182 of them (23% used research integrity and research ethics terminology in their codes, with differences across disciplines: while the terminology was common in professional organizations in social sciences (82%, mental health (71%, sciences (61%, other organizations had no statements (construction trades, fraternal social organizations, real estate or a few of them (management, media, engineering. A subsample of 158 professional organizations we judged to be directly involved in research significantly more often had statements on research integrity/ethics terminology than the whole sample: an average of 10.4% of organizations with a statement (95% CI = 10.4-23-5% on any of the 27 research integrity/ethics terms compared to 3.3% (95% CI = 2.1-4.6%, respectively (P<0.001. Overall, 62% of all statements addressing research integrity/ethics concepts used prescriptive language in describing the standard of practice. Professional organizations should define research integrity and research ethics issues in their ethics codes and collaborate within and across disciplines to adequately address responsible conduct of research and meet contemporary needs of their communities.
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)
Komić, Dubravka; Marušić, Stjepan Ljudevit; Marušić, Ana
2015-01-01
Professional codes of ethics are social contracts among members of a professional group, which aim to instigate, encourage and nurture ethical behaviour and prevent professional misconduct, including research and publication. Despite the existence of codes of ethics, research misconduct remains a serious problem. A survey of codes of ethics from 795 professional organizations from the Illinois Institute of Technology's Codes of Ethics Collection showed that 182 of them (23%) used research integrity and research ethics terminology in their codes, with differences across disciplines: while the terminology was common in professional organizations in social sciences (82%), mental health (71%), sciences (61%), other organizations had no statements (construction trades, fraternal social organizations, real estate) or a few of them (management, media, engineering). A subsample of 158 professional organizations we judged to be directly involved in research significantly more often had statements on research integrity/ethics terminology than the whole sample: an average of 10.4% of organizations with a statement (95% CI = 10.4-23-5%) on any of the 27 research integrity/ethics terms compared to 3.3% (95% CI = 2.1-4.6%), respectively (Presearch integrity/ethics concepts used prescriptive language in describing the standard of practice. Professional organizations should define research integrity and research ethics issues in their ethics codes and collaborate within and across disciplines to adequately address responsible conduct of research and meet contemporary needs of their communities.
Komić, Dubravka; Marušić, Stjepan Ljudevit; Marušić, Ana
2015-01-01
Professional codes of ethics are social contracts among members of a professional group, which aim to instigate, encourage and nurture ethical behaviour and prevent professional misconduct, including research and publication. Despite the existence of codes of ethics, research misconduct remains a serious problem. A survey of codes of ethics from 795 professional organizations from the Illinois Institute of Technology’s Codes of Ethics Collection showed that 182 of them (23%) used research integrity and research ethics terminology in their codes, with differences across disciplines: while the terminology was common in professional organizations in social sciences (82%), mental health (71%), sciences (61%), other organizations had no statements (construction trades, fraternal social organizations, real estate) or a few of them (management, media, engineering). A subsample of 158 professional organizations we judged to be directly involved in research significantly more often had statements on research integrity/ethics terminology than the whole sample: an average of 10.4% of organizations with a statement (95% CI = 10.4-23-5%) on any of the 27 research integrity/ethics terms compared to 3.3% (95% CI = 2.1–4.6%), respectively (Pethics concepts used prescriptive language in describing the standard of practice. Professional organizations should define research integrity and research ethics issues in their ethics codes and collaborate within and across disciplines to adequately address responsible conduct of research and meet contemporary needs of their communities. PMID:26192805
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
Reactivity feedback coefficients Pakistan research reactor-1 using PRIDE code
Energy Technology Data Exchange (ETDEWEB)
Mansoor, Ali; Ahmed, Siraj-ul-Islam; Khan, Rustam [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Inam-ul-Haq [Comsats Institute of Information Technology, Islamabad (Pakistan). Dept. of Physics
2017-05-15
Results of the analyses performed for fuel, moderator and void's temperature feedback reactivity coefficients for the first high power core configuration of Pakistan Research Reactor - 1 (PARR-1) are summarized. For this purpose, a validated three dimensional model of PARR-1 core was developed and confirmed against the reference results for reactivity calculations. The ''Program for Reactor In-Core Analysis using Diffusion Equation'' (PRIDE) code was used for development of global (3-dimensional) model in conjunction with WIMSD4 for lattice cell modeling. Values for isothermal fuel, moderator and void's temperature feedback reactivity coefficients have been calculated. Additionally, flux profiles for the five energy groups were also generated.
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
Research on the improvement of nuclear safety -Development of computing code system for level 3 PSA
International Nuclear Information System (INIS)
Jeong, Jong Tae; Kim, Dong Ha; Park, Won Seok; Hwang, Mi Jeong
1995-07-01
Among the various research areas of the level 3 PSA, the effect of terrain on the transport of radioactive material was investigated. These results will give a physical insight in the development of a new dispersion model. A wind tunnel experiment with bell shaped hill model was made in order to develop a new dispersion model. And an improved dispersion model was developed based on the concentration distribution data obtained from the wind tunnel experiment. This model will be added as an option to the atmospheric dispersion code. A stand-alone atmospheric code using MS Visual Basic programming language which runs at the Windows environment of a PC was developed. A user can easily select a necessary data file and type input data by clicking menus, and can select calculation options such building wake, plume rise etc., if necessary. And a user can easily understand the meaning of concentration distribution on the map around the plant site as well as output files. Also the methodologies for the estimation of radiation exposure and for the calculation of risks was established. These methodologies will be used for the development of modules for the radiation exposure and risks respectively. These modules will be developed independently and finally will be combined to the atmospheric dispersion code in order to develop a level 3 PSA code. 30 tabs., 56 figs., refs. (Author)
Research on the improvement of nuclear safety -The development of a severe accident analysis code-
Energy Technology Data Exchange (ETDEWEB)
Kim, Heui Dong; Cho, Sung Won; Park, Jong Hwa; Hong, Sung Wan; Yoo, Dong Han; Hwang, Moon Kyoo; Noh, Kee Man; Song, Yong Man [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1995-07-01
For prevention and mitigation of the containment failure during severe accident, the study is focused on the severe accident phenomena, especially, the ones occurring inside the cavity and is intended to improve existing models and develop analytical tools for the assessment of severe accidents. A correlation equation of the flame velocity of pre mixture gas of H{sub 2}/air/steam has been suggested and combustion flame characteristic was analyzed using a developed computer code. For the analysis of the expansion phase of vapor explosion, the mechanical model has been developed. The development of a debris entrainment model in a reactor cavity with captured volume has been continued to review and examine the limitation and deficiencies of the existing models. Pre-test calculation was performed to support the severe accident experiment for molten corium concrete interaction study and the crust formation process and heat transfer characteristics of the crust have been carried out. A stress analysis code was developed using finite element method for the reactor vessel lower head failure analysis. Through international program of PHEBUS-FP and participation in the software development, the research on the core degradation process and fission products release and transportation are undergoing. CONTAIN and MELCOR codes were continuously updated under the cooperation with USNRC and French developed computer codes such as ICARE2, ESCADRE, SOPHAEROS were also installed into the SUN workstation. 204 figs, 61 tabs, 87 refs. (Author).
Research on the improvement of nuclear safety -Development of computing code system for level 3 PSA
Energy Technology Data Exchange (ETDEWEB)
Jeong, Jong Tae; Kim, Dong Ha; Park, Won Seok; Hwang, Mi Jeong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1995-07-01
Among the various research areas of the level 3 PSA, the effect of terrain on the transport of radioactive material was investigated. These results will give a physical insight in the development of a new dispersion model. A wind tunnel experiment with bell shaped hill model was made in order to develop a new dispersion model. And an improved dispersion model was developed based on the concentration distribution data obtained from the wind tunnel experiment. This model will be added as an option to the atmospheric dispersion code. A stand-alone atmospheric code using MS Visual Basic programming language which runs at the Windows environment of a PC was developed. A user can easily select a necessary data file and type input data by clicking menus, and can select calculation options such building wake, plume rise etc., if necessary. And a user can easily understand the meaning of concentration distribution on the map around the plant site as well as output files. Also the methodologies for the estimation of radiation exposure and for the calculation of risks was established. These methodologies will be used for the development of modules for the radiation exposure and risks respectively. These modules will be developed independently and finally will be combined to the atmospheric dispersion code in order to develop a level 3 PSA code. 30 tabs., 56 figs., refs. (Author).
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.
Research and Design in Unified Coding Architecture for Smart Grids
Directory of Open Access Journals (Sweden)
Gang Han
2013-09-01
Full Text Available Standardized and sharing information platform is the foundation of the Smart Grids. In order to improve the dispatching center information integration of the power grids and achieve efficient data exchange, sharing and interoperability, a unified coding architecture is proposed. The architecture includes coding management layer, coding generation layer, information models layer and application system layer. Hierarchical design makes the whole coding architecture to adapt to different application environments, different interfaces, loosely coupled requirements, which can realize the integration model management function of the power grids. The life cycle and evaluation method of survival of unified coding architecture is proposed. It can ensure the stability and availability of the coding architecture. Finally, the development direction of coding technology of the Smart Grids in future is prospected.
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.
BURNUR.SYS: A 2-D code system for NUR research reactor burn up analysis
International Nuclear Information System (INIS)
Meftah, B.; Halilou, A.; Letaim, F.; Mazidi, S.; Mokeddem, M.Y.; Zeggar, F.
2008-01-01
Adequate knowledge of burn up levels of fuel elements within a research reactor is of great importance for its optimum operation. Such knowledge is required for the monitoring of reactivity parameters and flux and power distributions throughout the reactor core, the estimation of the radioactive source term needed in accidental situations analysis, the evaluation of the amount of fissile materials present at any moment within the fuel for safeguards purposes and the estimation of cooling and shielding requirements for interim storage or transport of spent fuel elements. This paper presents the approach of fuel burn up evaluation used at the NUR research reactor. The approach is essentially based upon the utilization of BURNUR.SYS code, an in-house developed software. BURNUR.SYS is an object oriented program under DELPHI 7 that integrates the cell calculation code WIMSD-4 and the core calculation code CITVAP. BURNUR.SYS calculates the evolution in time of pertinent quantities such as: the concentrations of U235 and others actinides, the concentrations of major poisons (Xe135 and Sm149), the distributions of power densities and burn up levels within fuel elements, the effective multiplication factor and core reactivity. The results are displayed in user friendly graphical and numerical formats
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)
Kodo: An Open and Research Oriented Network Coding Library
DEFF Research Database (Denmark)
Pedersen, Morten Videbæk; Heide, Janus; Fitzek, Frank
2011-01-01
We consider the problem of efficient decoding of a random linear code over a finite field. In particular we are interested in the case where the code is random, relatively sparse, and use the binary finite field as an example. The goal is to decode the data using fewer operations to potentially a...
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.
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.
Consensus Coding as a Tool in Visual Appearance Research
Directory of Open Access Journals (Sweden)
D R Simmons
2011-04-01
Full Text Available A common problem in visual appearance research is how to quantitatively characterise the visual appearance of a region of an image which is categorised by human observers in the same way. An example of this is scarring in medical images (Ayoub et al, 2010, The Cleft-Palate Craniofacial Journal, in press. We have argued that “scarriness” is itself a visual appearance descriptor which summarises the distinctive combination of colour, texture and shape information which allows us to distinguish scarred from non-scarred tissue (Simmons et al, ECVP 2009. Other potential descriptors for other image classes would be “metallic”, “natural”, or “liquid”. Having developed an automatic algorithm to locate scars in medical images, we then tested “ground truth” by asking untrained observers to draw around the region of scarring. The shape and size of the scar on the image was defined by building a contour plot of the agreement between observers' outlines and thresholding at the point above which 50% of the observers agreed: a consensus coding scheme. Based on the variability in the amount of overlap between the scar as defined by the algorithm, and the consensus scar of the observers, we have concluded that the algorithm does not completely capture the putative appearance descriptor “scarriness”. A simultaneous analysis of qualitative descriptions of the scarring by the observers revealed that other image features than those encoded by the algorithm (colour and texture might be important, such as scar boundary shape. This approach to visual appearance research in medical imaging has potential applications in other application areas, such as botany, geology and archaeology.
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)
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
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
Research and Development Program for transportation packagings at Sandia National Laboratories
International Nuclear Information System (INIS)
Hohnstreiter, G.F.; Sorenson, K.B.
1995-01-01
This document contains information about the research and development programs dealing with waste transport at Sandia National Laboratories. This paper discusses topics such as: Why new packaging is needed; analytical methodologies and design codes;evaluation of packaging components; materials characterization; creative packaging concepts; packaging engineering and analysis; testing; and certification support
Administrative mechanics of research fuel transportation
International Nuclear Information System (INIS)
Harmon, Diane W.
1983-01-01
This presentation contains the discussion on the multitude of administrative mechanics that have to be meshed for the successful completion of a shipment of spent fuel, HEU or LEU in the research reactors fuel cycle. The costs associated with transportation may be the equivalent of 'a black hole', so an overview of cost factors is given. At the end one could find that this black hole factor in the budget is actually a bargain. The first step is the quotation phase. The cost variables in the quotation contain the cost of packaging i.e. containers; the complete routing of the packages and the materials. Factors that are of outmost importance are the routing restrictions and regulations, physical security regulations. All of this effort is just to provide a valid quotation not to accomplish the goal of completing a shipment. Public relations cannot be omitted either
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.
Geisler, Cheryl
2018-01-01
Coding, the analytic task of assigning codes to nonnumeric data, is foundational to writing research. A rich discussion of methodological pluralism has established the foundational importance of systematicity in the task of coding, but less attention has been paid to the equally important commitment to language complexity. Addressing the interplay…
DeLyser, Dydia; Potter, Amy E.
2013-01-01
This article describes experiential-learning approaches to conveying the work and rewards involved in qualitative research. Seminar students interviewed one another, transcribed or took notes on those interviews, shared those materials to create a set of empirical materials for coding, developed coding schemes, and coded the materials using those…
Transportation Research – Safety and Sustainability
Indian Academy of Sciences (India)
However, many recent reports suggest that improvements in public transport and promotion of ... Most efforts to reduce environmental pollution due to road transport, therefore, focus on the control of exhaust ... Whenever sustainable transport issues are raised, discussions centre around vehicle emis- sions and pollution ...
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)
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
Validation of the TAC/BLOOST code (Contract research)
International Nuclear Information System (INIS)
Takamatsu, Kuniyoshi; Nakagawa, Shigeaki
2005-06-01
Safety demonstration tests using the High Temperature engineering Test Reactor (HTTR) are in progress to verify the inherent safety features for High Temperature Gas-cooled Reactors (HTGRs). The coolant flow reduction test by tripping gas circulators is one of the safety demonstration tests. The reactor power safely brings to a stable level without a reactor scram and the temperature transient of the reactor-core is very slow. The TAC/BLOOST code was developed to analyze reactor and temperature transient during the coolant flow reduction test taking account of reactor dynamics. This paper describes the validation result of the TAC/BLOOST code with the measured values of gas circulators tripping tests at 30% (9 MW). It was confirmed that the TAC/BLOOST code was able to analyze the reactor transient during the test. (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.)
Education in Transportation Systems Planning: Highway Research Record No. 462.
National Academy of Sciences - National Research Council, Washington, DC. Transportation Research Board.
The papers contained in the issue of Highway Research Record focus on current and emerging patterns of education and training related to transportation systems planning. The five papers are: Transportation Centers and Other Mechanisms to Encourage Interdisciplinary Research and Training Efforts in Transportation (Frederick J. Wegmann and Edward A.…
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)
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
TRANSPORTATION RESEARCH CONTRIBUTIONS TO SOCIETY BY UNIVERSITY TRANSPORTATION CENTERS
Directory of Open Access Journals (Sweden)
Robert C. JOHNS
2003-01-01
Full Text Available This paper discusses the importance of knowledge in the global economy and reviews the process in which knowledge is applied to develop innovations. It confirms the importance of innovation as a key factor for success in today's competitive environment. The paper discusses the contributions a university can make to the innovation process in the field of transportation, and offers a vision of how a university center can enhance and facilitate these contributions. It then describes the efforts of one center, including three examples of innovations facilitated by the center in traffic detection, regional planning, and pavement management. The paper concludes with suggestions that would strengthen the societal contributions of university transportation centers.
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
International Nuclear Information System (INIS)
Kim, Jung Do; Gil, Choong Sub; Lee, Jong Tai; Hwang, Won Guk
1992-01-01
A one-group cross section data base of the ORIGEN2 computer code was developed for research reactor applications. For this, ENDF/B-IV and -V data were processed using the NJOY code system into 69-group data. The burnup dependent weighting spectra for KMRR were calculated with the WIMS-KAERI computer code, and then the 69-group data were collapsed to one-group using the spectra. The ORlGEN2-predicted burnup-dependent actinide compositions of KMRR spent fuel using the newly developed data base show a good agreement with the results of detailed multigroup transport calculation. In addition, the burnup characteristics of KMRR spent fuel was analyzed with the new data base. (Author)
Market research for Idaho Transportation Department linear referencing system.
2009-09-02
For over 30 years, the Idaho Transportation Department (ITD) has had an LRS called MACS : (MilePoint And Coded Segment), which is being implemented on a mainframe using a : COBOL/CICS platform. As ITD began embracing newer technologies and moving tow...
ORPHEE research reactor: 3D core depletion calculation using Monte-Carlo code TRIPOLI-4®
Damian, F.; Brun, E.
2014-06-01
ORPHEE is a research reactor located at CEA Saclay. It aims at producing neutron beams for experiments. This is a pool-type reactor (heavy water), and the core is cooled by light water. Its thermal power is 14 MW. ORPHEE core is 90 cm height and has a cross section of 27x27 cm2. It is loaded with eight fuel assemblies characterized by a various number of fuel plates. The fuel plate is composed of aluminium and High Enriched Uranium (HEU). It is a once through core with a fuel cycle length of approximately 100 Equivalent Full Power Days (EFPD) and with a maximum burnup of 40%. Various analyses under progress at CEA concern the determination of the core neutronic parameters during irradiation. Taking into consideration the geometrical complexity of the core and the quasi absence of thermal feedback for nominal operation, the 3D core depletion calculations are performed using the Monte-Carlo code TRIPOLI-4® [1,2,3]. A preliminary validation of the depletion calculation was performed on a 2D core configuration by comparison with the deterministic transport code APOLLO2 [4]. The analysis showed the reliability of TRIPOLI-4® to calculate a complex core configuration using a large number of depleting regions with a high level of confidence.
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
Survey of research reports in transportation modelling. Part 1
Nijsse, A.; Wamsteker-Andriessen, S.J.
1993-01-01
A survey of research reports in transportation modelling in two parts. Part one is devided in reports concerning economic development and car mobility, analyzing large transportation data files and transportation planning and spatial development. Part two consists of reserach reports concerning
Survey of research reports in transportation modelling. Part 2
Nijsse, A.; Wamsteker-Andriessen, S.J.
1993-01-01
A survey of research reports in transportation modelling in two parts. Part one is devided in reports concerning economic development and car mobility, analyzing large transportation data files and transportation planning and spatial development. Part two consists of reserach reports concerning
Intelligent transportation system (ITS) international research exchange.
2014-01-01
ITS applications address surface transportation challenges in safety, mobility, and : sustainability that are similar in cause and impact worldwide. International ITS : exchange allows cooperating nations to benefit from each others pre-competitiv...
IAEA Code of Conduct on the Safety of Research Reactors and Suggestions for Effective Application
International Nuclear Information System (INIS)
Kim, W. S.; Choi, Y. S.; Choi, K. S.; Shin, D. S.
2006-01-01
In 1998, the International Nuclear Safety Advisory Group (INSAG) raised concerns about research reactors, especially those neither operating nor decommissioned (extended shutdown) in developing countries and recommended that the IAEA develop an international protocol or similar legal instrument to address these concerns. The board of IAEA requested the agency to develop and implement an international research reactor enhancement plan including preparation of a Code of Conduct on the Safety of Research Reactors. After holding two open-ended meetings to develop a draft code and circulating it to all Member States, the Code was adopted by the Board of IAEA in March 2004. This paper presents what the Code of Conduct is and what the Member States have to do. In addition, several suggestions are identified for effectively applying the Code of Conducts to domestic research reactors
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.
Systemization of burnup sensitivity analysis code (2) (Contract research)
International Nuclear Information System (INIS)
Tatsumi, Masahiro; Hyoudou, Hideaki
2008-08-01
Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant economic efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristic is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons: the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion
International Nuclear Information System (INIS)
Goko, Shinji
2017-01-01
As for the safety analysis to be carried out when a nuclear power company applies for installation permission of facility or equipment, business license, design approval etc., the Regulatory Standard and Research Department Secretariat of Nuclear Regulation Authority continuously conducts safety research for the introduction of various technologies and their improvement in order to evaluate the adequacy of this safety analysis. In the field of the shielding analysis of nuclear fuel transportation materials, this group improved the code to make PHITS applicable to this field, and has been promoting the improvement as a tool used for regulations since FY2013. This paper introduced the history and progress of this safety research. PHITS 2.88, which is the latest version as of November 2016, was equipped with the automatic generation function of variance reduction parameters [T-WWG] etc., and developed as the tool equipped with many effective functions in practical application to nuclear power regulations. In addition, this group conducted the verification analysis against nuclear fuel packages, which showed a good agreement with the analysis by MCNP, which is extensively used worldwide and abundant in actual results. It also shows a relatively good agreement with the measured values, when considering differences in analysis and measurement. (A.O.)
BNFL's experience in the sea transport of irradiated research reactor fuel to the USA
International Nuclear Information System (INIS)
Hudson, I.A.; Porter, I.
2000-01-01
BNFL provides worldwide transport for a wide range of nuclear materials. BNFL Transport manages an unique fleet of vessels, designed, built, and operated to the highest safety standards, including the highest rating within the INF Code recommended by the International Maritime Organisation. The company has some 20 years of experience of transporting irradiated research reactor fuel in support of the United States' programme for returning US obligated fuel from around the world. Between 1977 and 1988 BNFL performed 11 shipments of irradiated research reactor fuel from the Japan Atomic Energy Research Institute to the US. Since 1997, a further 3 shipments have been performed as part of an ongoing programme for Japanese research reactor operators. Where possible, shipments of fuel from European countries such as Sweden and Spain have been combined with those from Japan for delivery to the US. (author)
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)
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
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)
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
Computer codes for tasks in the fields of isotope and radiation research
International Nuclear Information System (INIS)
Friedrich, K.; Gebhardt, O.
1978-11-01
Concise descriptions of computer codes developed for solving problems in the fields of isotope and radiation research at the Zentralinstitut fuer Isotopen- und Strahlenforschung (ZfI) are compiled. In part two the structure of the ZfI program library MABIF is outlined and a complete list of all codes available is given
Moral, Cristian; de Antonio, Angelica; Ferre, Xavier; Lara, Graciela
2015-01-01
Introduction: In this article we propose a qualitative analysis tool--a coding system--that can support the formalisation of the information-seeking process in a specific field: research in computer science. Method: In order to elaborate the coding system, we have conducted a set of qualitative studies, more specifically a focus group and some…
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)
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.
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
Clinical code set engineering for reusing EHR data for research: A review.
Williams, Richard; Kontopantelis, Evangelos; Buchan, Iain; Peek, Niels
2017-06-01
The construction of reliable, reusable clinical code sets is essential when re-using Electronic Health Record (EHR) data for research. Yet code set definitions are rarely transparent and their sharing is almost non-existent. There is a lack of methodological standards for the management (construction, sharing, revision and reuse) of clinical code sets which needs to be addressed to ensure the reliability and credibility of studies which use code sets. To review methodological literature on the management of sets of clinical codes used in research on clinical databases and to provide a list of best practice recommendations for future studies and software tools. We performed an exhaustive search for methodological papers about clinical code set engineering for re-using EHR data in research. This was supplemented with papers identified by snowball sampling. In addition, a list of e-phenotyping systems was constructed by merging references from several systematic reviews on this topic, and the processes adopted by those systems for code set management was reviewed. Thirty methodological papers were reviewed. Common approaches included: creating an initial list of synonyms for the condition of interest (n=20); making use of the hierarchical nature of coding terminologies during searching (n=23); reviewing sets with clinician input (n=20); and reusing and updating an existing code set (n=20). Several open source software tools (n=3) were discovered. There is a need for software tools that enable users to easily and quickly create, revise, extend, review and share code sets and we provide a list of recommendations for their design and implementation. Research re-using EHR data could be improved through the further development, more widespread use and routine reporting of the methods by which clinical codes were selected. Copyright © 2017 The Author(s). Published by Elsevier Inc. All rights reserved.
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
Tavukcu, Tahir
2016-01-01
In this research, it is aimed to determine the effect of the attitudes of postgraduate students towards scientific research and codes of conduct, supported by digital script. This research is a quantitative study, and it has been formed according to pre-test & post-test research model of experiment and control group. In both groups, lessons…
THE CONTRIBUTION OF PUBLIC INTEREST RESEARCH TO TRANSPORTATION POLICY
Directory of Open Access Journals (Sweden)
Makoto ITOH
2003-01-01
Full Text Available Established in 1995 with the basic philosophy of serving as a bridge between research and practice, the Institute for Transport Policy Studies conducts activities in support of transportation policy research in the public interest. This paper aims to describe the contribution of public interest research to transportation policy as seen in the Institute's activities. Touching first on the context and events leading to its establishment, the paper then describes the Institute's guiding principles, organization and staff and summarizes research and other activities.
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)
Transportable Heavy Duty Emissions Testing Laboratory and Research Program
Energy Technology Data Exchange (ETDEWEB)
David Lyons
2008-03-31
The objective of this program was to quantify the emissions from heavy-duty vehicles operating on alternative fuels or advanced fuel blends, often with novel engine technology or aftertreatment. In the first year of the program West Virginia University (WVU) researchers determined that a transportable chassis dynamometer emissions measurement approach was required so that fleets of trucks and buses did not need to be ferried across the nation to a fixed facility. A Transportable Heavy-Duty Vehicle Emissions Testing Laboratory (Translab) was designed, constructed and verified. This laboratory consisted of a chassis dynamometer semi-trailer and an analytic trailer housing a full scale exhaust dilution tunnel and sampling system which mimicked closely the system described in the Code of Federal Regulations for engine certification. The Translab was first used to quantify emissions from natural gas and methanol fueled transit buses, and a second Translab unit was constructed to satisfy research demand. Subsequent emissions measurement was performed on trucks and buses using ethanol, Fischer-Tropsch fuel, and biodiesel. A medium-duty chassis dynamometer was also designed and constructed to facilitate research on delivery vehicles in the 10,000 to 20,000lb range. The Translab participated in major programs to evaluate low-sulfur diesel in conjunction with passively regenerating exhaust particulate filtration technology, and substantial reductions in particulate matter were recorded. The researchers also participated in programs to evaluate emissions from advanced natural gas engines with closed loop feedback control. These natural gas engines showed substantially reduced levels of oxides of nitrogen. For all of the trucks and buses characterized, the levels of carbon monoxide, oxides of nitrogen, hydrocarbons, carbon dioxide and particulate matter were quantified, and in many cases non-regulated species such as aldehydes were also sampled. Particle size was also
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
Femtosecond Laser System for Research on High-Speed Optical Transmultiplexing and Coding
National Research Council Canada - National Science Library
Weiner, Andrew
1997-01-01
.... This would fill an important need in both TDM packet networks and bit-parallel WDM linds. The research also aims at experimental tests of an ultrashort pulse code-division, multiple-access (CDMA...
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
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...
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)
Safe operation of research reactors and critical assemblies. Code of practice and annexes. 1984 ed
International Nuclear Information System (INIS)
1984-01-01
The safe operation of research reactors and critical assemblies (hereafter termed 'reactors') requires proper design, construction, management and supervision. This Code of Practice deals mainly with management and supervision. The provisions of the Code apply to the whole life of the reactor, including modification, updating and upgrading. The Code may be subject to revision in the light of experience and the state of technology. The Code is aimed at defining minimum requirements for the safe operation of reactors. Emphasis is placed on which safety requirements should be met rather than on specifying how these requirements may be met. The Code also provides guidance and information to persons and authorities responsible for the operation of reactors. The Code recommends that documents dealing with the operation of reactors and including safety analyses be prepared and submitted for review and approval to a regulatory body. Operation would be authorized on the understanding that it would comply with limits and conditions designed to ensure safety. The Code covers a wide range of reactor types, which gives rise to a variety of safety issues. Safety issues applicable to specific reactor types only (e.g. fast reactors) are not necessarily covered in this Code. Some of the recommendations in the Code are not directly applicable to critical assemblies. A recommendation may therefore be interpreted according to the type of reactor concerned. In such cases the words 'adequate' and 'appropriate' are used to mean 'adequate' or 'appropriate' for the type of reactor under consideration.
Compendium of computer codes for the researcher in magnetic fusion energy
Energy Technology Data Exchange (ETDEWEB)
Porter, G.D. (ed.)
1989-03-10
This is a compendium of computer codes, which are available to the fusion researcher. It is intended to be a document that permits a quick evaluation of the tools available to the experimenter who wants to both analyze his data, and compare the results of his analysis with the predictions of available theories. This document will be updated frequently to maintain its usefulness. I would appreciate receiving further information about codes not included here from anyone who has used them. The information required includes a brief description of the code (including any special features), a bibliography of the documentation available for the code and/or the underlying physics, a list of people to contact for help in running the code, instructions on how to access the code, and a description of the output from the code. Wherever possible, the code contacts should include people from each of the fusion facilities so that the novice can talk to someone ''down the hall'' when he first tries to use a code. I would also appreciate any comments about possible additions and improvements in the index. I encourage any additional criticism of this document. 137 refs.
Reclassification of ICD-9 Codes into Meaningful Categories for Oncology Survivorship Research
International Nuclear Information System (INIS)
Rassekh, S. R.; Lorenzi, M.; Lee, L.; Devji, S.; McBride, M.; Goddard, K.
2010-01-01
Background. The International Classification of Disease, ninth revision (ICD-9) is designed to code disease into categories which are placed into administrative databases. These databases have been used for epidemiological studies. However, the categories used in the ICD9-codes are not always the most effective for evaluating specific diseases or their outcomes, such as the outcomes of cancer treatment. Therefore a re-classification of the ICD-9 codes into new categories specific to cancer outcomes is needed. Methods. An expert panel comprised of two physicians created broad categories that would be most useful to researchers investigating outcomes and morbidities associated with the treatment of cancer. A Senior Data Coordinator with expertise in ICD-9 coding, then joined this panel and each code was re-classified into the new categories. Results. Consensus was achieved for the categories to go from the 17 categories in ICD-9 to 39 categories. The ICD-9 Codes were placed into new categories, and sub categories were also created for more specific outcomes. The results of this re-classification is available in tabular form. Conclusions. ICD-9 codes were re-classified by group consensus into categories that are designed for oncology survivorship research. The novel re-classification system can be used by those involved in cancer survivorship research
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.
Transportation Research – Safety and Sustainability
Indian Academy of Sciences (India)
basic mechanisms of impact dynamics involved in road safety research. Most current safety standards are based on human body and segment properties derived from impact tests done on human cadavers. In their article on finite element modelling of the human body, Mukerjee et al show that introduction of muscle forces ...
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.
Research on centrality of urban transport network nodes
Wang, Kui; Fu, Xiufen
2017-05-01
Based on the actual data of urban transport in Guangzhou, 19,150 bus stations in Guangzhou (as of 2014) are selected as nodes. Based on the theory of complex network, the network model of Guangzhou urban transport is constructed. By analyzing the degree centrality index, betweenness centrality index and closeness centrality index of nodes in the network, the level of centrality of each node in the network is studied. From a different point of view to determine the hub node of Guangzhou urban transport network, corresponding to the city's key sites and major transfer sites. The reliability of the network is determined by the stability of some key nodes (transport hub station). The research of network node centralization can provide a theoretical basis for the rational allocation of urban transport network sites and public transport system planning.
Working research codes into fluid dynamics education: a science gateway approach
Mason, Lachlan; Hetherington, James; O'Reilly, Martin; Yong, May; Jersakova, Radka; Grieve, Stuart; Perez-Suarez, David; Klapaukh, Roman; Craster, Richard V.; Matar, Omar K.
2017-11-01
Research codes are effective for illustrating complex concepts in educational fluid dynamics courses, compared to textbook examples, an interactive three-dimensional visualisation can bring a problem to life! Various barriers, however, prevent the adoption of research codes in teaching: codes are typically created for highly-specific `once-off' calculations and, as such, have no user interface and a steep learning curve. Moreover, a code may require access to high-performance computing resources that are not readily available in the classroom. This project allows academics to rapidly work research codes into their teaching via a minimalist `science gateway' framework. The gateway is a simple, yet flexible, web interface allowing students to construct and run simulations, as well as view and share their output. Behind the scenes, the common operations of job configuration, submission, monitoring and post-processing are customisable at the level of shell scripting. In this talk, we demonstrate the creation of an example teaching gateway connected to the Code BLUE fluid dynamics software. Student simulations can be run via a third-party cloud computing provider or a local high-performance cluster. EPSRC, UK, MEMPHIS program Grant (EP/K003976/1), RAEng Research Chair (OKM).
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.
A Systematic Review of Coding Systems Used in Pharmacoepidemiology and Database Research.
Chen, Yong; Zivkovic, Marko; Wang, Tongtong; Su, Su; Lee, Jianyi; Bortnichak, Edward A
2018-02-01
Clinical coding systems have been developed to translate real-world healthcare information such as prescriptions, diagnoses and procedures into standardized codes appropriate for use in large healthcare datasets. Due to the lack of information on coding system characteristics and insufficient uniformity in coding practices, there is a growing need for better understanding of coding systems and their use in pharmacoepidemiology and observational real world data research. To determine: 1) the number of available coding systems and their characteristics, 2) which pharmacoepidemiology databases are they adopted in, 3) what outcomes and exposures can be identified from each coding system, and 4) how robust they are with respect to consistency and validity in pharmacoepidemiology and observational database studies. Electronic literature database and unpublished literature searches, as well as hand searching of relevant journals were conducted to identify eligible articles discussing characteristics and applications of coding systems in use and published in the English language between 1986 and 2016. Characteristics considered included type of information captured by codes, clinical setting(s) of use, adoption by a pharmacoepidemiology database, region, and available mappings. Applications articles describing the use and validity of specific codes, code lists, or algorithms were also included. Data extraction was performed independently by two reviewers and a narrative synthesis was performed. A total of 897 unique articles and 57 coding systems were identified, 17% of which included country-specific modifications or multiple versions. Procedures (55%), diagnoses (36%), drugs (38%), and site of disease (39%) were most commonly and directly captured by these coding systems. The systems were used to capture information from the following clinical settings: inpatient (63%), ambulatory (55%), emergency department (ED, 34%), and pharmacy (13%). More than half of all coding
DOE's foreign research reactor transportation services contract: Perspective and experience
International Nuclear Information System (INIS)
Patterson, John
1997-01-01
DOE committed to low- and moderate-income countries participating in the foreign research reactor spent fuel returns program that the United States government would provide for the transportation of the spent fuel. In fulfillment of that commitment, DOE entered into transportation services contracts with qualified, private-sector firms. NAC will discuss its experience as a transportation services provider, including range of services available to the foreign reactors, advantages to DOE and to the foreign research reactors, access to contract services by high income countries and potential advantages, and experience with initial tasks performed under the contract. (author)
Initial verification and validation of RAZORBACK - A research reactor transient analysis code
Energy Technology Data Exchange (ETDEWEB)
Talley, Darren G. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)
2015-09-01
This report describes the work and results of the initial verification and validation (V&V) of the beta release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This initial V&V effort was intended to confirm that the code work to-date shows good agreement between simulation and actual ACRR operations, indicating that the subsequent V&V effort for the official release of the code will be successful.
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)
CARONTE project: Creating an Agenda for Research on Transportation Security
Energy Technology Data Exchange (ETDEWEB)
Leon Bello, J.; Gonzalez Viosca, E.
2016-07-01
Europe’s prosperity relies on effective transport systems. Any attacks and disturbances to land freight and passenger transport would have significant impact on economic growth, territorial cohesion, social development and the environment. Unfortunately, there are weaknesses in the land transport security.The objective of CARONTE project is define a future research agenda for security in land transport that focuses on core gaps caused by emerging risks while avoiding any doubling-up of research elsewhere. Its research agenda will cover all threats, including cyber-crime, and security aspects across all modes of land transportation. At the same time, it will respect the fundamental human rights and privacy of European citizens. The step-by-step method of CARONTE’s consortium has analyzed the state of the art and emerging risks; has identified gaps, analyses and assessments of potential solutions; and has produced an overall research agenda for the future. CARONTE’s results will answer the following questions among others: Which existing research projects merit a follow up and extension? Where are the combinations or synergy effects to be attended? Which themes and topics should be elaborated in new research projects? Who should be involved and integrated in future research projects (stakeholders, authorities, etc.)? The CARONTE consortium includes universities and research institutes, companies, and end-users providing with experience in research and consultancy in transportation, logistics, infrastructure management, security and communications. ITENE - Instituto Tecnológico del Embalaje, Transporte y Logística-has been one of the Project partners among a total of 11 members from eight different countries in the European Union which have also been supported via a High Level Advisory Board. (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)
Research on the coding and decoding technology of the OCDMA system
Li, Ping; Wang, Yuru; Lan, Zhenping; Wang, Jinpeng; Zou, Nianyu
2015-12-01
Optical Code Division Multiplex Access, OCDMA, is a kind of new technology which is combined the wireless CDMA technology and the optical fiber communication technology together. The address coding technology in OCDMA system has been researched. Besides, the principle of the codec based on optical fiber delay line and non-coherent spectral domain encoding and decoding has been introduced and analysed, and the results was verified by experiment.
International Nuclear Information System (INIS)
Akimoto, Hajime; Kukita; Ohnuki, Akira
1997-01-01
The Japan Atomic Energy Research Institute (JAERI) is conducting several research programs related to thermal-hydraulic and neutronic behavior of light water reactors (LWRs). These include LWR safety research projects, which are conducted in accordance with the Nuclear Safety Commission's research plan, and reactor engineering projects for the development of innovative reactor designs or core/fuel designs. Thermal-hydraulic and neutronic codes are used for various purposes including experimental analysis, nuclear power plant (NPP) safety analysis, and design assessment
ABC transporter research: going strong 40 years on.
Theodoulou, Frederica L; Kerr, Ian D
2015-10-01
In most organisms, ABC transporters constitute one of the largest families of membrane proteins. In humans, their functions are diverse and underpin numerous key physiological processes, as well as being causative factors in a number of clinically relevant pathologies. Advances in our understanding of these diseases have come about through combinations of genetic and protein biochemical investigations of these transporters and the power of in vitro and in vivo investigations is helping to develop genotype-phenotype understanding. However, the importance of ABC transporter research goes far beyond human biology; microbial ABC transporters are of great interest in terms of understanding virulence and drug resistance and industrial biotechnology researchers are exploring the potential of prokaryotic ABC exporters to increase the capacity of synthetic biology systems. Plant ABC transporters play important roles in transport of hormones, xenobiotics, metals and secondary metabolites, pathogen responses and numerous aspects of development, all of which are important in the global food security area. For 3 days in Chester, this Biochemical Society Focused Meeting brought together researchers with diverse experimental approaches and with different fundamental questions, all of which are linked by the commonality of ABC transporters. © 2015 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)
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
Human behavior research and the design of sustainable transport systems
Schauer, James J.
2011-09-01
reduced carbon emissions are central to the design and optimization of future low carbon transport systems. Gaker et al (2011) suggest a framework, and provide insight into the willingness of transport consumers to pay for emission reductions of carbon dioxide from their personal transport choices within the context of other attributes of transport variables. The results of this study, although limited to a small demographic segment of the US population, demonstrate that people can integrate information on greenhouse gas emissions with other transport attributes including cost and time. Likewise, the research shows that the study group was willing to pay for reduction in greenhouse gas emissions associated with their transport choices. The study examined auto purchase choice, transport mode choice and transport route choice, which represent key decisions associated with transport that impact greenhouse gas emissions. Interestingly, they found that the study group was willing to pay for reductions in greenhouse gas emissions at a relatively consistent price across these transport choices. Clearly, the study results may not broadly apply to all demographics of users of transport, even in the study domain, due to the small demographic segment that was examined and the fact that the study was conducted in the laboratory. However, the methods used by Gaker et al (2011) are cause for optimism that future studies can obtain much needed mapping of transport preferences and willingness to pay for greenhouse gas emission reductions associated with personal transport choices. Although the Gaker et al (2011) study is directed at understanding the promotion of low carbon transport in the context of existing infrastructures, the ability of these studies to elucidate human behavior and preferences within the trade-offs of transport are critical to the design of future transport systems that seek to meet transport demand with constrained greenhouse gas emissions. Additional studies of
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
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...
Johansson, Mats; Broström, Linus
2012-08-01
Research on incompetent humans raises ethical challenges, especially when there is no direct benefit to these research subjects. Contemporary codes of research ethics typically require that such research must specifically serve to benefit the population to which the research subjects belong. The article critically examines this "same-population condition", raising issues of both interpretation and moral justification. Of particular concern is the risk that the way in which the condition is articulated and rationalized in effect disguises or downplays the instrumentalization of incompetent individuals.
International Nuclear Information System (INIS)
Cox, J.
1971-01-01
This book is in two parts. The first is a Code of Practice for the Safe Operation of Critical Assemblies and Research Reactors, prepared as a result of a meeting of experts which took place in Vienna on 20-24 May 1968. The Code has been prepared by the International Atomic Energy Agency in co-operation with the World Health Organization, and its publication is sponsored by both organizations. In addition, the Code was approved by the Board of Governors of the International Atomic Energy Agency on 16 December 1968 as part of the Agency's safety standards, which are applied to operations undertaken by Member States with the assistance of the Agency. The Board, in approving the publication of the present book, also recommended Member States to take the Code into account in the formulation of national regulations and recommendations. The second part of the book is a Technical Appendix to give information and illustrative samples that would be helpful in implementing the Code of Practice. This second part, although published under the same cover, is not part of the Code. An extensive Bibliography, amplifying the Technical Appendix, is included at the end.
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
International Nuclear Information System (INIS)
Carvalho Filho, Carlos Alberto de; Branco, Otavio Eurico de Aquino; Loureiro, Celso de Oliveira
1996-01-01
The Engenho Nogueira Hydrogeological Project, PROHBEN, was idealized with the goal of implementing an Experimental Hydrogeological basin within its limits, in order to permit the development of hydrogeological studies and techniques, mainly in the modeling of flow and transport of contaminants (radionuclides) in the saturated and unsaturated porous media. The PROHBEN is located in Belo Horizonte, Minas Gerais, amounting a 5 km 2 area. The local porous-granular, heterogeneous and anisotropic, water-table aquifer reaches 40 meters of thickness, and is compound mainly by alluvial deposits and alteration rocks products, with a sandy texture. The flow and transport modeling are being done using the Modflow and MT3D codes. Three master degree researches are being done in the PROHBEN area and one expects is that more researchers come to use this experimental site. (author)
A code of best practice for judgement-based operational research
Wijnmalen, D.J.D.; Curtis, N.J.
2013-01-01
Judgement-based (or ‘soft’) Operational Research (OR) is used in Defence, although it may not be well known and understood and it may not be perceived to have the rigour of more quantitative techniques. A NATO task group was set up to address these features and to produce a Code of Best Practice
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
Preface: Research advances in vadose zone hydrology through simulations with the TOUGH codes
International Nuclear Information System (INIS)
Finsterle, Stefan; Oldenburg, Curtis M.
2004-01-01
Numerical simulators are playing an increasingly important role in advancing our fundamental understanding of hydrological systems. They are indispensable tools for managing groundwater resources, analyzing proposed and actual remediation activities at contaminated sites, optimizing recovery of oil, gas, and geothermal energy, evaluating subsurface structures and mining activities, designing monitoring systems, assessing the long-term impacts of chemical and nuclear waste disposal, and devising improved irrigation and drainage practices in agricultural areas, among many other applications. The complexity of subsurface hydrology in the vadose zone calls for sophisticated modeling codes capable of handling the strong nonlinearities involved, the interactions of coupled physical, chemical and biological processes, and the multiscale heterogeneities inherent in such systems. The papers in this special section of ''Vadose Zone Journal'' are illustrative of the enormous potential of such numerical simulators as applied to the vadose zone. The papers describe recent developments and applications of one particular set of codes, the TOUGH family of codes, as applied to nonisothermal flow and transport in heterogeneous porous and fractured media (http://www-esd.lbl.gov/TOUGH2). The contributions were selected from presentations given at the TOUGH Symposium 2003, which brought together developers and users of the TOUGH codes at the Lawrence Berkeley National Laboratory (LBNL) in Berkeley, California, for three days of information exchange in May 2003 (http://www-esd.lbl.gov/TOUGHsymposium). The papers presented at the symposium covered a wide range of topics, including geothermal reservoir engineering, fracture flow and vadose zone hydrology, nuclear waste disposal, mining engineering, reactive chemical transport, environmental remediation, and gas transport. This Special Section of ''Vadose Zone Journal'' contains revised and expanded versions of selected papers from the
International Nuclear Information System (INIS)
1994-09-01
Transport Safety Research Abstracts (TSRA) was first published by the IAEA in 1991 as a means of disseminating information on research in radioactive material transport. This second edition utilizes International Nuclear Information System (INIS) protocol for data processing and report preparation for a research-in-progress database established by the IAEA's Division of Scientific and Technical Information. INIS subject categories and descriptors are included in the information about each project
Field research program for unsaturated flow and transport experimentation
International Nuclear Information System (INIS)
Tidwell, V.C.; Rautman, C.A.; Glass, R.J.
1992-01-01
As part of the Yucca Mountain Site Characterization Project, a field research program has been developed to refine and validate models for flow and transport through unsaturated fractured rock. Validation of these models within the range of their application for performance assessment requires a more sophisticated understanding of the processes that govern flow and transport within fractured porous media than currently exists. In particular, our research is prioritized according to understanding and modeling processes that, if not accurately incorporated into performance assessment models, would adversely impact the project's ability to evaluate repository performance. For this reason, we have oriented our field program toward enhancing our understanding of scaling processes as they relate to effective media property modeling, as well as to the conceptual modeling of complex flow and transport phenomena
International Nuclear Information System (INIS)
Koide, M.C.M.
1983-01-01
The evaluation and improvement of the diffusion code package developed by the RIS0 Research Center of Denmark have been performed. The improvements made in the package consisted in the presentation of their manuals. In order to reduce the process time of the codes an analitical boundary condition capable of representing the effects of the baffle and the reflector on the flux distribution has been calculated. Such boundary condition was obtained using a one-dimensional medium in the framework of the two group diffusion theory. The results showed that the application of this boundary condition produces very accurate results and an appreciable economy of processing time. (author) [pt
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
Transport systems research vehicle color display system operations manual
Easley, Wesley C.; Johnson, Larry E.
1989-01-01
A recent upgrade of the Transport Systems Research Vehicle operated by the Advanced Transport Operating Systems Program Office at the NASA Langley Research Center has resulted in an all-glass panel in the research flight deck. Eight ARINC-D size CRT color displays make up the panel. A major goal of the display upgrade effort was ease of operation and maintenance of the hardware while maintaining versatility needed for flight research. Software is the key to this required versatility and will be the area demanding the most detailed technical design expertise. This document is is intended to serve as a single source of quick reference information needed for routine operation and system level maintenance. Detailed maintenance and modification of the display system will require specific design documentation and must be accomplished by individuals with specialized knowledge and experience.
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
Harnessing innovation in passenger transport research in Africa
CSIR Research Space (South Africa)
Mokonyama, Mathetha T
2006-07-01
Full Text Available The paper provides the framework proposed by the newly established Built Environment Unit of CSIR, a public institution, to provide foresight driven research input into the passenger transport domain. This is modelled on the mandate of the CSIR...
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
Joint University Program for Air Transportation Research, 1984
Morrell, Frederick R. (Compiler)
1987-01-01
The research conducted during 1984 under the NASA/FAA sponsored Joint University Program for Air Transportation Research is summarized. The Joint University Program is a coordinated set of three grants sponsored by NASA Langley Research Center and the Federal Aviation Administration, one each with the Massachusetts Institute of Technology, Ohio University, and Princeton University. Completed works, status reports, and bibliographies are presented for research topics, which include navigation, guidance, control and display concepts. An overview of the year's activities for each of the schools is also presented.
Joint University Program for Air Transportation Research, 1988-1989
Morrell, Frederick R. (Compiler)
1990-01-01
The research conducted during 1988 to 1989 under the NASA/FAA-sponsored Joint University Program for Air Transportation Research is summarized. The Joint University Program is a coordinated set of three grants sponsored by NASA Langley Research Center and the Federal Aviation Administration, one each with the Massachusetts Institute of Technology, Ohio University, and Princeton University. Completed works, status reports, and annotated bibliographies are presented for research topics, which include computer science, guidance and control theory and practice, aircraft performance, flight dynamics, and applied experimental psychology. An overview of the year's activities for each university is also presented.
Energy Technology Data Exchange (ETDEWEB)
Akimoto, Hajime; Kukita; Ohnuki, Akira [Japan Atomic Energy Research Institute, Ibaraki (Japan)
1997-07-01
The Japan Atomic Energy Research Institute (JAERI) is conducting several research programs related to thermal-hydraulic and neutronic behavior of light water reactors (LWRs). These include LWR safety research projects, which are conducted in accordance with the Nuclear Safety Commission`s research plan, and reactor engineering projects for the development of innovative reactor designs or core/fuel designs. Thermal-hydraulic and neutronic codes are used for various purposes including experimental analysis, nuclear power plant (NPP) safety analysis, and design assessment.
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
Windows user-friendly code package development for operation of research reactors
International Nuclear Information System (INIS)
Hoang Anh Tuan
1998-01-01
The content of the project was to developed: 1. MS Windows interface to spectral codes like THERMOS, PEACO-COLLIS, GRACE and burn-up code. 2. MS Windows C-language burn-up diffusion hexagonal lattice code. The overall scope of the project was to develop a PC-based MS Windows code package for operation of Dalat research reactor. Various problems relating to neutronic physics like thermalization, resonance treatment, fast spectral treatment, change of isotopic concentration during burn-up time as well as burn-up distribution in the reactor core are considered in parallel to application of informatics technique. The developing process is a subject of the concept of user-friendly interface between end-users and the code package. High level input features through system of icon, menu, dialog box with regard to Common User Access (CUA) convention and sophisticated graphical output in MS Windows environment was used. The user-computer interface is also enhanced by using both keyboard and mouse, which creates a very natural manner for end-user. (author)
Expanding the notion of researcher distress: the cumulative effects of coding.
Woodby, Lesa L; Williams, Beverly Rosa; Wittich, Angelina R; Burgio, Kathryn L
2011-06-01
Qualitative researchers who explore the individual's experience of health, illness, death, and dying often experience emotional stress in their work. In this article, we describe the emotional stress we experienced while coding semistructured, after-death interviews conducted with 38 next of kin of deceased veterans. Coding sensitive topic data required an unexpected level of emotional labor, the impact of which has not been addressed in the literature. In writing this discussion article, we stepped back from our roles as interviewers/coders and reflected on how our work affected us individually and as a team, and how a sequence of exposures could exert a cumulative effect for researchers in such a dual role. Through this article, we hope to generate an expanded discourse on how qualitative inquiry impacts the emotional well-being of researchers.
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
2008-09-01
The Case Western Reserve University Department of Civil Engineering is in the process of expanding its teaching and research activities, Transportation Engineering as part of its initiative in the overall area of Infrastructure Performance and Reliab...
OpenMC: A state-of-the-art Monte Carlo code for research and development
International Nuclear Information System (INIS)
Romano, Paul K.; Horelik, Nicholas E.; Herman, Bryan R.; Nelson, Adam G.; Forget, Benoit; Smith, Kord
2015-01-01
Highlights: • OpenMC is an open source Monte Carlo particle transport code. • Solid geometry and continuous-energy physics allow high-fidelity simulations. • Development has focused on high performance and modern I/O techniques. • OpenMC is capable of scaling up to hundreds of thousands of processors. • Other features include plotting, CMFD acceleration, and variance reduction. - Abstract: This paper gives an overview of OpenMC, an open source Monte Carlo particle transport code recently developed at the Massachusetts Institute of Technology. OpenMC uses continuous-energy cross sections and a constructive solid geometry representation, enabling high-fidelity modeling of nuclear reactors and other systems. Modern, portable input/output file formats are used in OpenMC: XML for input, and HDF5 for output. High performance parallel algorithms in OpenMC have demonstrated near-linear scaling to over 100,000 processors on modern supercomputers. Other topics discussed in this paper include plotting, CMFD acceleration, variance reduction, eigenvalue calculations, and software development processes
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.)
Air medical transport personnel experiences with and opinions about research.
Fox, Jolene; Thomas, Frank; Carpenter, Judi; Handrahan, Diana
2010-01-01
This study examined air medical transport (AMT) personnel's experiences with and opinions about prehospital and AMT research. A Web-based questionnaire was sent to eight randomly selected AMT programs from each of six Association of Air Medical Services (AAMS) regions. Responders were defined by university association (UA) and AMT professional role. Forty-eight of 54 (89%) contacted programs and 536 of 1,282 (42%) individuals responded. Non-UA responders (74%) had significantly more work experience in emergency medical services (EMS) (13.5 +/- 8.5 vs. 10.8 +/- 8.3 years, P = .002) and AMT (8.3 +/- 6.3 vs. 6.8 +/- 5.7 years, P = .008), whereas UA responders (26%) had more research training (51% vs. 37%, P = .006), experience (79% vs. 59%, P < .001), and grants (7% vs. 2%, P = .006). By AMT role, administrators had the most work experience, and physicians had the most research experience. Research productivity of responders was low, with only 9% having presented and 10% having published research; and UA made no difference in productivity. A majority of responders advocated research: EMS (66%) and AMT (68%), program (53%). Willingness to participate in research was high for both EMS research (87%) and AMT research (92%). Although AMT personnel were strong advocates of and willing to participate in research, few had research knowledge. For AMT personnel, disparity exists between advocating for and producing research. Copyright 2010 Air Medical Journal Associates. Published by Elsevier Inc. All rights reserved.
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)
Power transients of Ghana research reactor-1 using PARET/ANL thermal hydraulic code
International Nuclear Information System (INIS)
Ampomah-Amoaka, E.; Akaho, E.H.K.; Anim-Sampong, S.; Nyarko, B.J.B.
2010-01-01
PARET/ANL(Version 7.3 of 2007) thermal-hydraulic code was used to perform transient analysis of the Ghana Research Reactor-1.The reactivities inserted were 2.1mk and 4mk.The peak power of 5.81kW was obtained for 2.1 mk insertion whereas the peak power for 4mk insertion of reactivity was 92.32kW.These results compare closely with experiments and theoretical studies conducted previously.
Non-US data compression and coding research. FASAC Technical Assessment Report
Energy Technology Data Exchange (ETDEWEB)
Gray, R.M.; Cohn, M.; Craver, L.W.; Gersho, A.; Lookabaugh, T.; Pollara, F.; Vetterli, M.
1993-11-01
This assessment of recent data compression and coding research outside the United States examines fundamental and applied work in the basic areas of signal decomposition, quantization, lossless compression, and error control, as well as application development efforts in image/video compression and speech/audio compression. Seven computer scientists and engineers who are active in development of these technologies in US academia, government, and industry carried out the assessment. Strong industrial and academic research groups in Western Europe, Israel, and the Pacific Rim are active in the worldwide search for compression algorithms that provide good tradeoffs among fidelity, bit rate, and computational complexity, though the theoretical roots and virtually all of the classical compression algorithms were developed in the United States. Certain areas, such as segmentation coding, model-based coding, and trellis-coded modulation, have developed earlier or in more depth outside the United States, though the United States has maintained its early lead in most areas of theory and algorithm development. Researchers abroad are active in other currently popular areas, such as quantizer design techniques based on neural networks and signal decompositions based on fractals and wavelets, but, in most cases, either similar research is or has been going on in the United States, or the work has not led to useful improvements in compression performance. Because there is a high degree of international cooperation and interaction in this field, good ideas spread rapidly across borders (both ways) through international conferences, journals, and technical exchanges. Though there have been no fundamental data compression breakthroughs in the past five years--outside or inside the United State--there have been an enormous number of significant improvements in both places in the tradeoffs among fidelity, bit rate, and computational complexity.
Asarta, Carlos J.
2016-01-01
Carlos Asarta comments here that Arbaugh, Fornaciari, and Hwang (2016) are to be commended for their work ("Identifying Research Topic Development in Business and Management Education Research Using Legitimation Code Theory" "Journal of Management Education," Dec 2016, see EJ1118407). Asarta says that they make several…
Bacon, Donald R.
2016-01-01
In this rejoinder to "Identifying Research Topic Development in Business and Management Education Research Using Legitimation Code Theory," published in the "Journal of Management Education," Dec 2016 (see EJ1118407), Donald R. Bacon discusses the similarities between Arbaugh et al.'s (2016) findings and the scholarship…
A program for undergraduate research into the mechanisms of sensory coding and memory decay
Energy Technology Data Exchange (ETDEWEB)
Calin-Jageman, R J
2010-09-28
This is the final technical report for this DOE project, entitltled "A program for undergraduate research into the mechanisms of sensory coding and memory decay". The report summarizes progress on the three research aims: 1) to identify phyisological and genetic correlates of long-term habituation, 2) to understand mechanisms of olfactory coding, and 3) to foster a world-class undergraduate neuroscience program. Progress on the first aim has enabled comparison of learning-regulated transcripts across closely related learning paradigms and species, and results suggest that only a small core of transcripts serve truly general roles in long-term memory. Progress on the second aim has enabled testing of several mutant phenotypes for olfactory behaviors, and results show that responses are not fully consistent with the combinitoral coding hypothesis. Finally, 14 undergraduate students participated in this research, the neuroscience program attracted extramural funding, and we completed a successful summer program to enhance transitions for community-college students into 4-year colleges to persue STEM fields.
Testing of a transport cask for research reactor spent fuel
International Nuclear Information System (INIS)
Mourao, Rogerio P.; Silva, Luiz Leite da; Miranda, Carlos A.; Mattar Neto, Miguel; Quintana, Jose F.A.; Saliba, Roberto O.; Novara, Oscar E.
2011-01-01
Since the beginning of the last decade three Latin American countries which operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the reactors operated in the region. As a step in this direction, a packaging for the transport of irradiated fuel from research reactors was designed by a tri-national team and a half-scale model for MTR fuel constructed in Argentina and tested in Brazil. Two test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. Although the specimen has not successfully performed the tests, its overall performance was considered very satisfactory, and improvements are being introduced to the design. A third test sequence is planned for 2011. (author)
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...
Researching on knowledge architecture of design by analysis based on ASME code
International Nuclear Information System (INIS)
Bao Shiyi; Zhou Yu; He Shuyan
2003-01-01
The quality of knowledge-based system's knowledge architecture is one of decisive factors of knowledge-based system's validity and rationality. For designing the ASME code knowledge based system, this paper presents a knowledge acquisition method which is extracting knowledge through document analysis consulted domain experts' knowledge. Then the paper describes knowledge architecture of design by analysis based on the related rules in ASME code. The knowledge of the knowledge architecture is divided into two categories: one is empirical knowledge, and another is ASME code knowledge. Applied as the basement of the knowledge architecture, a general procedural process of design by analysis that is met the engineering design requirements and designers' conventional mode is generalized and explained detailed in the paper. For the sake of improving inference efficiency and concurrent computation of KBS, a kind of knowledge Petri net (KPN) model is proposed and adopted in expressing the knowledge architecture. Furthermore, for validating and verifying of the empirical rules, five knowledge validation and verification theorems are given in the paper. Moreover the research production is applicable to design the knowledge architecture of ASME codes or other engineering standards. (author)
Research and development of electric vehicles for clean transportation.
Wada, Masayoshi
2009-01-01
This article presents the research and development of an electric vehicle (EV) in Department of Human-Robotics Saitama Institute of Technology, Japan. Electric mobile systems developed in our laboratory include a converted electric automobile, electric wheelchair and personal mobile robot. These mobile systems contribute to realize clean transportation since energy sources and devices from all vehicles, i.e., batteries and electric motors, does not deteriorate the environment. To drive motors for vehicle traveling, robotic technologies were applied.
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
Evaluation of KFB-funded research on transport systems
Energy Technology Data Exchange (ETDEWEB)
Boyce, D.; Knudsen, T.; Wegener, M.
1999-09-01
This report presents an evaluation of two research projects on transport systems, which have been financed fully or partially by KFB. The projects are: l. Systems analysis of transport markets at the Division of Transport and Location Analysis in the Department of Infrastructure and Planning of the Royal Institute of Technology, Stockholm; and 2. Planning, analysis and management in traffic networks - optimization models and methods at the Division of Optimization in the Department of Mathematics at Linkoeping University. The evaluation seeks to examine the scientific quality of the research and its relevance to the academic world and society. The two project teams prepared a self-assessment of their research activities and submitted copies of relevant publications. The evaluation committee visited both institutions and engaged the teams in discussions of their results and methodology. These visits occurred on June 1 and 2, 1999. This report is based on the self-assessments of the teams, the materials submitted and the meetings with the project teams. The evaluation and recommendations presented in the report are those of the reviewers and do not necessarily represent the views of KFB
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:
Responsible Code of Conduct for the Life Science and Dual-Use Research
International Nuclear Information System (INIS)
Bokan, S.
2007-01-01
The potential threat from misuse of current and future Dual-Use research in the field of NBC Defense is challenge to which scientific community must respond. The rapid advances in the life sciences and the worldwide growth of biotechnology industry only add urgency of this task. Code of conduct is formal statement of values and professional practices of a group of individuals with a common focus, either an occupation, academic field, or social doctrine. Codes of conduct can help to reduce the risk that scientific research will be misused. 'Dual-use' is a term often used in politics and diplomacy to refer to technology which can be used for both peaceful and military aims, usually in regard to the proliferation of nuclear weapons. Dual-use information and 'know-how' in the field of NBC defense are covered under the Export control regimes. Nearly all WMD production equipment is 'dual-use' and only very large capacity equipment is export controlled. Research in the life sciences, including NBC defense research must be conducted safely, securely, and ethically. Development of an international harmonized regime for control of biological and chemical warfare agents within and between laboratories and facilities is very important. This paper will present very important consideration of the content, promulgation and adoption of codes of conduct for scientists in the field of NBC research, for inducing of discussion between scientists into group of CBMTS members with aim how improve protection of sensitive research results and information in the field of NBC Defense sciences. (author)
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
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.
Loss of coolant acident analyses on Osiris research reactor using the RELAP5 code
International Nuclear Information System (INIS)
Soares, Humberto Vitor; Costa, Antonella Lombardi; Lima, Claubia Pereira Bezerra; Veloso, Maria Auxiliadora Fortini
2011-01-01
RELAP5/MOD 3.3 code is widely used for thermal hydraulic studies of commercial nuclear power plants. However, several current investigations have shown that RELAP5 code can also be applied for thermal hydraulic analysis of nuclear research systems with good predictions. In this paper, a nodalization of the core and the most important components of the primary cooling system of the OSIRIS reactor developed for RELAP5 thermal hydraulic code are presented as well as results of steady state and transient simulations. OSIRIS has thermal power of 70 MW and it is an open pool type research reactor moderated and cooled by water. The OSIRIS reactor characteristics have been used as a base for the development of a model for the Multipurpose Brazilian Reactor (RMB). The aim of the present work is to investigate the behavior of the core during a loss of coolant accident and the possible damage of the fuel elements due an inadequate heat removal. Although the core coolant reached the saturation point due the large break, the fuel element conditions were out of the damage zone. (author)
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
Subsurface transport with emphasis on hydrology: research needs. Subsurface Transport Program
International Nuclear Information System (INIS)
Zachara, J.M.; Wildung, R.E.
1982-03-01
A number of energy technologies presently in operation or under development generate solid wastes in large quantities as a major byproduct. These wastes will, for the most part, be disposed to the ground in landfills or inactive mine sites. Although the waste materials differ significantly among technologies, most contain residual, water-soluble chemical components which are of ecological and human health concern. Thus, in ground disposal may have a significant long-term impact on water supplies and human health if not properly conducted. With the growing magnitude of solid waste disposal operations, it becomes imperative to establish common ground between technologies such that research in this complex area can be efficiently managed to benefit a variety of users. This report develops the concept of multitechnology or generic research in subsurface transport with emphasis on hydrogeochemistry. Initially, a generic research approach was developed independent of waste characteristics. This approach both identified and prioritized the research information or experimentation and data management tools (models) required to resolve major technical concerns for in ground disposal. Waste characteristics were then evaluated to identify the common, cross-technology information needs. This evaluation indicated that solid wastes from energy producing technologies have physiocochemical properties in common which serve as a useful basis for identification of fundamental, generic research needs. Priority research projects are suggested for addressing contaminant identification, solubilization, transformation and transport. 38 references, 3 tables
OpenMC: a state-of-the-Art Monte Carlo code for research and development
International Nuclear Information System (INIS)
Romano, P.K.; Horelik, N.E.; Herman, B.R.; Forget, B.; Smith, K.; Nelson, A.G.
2013-01-01
This paper gives an overview of OpenMC, an open source Monte Carlo particle transport code recently developed at the Massachusetts Institute of Technology. OpenMC uses continuous-energy cross sections and a constructive solid geometry representation, enabling high-fidelity modeling of nuclear reactors and other systems. Modern, portable input/output file formats are used in OpenMC: XML for input, and HDF5 for output. High performance parallel algorithms in OpenMC have demonstrated near-linear scaling to over 100,000 processors on modern supercomputers. Other topics discussed in this paper include plotting, CMFD acceleration, variance reduction, eigenvalue calculations, and software development processes. (authors)
International Nuclear Information System (INIS)
Hursin, Mathieu; Xiao Shanjie; Jevremovic, Tatjana
2006-01-01
This paper summarizes the theoretical and numerical aspects of the AGENT code methodology accurately applied for detailed three-dimensional (3D) multigroup steady-state modeling of neutron interactions in complex heterogeneous reactor domains. For the first time we show the fine-mesh neutron scalar flux distribution in Purdue research reactor (that was built over forty years ago). The AGENT methodology is based on the unique combination of the three theories: the method of characteristics (MOC) used to simulate the neutron transport in two-dimensional (2D) whole core heterogeneous calculation, the theory of R-functions used as a mathematical tool to describe the true geometry and fuse with the MOC equations, and one-dimensional (1D) higher-order diffusion correction of 2D transport model to account for full 3D heterogeneous whole core representation. The synergism between the radial 2D transport and the 1D axial transport (to take into account the axial neutron interactions and leakage), called the 2D/1D method (used in DeCART and CHAPLET codes), provides a 3D computational solution. The unique synergism between the AGENT geometrical algorithm capable of modeling any current or future reactor core geometry and 3D neutron transport methodology is described in details. The 3D AGENT accuracy and its efficiency are demonstrated showing the eigenvalues, point-wise flux and reaction rate distributions in representative reactor geometries. The AGENT code, comprising this synergism, represents a building block of the computational system, called the virtual reactor. Its main purpose is to perform 'virtual' experiments and demonstrations of various mainly university research reactor experiments
Application of Code Of Conduct on the Safety of Research Reactor (RTP)
International Nuclear Information System (INIS)
Ligam, A.S.; Ahmad Nabil Abd Rahim; Zarina Masood
2014-01-01
The implementation and the practices of the effective safety system at research reactors are important to ensure that the worker, public and environment do not receive any abnormal causes. Many international safety related support agencies for research reactor such as International Atomic Energy Agency (IAEA) providing guidelines that can be applied to enhance and strengthen the enforcement of safety namely Code of Conduct on the Safety of Research Reactor (IAEA/CODEOC/RR/2006). The excellent safety management, reliability, and maintainability of RTP reactor structures, coupled with personnel numerous lessons and experiences learned, Reactor TRIGA PUSPATI research reactor providing Nuclear Malaysia personnel and visitor the very safe working and visiting environment. This paper will discuss the status, practices and improvement strategies over the past few years. (author)
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
Arbaugh, J. B.; Fornaciari, Charles J.; Hwang, Alvin
2016-01-01
Although the volume of business and management education (BME) research has expanded substantially, concerns remain about the field's legitimacy and its ability to attract new and dedicated scholars. An obstacle that may impede field development is lack of knowledge about influential works and authors to frame topical areas of inquiry and future…
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
77 FR 38709 - Surface Transportation Environment and Planning Cooperative Research Program (STEP)
2012-06-28
... for evaluating transportation measures and developing indicators of economic, social, and environmental performance of transportation systems to facilitate alternative analysis; (4) Developing and deploying research to address congestion reduction efforts; (5) Developing transportation safety planning...
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.
Energy Technology Data Exchange (ETDEWEB)
Kida, Takashi; Umeda, Miki; Sugikawa, Susumu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2003-03-01
MOX dissolution using silver-mediated electrochemical method will be employed for the preparation of plutonium nitrate solution in the criticality safety experiments in the Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF). A simulation code for the MOX dissolution has been developed for the operating support. The present report describes the outline of the simulation code, a comparison with the experimental data and a parameter study on the MOX dissolution. The principle of this code is based on the Zundelevich's model for PuO{sub 2} dissolution using Ag(II). The influence of nitrous acid on the material balance of Ag(II) is taken into consideration and the surface area of MOX powder is evaluated by particle size distribution in this model. The comparison with experimental data was carried out to confirm the validity of this model. It was confirmed that the behavior of MOX dissolution could adequately be simulated using an appropriate MOX dissolution rate constant. It was found from the result of parameter studies that MOX particle size was major governing factor on the dissolution rate. (author)
Study of steam condensation at sub-atmospheric pressure: setting a basic research using MELCOR code
Manfredini, A.; Mazzini, M.
2017-11-01
One of the most serious accidents that can occur in the experimental nuclear fusion reactor ITER is the break of one of the headers of the refrigeration system of the first wall of the Tokamak. This results in water-steam mixture discharge in vacuum vessel (VV), with consequent pressurization of this container. To prevent the pressure in the VV exceeds 150 KPa absolute, a system discharges the steam inside a suppression pool, at an absolute pressure of 4.2 kPa. The computer codes used to analyze such incident (eg. RELAP 5 or MELCOR) are not validated experimentally for such conditions. Therefore, we planned a basic research, in order to have experimental data useful to validate the heat transfer correlations used in these codes. After a thorough literature search on this topic, ACTA, in collaboration with the staff of ITER, defined the experimental matrix and performed the design of the experimental apparatus. For the thermal-hydraulic design of the experiments, we executed a series of calculations by MELCOR. This code, however, was used in an unconventional mode, with the development of models suited respectively to low and high steam flow-rate tests. The article concludes with a discussion of the placement of experimental data within the map featuring the phenomenon characteristics, showing the importance of the new knowledge acquired, particularly in the case of chugging.
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)
Prediction of Radio Code Performance--Recent Research and Need for New ARC Test
1962-09-01
A RMY( Army Project Numiber New Classification Techniques c-O0 OJ95-6o-ool Research Mem•av ’ w 3 - - PRM•ICfION 0,-Q J IO PDE JERFORMWICE...by static and -S~annel Doise than other methods of communica.tion more commonly used. Proficiency in the skill remains critical to performance in...Highland, R. W ., and Fleishman, E. A. An empirical classification of error patterns in receiving Morse code. J. Appl. Psych., 1958, 42, 112-119
Guide to federal intelligent transportation system (ITS) research.
2013-01-01
The U.S. Department of Transportations (USDOT) Intelligent Transportation System (ITS) Program aims to bring connectivity to transportation through the use of advanced wireless technologies powerful technologies that enable transformative chan...
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.
International Nuclear Information System (INIS)
Di Pasquantonio, F.
1987-01-01
The purpose of this report is that of analyzing tha problems connected with the vectorization and/or multitasking of several computer codes utilized in the ENEL-DSR Centro Ricerca Termica e Nucleare. After some general remarks on vector computers the analysis is focused on some topic relating to vectorization and multitasking of programs written for scalar computers. The priority for vectorization and/or multitasking has been given at the following codes: 1) DOT 4.2 (radiation transport and shielding); 2) QUANDRY (accidental and operating transients in LWR cores); 3) MORSE and MCNP (Monte Carlo codes for radiation transport and shielding); 4) RELAP (accidental and operating transients in LWR plants); 5) TRAP-MELT and NAUA (evaluation of source term). The principal results of the study are the following: 1) For the DOT 4.2 code it is convenients to improve the vectorized version DOT IV/C developed by Swanson introducing the parallel S.O.R. iterative method; 2) For the code QUANDRY it is proposed to introduce the three dimensional red-black mesh point ordering named ''diagonal method''; 3) To implements, on the CRAY X/MP 48, the multitasked version of the code MCNP developed by the Los Alamos National Lboratory; 4) To implements, on the CRAY X/MP 12, the vectorized and optimized version of the codes RELAP5/MOD1-MOD2 developed by J.R.C. EUROATOM-Ispra; 5) For the codes TRAP-MELT and NAUA the insertion of the vectorized routines LSODP na LSODPK for dominanting stiff cases
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.
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)
Tahir Tavukcu
2016-07-01
Full Text Available In this research, it is aimed to determine the effect of the attitudes of postgraduate students towards scientific research and codes of conduct, supported by digital script. This research is a quantitative study, and it has been formed according to pre-test & post-test research model of experiment and control group. In both groups, lessons were performed in the ways of distance education on the YDU-UZEM system and co-education. Besides, the experimental group was supported by digital scripts. Course materials have been shared through the system onto each group’s own page. The distance education lessons were performed simultaneously and non-simultaneously. The simultaneous lessons were performed through Big Blue Button virtual class add-in, and non-simultaneous lessons were performed through chatting panel and integration of the recorded lessons onto the system in order to review the lessons whenever needed. In the both groups, there are 40 (80 in total postgraduate students from the programs of the institutions of Near East University. The groups were designated, as a result of achievement test applied as a pre-test before the study, homogeneously in accordance with their school numbers with regards to success and gender; that the ones with school numbers of which last digits are odd number are the control group, and the ones with school numbers of which last digits are even number are the experimental group. In order to collect the required data, research-directed attitude scale was used after getting required permission. The obtained data were analyzed with appropriate analyzing techniques. With the findings acquired from this research, it is concluded that there is a meaningful difference in favor of the experimental group supported by the digital scripts after examining the both groups’ attitudes towards scientific research and ethics.
2014-06-01
The Research and Implementation Manual describes the administrative processes used by : Research Administration to develop and implement the Michigan Department of Transportation : (MDOT) research program. Contents of this manual include a discussion...
2015-01-01
The Research and Implementation Manual describes the administrative processes used by Research Administration to develop and implement the Michigan Department of Transportation (MDOT) research program. Contents of this manual include a discussion of ...
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.
Code-B-1 for stress/strain calculation for TRISO fuel particle (Contract research)
International Nuclear Information System (INIS)
Aihara, Jun; Ueta, Shohei; Shibata, Taiju; Sawa, Kazuhiro
2011-12-01
We have developed Code-B-1 for the prediction of the failure probabilities of the coated fuel particles for the high temperature gas-cooled reactors (HTGRs) under operation by modification of an existing code. A finite element method (FEM) is employed for the stress calculation part and Code-B-1 can treat the plastic deformation of the coating layer of the coated fuel particles which the existing code cannot treat. (author)
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
Inclusive innovation; a research project to assess the implementation of codes of conduct
Nijhof, A.H.J.; Fisscher, O.A.M.; Laan, Albertus
2002-01-01
More and more organizations formulate a code of conduct to stimulate responsible action of people within the organization. Usually much time and energy is spent fixing the content of the code. Then there is the challenge of implementing and maintaining the code. This is a tricky process in which too
research efforts on intelligent transportation system in nigeria
African Journals Online (AJOL)
The critical situation of unwelcome frustration experienced by urban trip makers and roadside dwellers alike, calls for a very strong push by all stakeholders in the transportation sector to enhance the service performance of transportation facilities using Intelligent Transportation Systems (ITS). Needed strategies for ...
Sweeney, Angela; Greenwood, Kathryn E; Williams, Sally; Wykes, Til; Rose, Diana S
2013-12-01
Health research is frequently conducted in multi-disciplinary teams, with these teams increasingly including service user researchers. Whilst it is common for service user researchers to be involved in data collection--most typically interviewing other service users--it is less common for service user researchers to be involved in data analysis and interpretation. This means that a unique and significant perspective on the data is absent. This study aims to use an empirical report of a study on Cognitive Behavioural Therapy for psychosis (CBTp) to demonstrate the value of multiple coding in enabling service users voices to be heard in team-based qualitative data analysis. The CBTp study employed multiple coding to analyse service users' discussions of CBT for psychosis (CBTp) from the perspectives of a service user researcher, clinical researcher and psychology assistant. Multiple coding was selected to enable multiple perspectives to analyse and interpret data, to understand and explore differences and to build multi-disciplinary consensus. Multiple coding enabled the team to understand where our views were commensurate and incommensurate and to discuss and debate differences. Through the process of multiple coding, we were able to build strong consensus about the data from multiple perspectives, including that of the service user researcher. Multiple coding is an important method for understanding and exploring multiple perspectives on data and building team consensus. This can be contrasted with inter-rater reliability which is only appropriate in limited circumstances. We conclude that multiple coding is an appropriate and important means of hearing service users' voices in qualitative data analysis. © 2012 John Wiley & Sons Ltd.
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.
2016-05-01
The Accessible Transportation Technologies Research Initiative (ATTRI) is a joint U.S. Department of Transportation (U.S. DOT) initiative that is co-led by the Federal Highway Administration (FHWA) and the Federal Transit Administration (FTA). ATTRI ...
Evaluation of Tehran research reactor (TRR) control rod worth using MCNP4C computer code
International Nuclear Information System (INIS)
Hosseini, Mohammad; Vosoughi, Naser; Hosseini, Seyed Abolfazl
2010-01-01
The main objective of reactor control system is to provide a safe reactor starting up, operation and shutting down. Calculation or measurement of precise values of control rod worth is of great importance in Tehran Research Reactor (TRR), considering the fact that they are the only controlling tools in the reactor. In present paper, simulation of TRR in First Operation Cycle (FOC) and in cold and clean core for the calculation of total and integral worth of control nods is reported. MCNP4C computer code has been used for all simulation process. Two method have been used for control rods worth calculation in this paper, namely the direct approach and perturbation method. It is shown that while the direct approach is appropriate for worth calculation of both the shim and the regulating control rods, the perturbation method is just suitable for tiny reactivity changes, i.e. for small initial part of regulating rods. Results of simulation are compared with the reported data in Safety Analysis Report (SAR) of Tehran research reactor and showed satisfactory agreement. (author)
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
Development of a computer code for Dalat research reactor transient analysis
International Nuclear Information System (INIS)
Le Vinh Vinh; Nguyen Thai Sinh; Huynh Ton Nghiem; Luong Ba Vien; Pham Van Lam; Nguyen Kien Cuong
2003-01-01
DRSIM (Dalat Reactor SIMulation) computer code has been developed for Dalat reactor transient analysis. It is basically a coupled neutronics-hydrodynamics-heat transfer code employing point kinetics, one dimensional hydrodynamics and one dimensional heat transfer. The work was financed by VAEC and DNRI in the framework of institutional R and D programme. Some transient problems related to reactivity and loss of coolant flow was carried out by DRSIM using temperature and void coefficients calculated by WIMS and HEXNOD2D codes. (author)
Carey, Lindsay B; Cohen, Jeffrey
2015-10-01
The World Health Organization (WHO) 'Pastoral Intervention Codings' were first released in 2002 as part of the 'International Statistical Classification of Diseases and Related Health Problems' (WHO 2002). The purpose of the WHO pastoral intervention codings (colloquially abbreviated as 'WHO-PICs') was to record and account for the religious, pastoral and/or spiritual interventions of chaplains and volunteers providing care to patients and other clients experiencing religious and/or spiritual health and well-being issues. The intent of such WHO codings was to provide information in five areas: statistical, research, clinical, education and policy. The purpose of this paper predominantly accounts for research although it does intersect and relate to other WHO priorities. Over the past 10 years, research by the current and associated authors to test the efficacy of the WHO-PICs has been implemented in a number of different health and welfare contexts that have engaged chaplaincy personnel. In summary, while the WHO-PICs are yet to be more widely utilized internationally, the codings have largely proven to be valuable indices appropriate to a variety of contexts. Research utilizing the WHO-PICs, however, has also revealed the necessity for a number of changes and inclusions to be implemented. Recommendations concerning the future utilisation of the WHO-PICs are made, as are recommendations for these codings to be further developed and promoted by the WHO, so as to more accurately record religious, pastoral and spiritual interventions.
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
2008-05-01
The Montana Department of Transportation (MDT) contracted the Bureau of Business and Economic Research at the University of Montana Missoula to conduct research to determine how other states solicit, prioritize, and select research problem statem...
Evaluating department of transportation's research program : a methodology and case study.
2012-06-01
An effective research program within a transportation organization can be a valuable asset to accomplish the goals of the overall : mission. Determining whether a research program is pursuing relevant research projects and obtaining results for the s...
The world anti-doping code : a South African perspective : research ...
African Journals Online (AJOL)
During February 2003 the World Anti-Doping Agency adopted the World-Anti Doping Code in Copenhagen in an effort to create and independent anti-doping body and to co-ordinate the harmonisation of doping regulations. The Code encompasses the principles around which the anti-doping effort in sport will revolve in ...
The Evolution of a Coding Schema in a Paced Program of Research
Winters, Charlene A.; Cudney, Shirley; Sullivan, Therese
2010-01-01
A major task involved in the management, analysis, and integration of qualitative data is the development of a coding schema to facilitate the analytic process. Described in this paper is the evolution of a coding schema that was used in the analysis of qualitative data generated from online forums of middle-aged women with chronic conditions who…
The research of atmospheric 2D optical PPM CDMA system with turbo coding
Zhou, Xiuli; Li, Zaoxia
2007-11-01
The atmospheric two-dimensional optical code-division multiple-access (CDMA) systems using pulse-position modulation (PPM) and Turbo-coded were presented. We analyzed the bit-error rate (BER) of the proposed system using pulse-position modulation (PPM) with considering the effects of the scintillation, avalanche photodiode noise, thermal noise, and multi-user interference. We showed that the atmospheric two dimensional (2D) optical PPM CDMA systems can realize high-speed communications when the logarithm variance of the scintillation is less than 0.1, and the turbo-coded atmospheric optical CDMA system has better bit error rate(BER) performance than the atmospheric optical PPM CDMA systems without turbo-coded. We also showed that the turbo-coded system has better performance than the multi-user detection system.
Antonacopoulou, Elena P.
2016-01-01
In "Identifying Research Topic Development in Business and Management Education Research Using Legitimation Code Theory," authors J.B. Arbaugh, Charles J. Fornaciari, and Alvin Hwang ("Journal of Management Education," December 2016 vol. 40 no. 6 p654-691, see EJ1118407) used citation analysis to track the development of…
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)
Singh, Tej; Kumar, Jainendra; Sharma, Archana; Singh, Kanchhi; Raina, V.K.; Srinivasan, P.
2009-01-01
At present Dhruva and Cirus reactors provide majority of research reactor based experimental/irradiation facilities to cater to various needs of the vast pool of researchers in the field of sciences research and development work for nuclear power plants and production of radioisotopes. With a view to further consolidate and expand the scope of research and development in nuclear and allied sciences, a new 30 MWt Multi Purpose Research Reactor is proposed to be constructed. This paper describes some of the physics design features of this reactor using MCNP code to validate the deterministic methods. The criticality calculations for 100 material testing reactor (JHR) of France and 610 MW SAVANNAH thermal reactor were performed using MCNP computer codes to boost the confidence level in designing the physics design of reactor core. (author)
Experience of IEA-R1 research reactor spent fuel transportation back to United States
Energy Technology Data Exchange (ETDEWEB)
Frajndlich, Roberto [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Div. de Operacao do Reator IEAR-R1m]. E-mail: frajndli@net.ipen.br; Perrotta, Jose A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Div.de Engenharia do Nucleo]. E-mail: perrotta@net.ipen.br; Maiorino, Jose Rubens [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Diretoria de Reatores]. E-mail: maiorino@net.ipen.br; Soares, Adalberto Jose [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Dept. de Reatores]. E-mail: ajsoares@net.ipen.br
1998-07-01
IPEN/CNEN-SP is sending the IEA-R1 Research Reactor spent fuels from USA origin back to this country. This paper describes the experience in organizing the negotiations, documents and activities to perform the transport. Subjects as cask licensing, transport licensing and fuel failure criteria for transportation are presented. (author)
Scientific research about climate change mitigation in transport: a critical review
Schwanen, T.; Banister, D.; Anable, J.
2011-01-01
This paper seeks to develop a deeper understanding of the research on climatechangemitigation in transport. We suggest that work to date has focused on the effects of improvements in transport technologies, changes in the price of transport, physical infrastructure provision, behavioural change and
Energy Technology Data Exchange (ETDEWEB)
Talley, Darren G.
2017-04-01
This report describes the work and results of the verification and validation (V&V) of the version 1.0 release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, the equation of motion for fuel element thermal expansion, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This V&V effort was intended to confirm that the code shows good agreement between simulation and actual ACRR operations.
Research on magnetic materials of interest in transportation
2000-04-01
This paper reports the results of an investigation on magnetic materials of interest in the transportation field. It includes information about the present state of magnetic materials and examines the recently discovered phenomenon referred to as col...
Final report, Portland State University intelligent transportation research initiative.
2006-07-01
This FY 2004 ITS Integration grant has provided partial funding for design, outfitting, and interior fit up for the new regional Intelligent Transportation Systems (ITS) Laboratory suite located in the new $60 million Northwest Center for Engin...
Sign Life-Cycle Policies and Practices : Transportation Research Synthesis
2017-10-01
MnDOT Metro District Traffic Engineering is interested in the practices that other state departments of transportation (DOTs) use to determine traffic sign life expectancy and replacement. Of particular interest is the state of the practice regarding...
Development of dynamic analysis code for HTTR hydrogen production system (Contract research)
International Nuclear Information System (INIS)
Maeda, Yukimasa; Nishihara, Tetsuo; Ohashi, Hirohumi; Sato, Hiroyuki; Inagaki, Yoshiyuki
2005-03-01
A heat and mass balance analysis code (N-HYPAC) has been developed to investigate transient behavior in the HTTR hydrogen production system. The code can analyze heat and mass transfer (temperature and mass and pressure distributions of process and helium gases) and behavior of the control system under both static state (case of steady operation) and dynamic state (case of transient operation). Analysis model of helium and process gases from IHX to secondary helium loop and hydrogen production system has been constructed. This report describes analytical flow sheet, construction of the code, basic equations, method to treat the input data, estimation of the preliminary analysis. (author)
Research on the Logistics Supply Chain in Port Logistics Transportation
Wang Yan-liang
2013-01-01
The aim of this study is to improve and increase the logistics system effectiveness and to solve the problem of optimal movement of different flows. Logistics transport carrying the world on material resources transfer exchange important mission and economic development and our lives are closely linked, logistics chain logistics transport occupies an important position and in the e logistics chain in port logistics has play a decisive role. For many coastal countries port logistics is the eco...
Energy Technology Data Exchange (ETDEWEB)
Galonska, Andreas
2010-03-15
In the present master thesis the development oa an automatic validation system for the simulation code ERO is documented. This 3D Monte-carlo code models the transport of impurities as well as plasma-wall interaction processes and has great importance for the fusion research. The validation system is based on JuBE (Julich Benchmarking Environment), the flexibility of which allows a slight extension of the system to other codes, for instance such, which are operated in the framework of the EU Task Force ITM (Integrated Tokamak Modelling). The chosen solution - JuBE and a special program for the ''intellectual'' comparison of actual and reference-edition data of ERO is described and founded. The use of this program and the configuration of JuBE are detailedly described. Simulations to different plasma experiments, which serve as reference cases for the automatic validation, are explained. The working of the system is illustrated by the description of a test case. This treats the failure localization and improvement in the parallelization of an important ERO module (tracking of physically eroded particle). It is demonstrated, how the system reacts in an erroneous validation and the subsequently performed error correction leads to a positive result. Finally a speed-up curve of the parallelization is established by means of the output data of JuBE.
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.)
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.
Gläser, Jochen; Laudel, Grit
2013-01-01
Qualitative research aimed at "mechanismic" explanations poses specific challenges to qualitative data analysis because it must integrate existing theory with patterns identified in the data. We explore the utilization of two methods—coding and qualitative content analysis—for the first steps in the
Nursing students and teaching of codes of ethics: an empirical research study.
Numminen, O H; Leino-Kilpi, H; van der Arend, A; Katajisto, J
2009-12-01
To explore graduating nursing students' perception of nurse educators' teaching of codes of ethics in polytechnics providing basic nursing education in Finland. Codes of ethics are regarded as an essential content in most nursing ethics curricula. However, little is known about how their teaching is implemented. Descriptive, cross-sectional design was used in this study. A total of 214 nursing students responded to a structured questionnaire with one open-ended question. The data was analysed statistically by SPSS and content analysis. Students perceived teaching of the codes as fairly extensive. The emphasis was on the nurse-patient relationship. Less attention was paid to nursing in wider social contexts. Educators' use of teaching and evaluation methods was narrow. Students whose teaching had been integrated into clinical training perceived that teaching had been more extensive. However, students did not perceive integration to clinical training as a much used teaching format. Students assessed their own knowledge and ability to apply the codes as mediocre. Those educators, whose knowledge about the codes students had assessed as adequate, were also perceived to teach the codes more extensively. Regardless of the responding students' positive description of the teaching, the findings should be interpreted with caution, due to the students' limited interest to respond. In teaching ethics, particular attention should be paid to more versatile use of teaching and evaluation methods, organization of integrated teaching, educators' competence in ethics, and student outcomes so that the importance of ethics would come across to all nursing students.
Energy Technology Data Exchange (ETDEWEB)
Hainoun, A., E-mail: pscientific2@aec.org.sy [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Doval, A. [Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400 S.C de Bariloche, Rio Negro (Argentina); Umbehaun, P. [Centro de Engenharia Nuclear – CEN, IPEN-CNEN/SP, Av. Lineu Prestes 2242-Cidade Universitaria, CEP-05508-000 São Paulo, SP (Brazil); Chatzidakis, S. [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States); Ghazi, N. [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Park, S. [Research Reactor Design and Engineering Division, Basic Science Project Operation Dept., Korea Atomic Energy Research Institute (Korea, Republic of); Mladin, M. [Institute for Nuclear Research, Campului Street No. 1, P.O. Box 78, 115400 Mioveni, Arges (Romania); Shokr, A. [Division of Nuclear Installation Safety, Research Reactor Safety Section, International Atomic Energy Agency, A-1400 Vienna (Austria)
2014-12-15
Highlights: • A set of advanced system thermal hydraulic codes are benchmarked against IFA of IEA-R1. • Comparative safety analysis of IEA-R1 reactor during LOFA by 7 working teams. • This work covers both experimental and calculation effort and presents new out findings on TH of RR that have not been reported before. • LOFA results discrepancies from 7% to 20% for coolant and peak clad temperatures are predicted conservatively. - Abstract: In the framework of the IAEA Coordination Research Project on “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal hydraulic computational methods and tools for operation and safety analysis of research reactors” the Brazilian research reactor IEA-R1 has been selected as reference facility to perform benchmark calculations for a set of thermal hydraulic codes being widely used by international teams in the field of research reactor (RR) deterministic safety analysis. The goal of the conducted benchmark is to demonstrate the application of innovative reactor analysis tools in the research reactor community, validation of the applied codes and application of the validated codes to perform comprehensive safety analysis of RR. The IEA-R1 is equipped with an Instrumented Fuel Assembly (IFA) which provided measurements for normal operation and loss of flow transient. The measurements comprised coolant and cladding temperatures, reactor power and flow rate. Temperatures are measured at three different radial and axial positions of IFA summing up to 12 measuring points in addition to the coolant inlet and outlet temperatures. The considered benchmark deals with the loss of reactor flow and the subsequent flow reversal from downward forced to upward natural circulation and presents therefore relevant phenomena for the RR safety analysis. The benchmark calculations were performed independently by the participating teams using different thermal hydraulic and safety
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)
O'Connor, Rory
2012-01-01
The primary aim of this chapter is to outline a potentially powerful framework for the combination of research approaches utilizing the Grounded Theory coding mechanism for Case Study, and Focus Groups data analysis. A secondary aim of this chapter is to provide a roadmap for such a usage by way of an example research project. The context for this project is the need to study and evaluate the actual practice of software development processes in real world commercial settings of software compa...
2014-05-01
Land use and transportation are inextricably linked. Models that capture the dynamics and interactions : of both systems are indispensable for evaluating alternative courses of action in policy and investment. : These models must be spatially disaggr...
2014-01-01
This issue of Research Showcase features articles on two successful research efforts, one on quiet : pavements and the other on the bene ts of prismatic sign sheeting, and an article on university : transportation center participation in Florida.
2009-01-30
This report documents the results of the research program completed by the Advanced Technologies for Transportation Research Program (ATTRP) at the University of Tennessee at Chattanooga (UTC) under Federal Transit Administration Cooperative Agreemen...
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)
Validation Calculations for the Application of MARS Code to the Safety Analysis of Research Reactors
International Nuclear Information System (INIS)
Park, Cheol; Kim, H.; Chae, H. T.; Lim, I. C.
2006-10-01
In order to investigate the applicability of MARS code to the accident analysis of the HANARO and other RRs, the following test data were simulated. Test data of the HANARO design and operation, Test data of flow instability and void fraction from published documents, IAEA RR transient data in TECDOC-643, Brazilian IEA-R1 experimental data. For the simulation of the HANARO data with finned rod type fuels at low pressure and low temperature conditions, MARS code, developed for the transient analysis of power reactors, was modified. Its prediction capability was assessed against the experimental data for the HANARO. From the assessment results, it can be said that the modified MARS code could be used for analyzing the thermal hydraulic transient of the HANARO. Some other simulations such as flow instability test and reactor transients were also done for the application of MARS code to RRs with plate type fuels. In the simulation for these cases, no modification was made. The results of simulated cases show that the MARS code can be used to the transient analysis of RRs with careful considerations. In particular, it seems that an improvement on a void model may be necessary for dealing with the phenomena in high void conditions
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.
Challenges and future research needs towards international freight transport modelling
Meersman, H.; Ehrler, C.C.; Bruckmann, D.; Chen, T.M.; Francke, J.; Hill, P.; Jackson, C.; Klauenberg, J.; Kurowski, M.; Seidel, S.; Vierth, I.
2016-01-01
The advanced internationalisation of markets and production processes continuously adds to the complexity of supply chains. At the same time improving the sustainability of the related international freight transport processes and optimising their efficiency is becoming a topic of central relevance.
Studies and research concerning BNFP. Nuclear spent fuel transportation studies
International Nuclear Information System (INIS)
Anderson, R.T.; Maier, J.B.
1979-11-01
Currently, there are a number of institutional problems associated with the shipment of spent fuel assemblies from commercial nuclear power plants: new and conflicting regulations, embargoing of certain routes, imposition of transport safeguards, physical security in-transit, and a lack of definition of when and where the fuel will be moved. This report presents a summary of these types and kinds of problems. It represents the results of evaluations performed relative to fuel receipt at the Barnwell Nuclear Fuel Plant. Case studies were made which address existing reactor sites with near-term spent fuel transportation needs. Shipment by either highway, rail, water, or intermodal water-rail was considered. The report identifies the impact of new regulations and uncertainty caused by indeterminate regulatory policy and lack of action on spent fuel acceptance and storage. This stagnant situation has made it impossible for industry to determine realistic transportation scenarios for business planning and financial risk analysis. A current lack of private investment in nuclear transportation equipment is expected to further prolong the problems associated with nuclear spent fuel and waste disposition. These problems are expected to intensify in the 1980's and in certain cases will make continuing reactor plant operation difficult or impossible
Improving School Bus Safety. Transportation Research Board Special Report 222.
National Academy of Sciences - National Research Council, Washington, DC. Transportation Research Board.
While school buses transport more passengers per trip, the rate of occupant fatalities per mile driven for school buses is one-quarter that for passenger cars. Nevertheless, the public expects school districts and other school bus operators to take all reasonable precautions to protect children as they travel to and from school. Although a variety…
research efforts on intelligent transportation system in nigeria
African Journals Online (AJOL)
user
INTELLIGENT TRANSPORTATION SYSTEM IN NIGERIA: DEVELOPMENT OF TRIP PLANNING MODELS. O. Adeleke, et al. Nigerian Journal of Technology. Vol. 35. No. 3, July 2016. 492. Since the introduction of ITS, there has been a proliferation of interest by the various stakeholders. (government and industry) because ...
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.
Research on verification and validation strategy of detonation fluid dynamics code of LAD2D
Wang, R. L.; Liang, X.; Liu, X. Z.
2017-07-01
The verification and validation (V&V) is an important approach in the software quality assurance of code in complex engineering application. Reasonable and efficient V&V strategy can achieve twice the result with half the effort. This article introduces the software-Lagrangian adaptive hydrodynamics code in 2D space (LAD2D), which is self-developed software in detonation CFD with plastic-elastic structure. The V&V strategy of this detonation CFD code is presented based on the foundation of V&V methodology for scientific software. The basic framework of the module verification and the function validation is proposed, composing the detonation fluid dynamics model V&V strategy of LAD2D.
Wu, Menglong; Han, Dahai; Zhang, Xiang; Zhang, Feng; Zhang, Min; Yue, Guangxin
2014-03-10
We have implemented a modified Low-Density Parity-Check (LDPC) codec algorithm in ultraviolet (UV) communication system. Simulations are conducted with measured parameters to evaluate the LDPC-based UV system performance. Moreover, LDPC (960, 480) and RS (18, 10) are implemented and experimented via a non-line-of-sight (NLOS) UV test bed. The experimental results are in agreement with the simulation and suggest that based on the given power and 10(-3)bit error rate (BER), in comparison with an uncoded system, average communication distance increases 32% with RS code, while 78% with LDPC code.
International Nuclear Information System (INIS)
Lim, I.C.; Hwang, S.Y.; Woo, J.S.; Lee, M.; Jun, B.J.
2003-01-01
Full text: The safety culture in HANARO was self-assessed in accordance with the Code of Conduct on the Safety of Research Reactor drafted by IAEA. From 2002, IAEA has worked on the development of the Code of Conduct to achieve and maintain high level of nuclear safety in research reactors worldwide through the enhancement of national measures and international co-operation including, where appropriate, safety related technical cooperation. It defines the role of the state, the role of the regulatory body, the role of the operating organization and the role of the IAEA. As for the role of operating organization, the code specifies general requirements in assessment and verification of safety, financial and human resources, quality assurance, human factors, radiation protection and emergency preparedness. It also defines the role of operating organization for safety of research reactor in siting, design, operation, maintenance, modification and utilization as well. All of these items are the subjects for safety culture implementation, which means the Code could be a guideline for an operating organization to assess its safety culture. The self-assessment of safety culture in HANARO was made by using the sections of the Code describing the role of the operating organization for safety of research reactor. The major assessment items and the practices in HANARO for each items are as follow: The SAR of HANARO was reviewed by the regulatory body before the construction and the fuel loading of HANARO. Major design modifications and new installation of utilization facility needs the approval from regulatory body and safety assessment is a requirement for the approval. The Tech. Spec. for HANARO Operation specifies the analysis, surveillance, testing and inspection for HANARO operation. The reactor operation is mainly supported by the government and partly by nuclear R and D fund. The education and training of operation staff are one of major tasks of operating organization
Basic mechanisms of gas transport and past research using perfluorocarbons.
Spiess, Bruce D
2010-03-01
Perfluorocarbon compounds have been utilized either in pure (neat) form or as emulsions suspended in aqueous fluids. These man-made chemicals possess a unique physical property allowing them to dissolve much more respiratory gases than any water-based system. Understanding the basic physical chemistry surrounding these emerging medical technologies will assure they are utilized to maximum benefit for mankind. It is clear they should not simply be viewed as 'blood substitutes' but rather as enhanced gas transport pharmaceuticals.
Operations Research In Maritime Transport And Freight Logistics
Directory of Open Access Journals (Sweden)
Shubham Tuslyan
2017-11-01
Full Text Available Todays globalization would be impossible without modern cost-effective merchant ships crossing the seas. World trade was 17 times as high at the end of the 20th century as it was 50 years previously. A shipping industry that has steadily lowered its costs has been a prerequisite of this development and there are no signs that the world economy will rely any less heavily on sea transport in the future. The current decade has witnessed a remarkable growth in container transportation and vessel sizes India is the 20th largest maritime country in the world. Its strategic location of a long coastline that flanks important global shipping routes makes it a major maritime nation. The maritime sector in India comprises of ports shipping shipbuilding and ship repair as well as inland water transport systems. About 95 of the countrys trade by volume and 70 by value is moved through maritime transport.Among the problems to be solved there are the spatial allocation of containers on the terminal yard optimization of shipping routes allocation of ships to berths and cranes allocation of cargo to ships scheduling priorities and operations in order to maximize performances based on some economic indicators. During the evaluation of the identified studies it becomes clear that the existing literature can be further subdivided into analytical simulation and combined approaches. The majority of the papers 212 out of 243 or 87 adopted analytical approaches that exclusively apply optimization algorithms to optimize container terminal operations. However in order to optimize the entire container terminal operations the use of this approach to simultaneously deal with different types of problems is difficult although not impossible especially in regard to stand-alone components. This is a major limitation of the widely used analytical approaches in traditional literature.
Case studies of market research for three transportation communication products
1994-03-01
This report completes a two-part project in support of the Volpe Center program, Public Acceptance and Markets for Various IVHS Services. The first report, A Primer on Marketing Research, provides an overview of the research approaches an...
Research on unequal error protection with punctured turbo codes in jpeg image transmission system
Directory of Open Access Journals (Sweden)
Lakhdar Moulay A.
2007-01-01
Full Text Available An investigation of Unequal Error Protection (UEP methods applied to JPEG image transmission using turbo codes is presented. The JPEG image is partitioned into two groups, i.e., DC components and AC components according to their respective sensitivity to channel noise. The highly sensitive DC components are better protected with a lower coding rate, while the less sensitive AC components use a higher coding rate. While we use the s-random interleaver and s-random odd-even interleaver combined with odd-even puncturing, we can fix easily the local rate of turbo-code. We propose to modify the design of s-random interleaver to fix the number of parity bits. A new UEP scheme for the Soft Output Viterbi Algorithm (SOVA is also proposed to improve the performances in terms of Bit Error Rate (BER and Peak Signal to Noise Ratio (PSNR. Simulation results are given to demonstrate how the UEP schemes outperforms the equal error protection (EEP scheme in terms of BER and PSNR.
Research progress on the numerical simulation of Z-pinch implosion using mared code
International Nuclear Information System (INIS)
Ding Ning; Wu Jiming; Yang Zhenhua; Fu Shangwu; Ning Cheng; Shu Xiaojian; Zhang Yang; Dai Zihuan; Yao Yanzhong; Yin Li; Sun Shunkai
2010-01-01
The physical scheme of the MARED code, a two-dimensional three-temperature radiation magneto-hydrodynamics code for Z-pinch implosion simulation, is described. Results from the one- and two-dimensional calculation tests demonstrate the MARED code is able to simulate Z-pinch implosions of a wide range of accelerator and load parameters. It is able to present the primary dynamic characteristics of Z-pinch implosions, and the calculated images and rules qualitatively agree with the theoretical analyses and experimental observations. Compared with the experimental data, simulation results show that, under the same condition, the tungsten wire-array implosion has higher X-ray radiation power output than aluminum wire arrays. With same load parameters, the X-ray radiation power increases with the load current. Under the certain drive condition, the X-ray output decreases with the load mass. The MARED code is also used to simulate the radiation field formation of the wire-array filled with foam. The preliminary results on the Z machine are qualitative consistent with the simulation results from the Sandia laboratory. (authors)
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)
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)
Know-How Transfer and Training Issues for the Transport Research Professional
Directory of Open Access Journals (Sweden)
Prof. George A. Giannopoulos
2015-06-01
Other relevant actions could be taken within the existing collaborative Transport research programmes e.g. the Transport pillar of the “societal challenges” part of the H2020 programme and could consist of specific provisions, impeded in the research contracts, allowing funding for activities such as web-training short courses and workshops, formulation and provision of training materials, holding workshops with the involvement of senior research personnel or leading international academics, etc.
International Nuclear Information System (INIS)
Brockmeier, U.; Unger, H.
1992-01-01
Within the scope of the project BMFT No. 15008317 entitled ''Comparative Assessment of Different Computer Codws for Severe Accident Analysis, Contribution to the ATHLET/SA-Code Development'' the codes ATHLET/SA, CATHARE/ICARE, MELCOR and SCDAP/RELAP5 are investigated. Emphasis is put on a comparison and an assessment of the governing modelling features implemented and operating in the codes under consideration. The codes are evaluated and compared on the base of selected experiments (especially the CORA experimental program of the Karlsruhe Research Center) and relevant severe accident scenarios. The present report is a reference study dealing with the governing models implemented in the severe accident codes SCDAP/RELAP5, ATHLET/SA, CATHARE/ICARE, MELCOR, KESS-III, MAAP and MELPROG/TRAC. Emphaisis is laid on the following models (molstly implemented in form of modules in the respective codes) dealing with: - thermal hydraulics; - heat generation and heat structures; - Radiation heat transfer; - mechanical (rod) behaviour; - core heatup, meltdown and relocation; - chemical reaction; - fission product release and transport; - material properties; - specific components. (orig.) [de
International Nuclear Information System (INIS)
Asakura, Toshihide; Sato, Makoto; Matsumura, Masakazu; Morita, Yasuji
2005-01-01
This paper reviews the succeeding development and utilization of Extraction System Simulation Code for Advanced Reprocessing (ESSCAR). From the viewpoint of development, more tests with spent fuel and calculations should be performed with better understanding of the physico-chemical phenomena in a separation process. From the viewpoint of process safety research on fuel cycle facilities, it is important to know the process behavior of a key substance; being highly reactive but existing only trace amount. (author)
Directory of Open Access Journals (Sweden)
Z. Gholamzadeh
2018-02-01
Full Text Available The neutron powder diffractometer (NPD is used to study a variety of technologically important and scientifically driven materials such as superconductors, multiferroics, catalysts, alloys, ceramics, cements, colossal magnetoresistance perovskites, magnets, thermoelectrics, zeolites, pharmaceuticals, etc. Monte Carlo–based codes are powerful tools to evaluate the neutronic behavior of the NPD. In the present study, MCNPX 2.6.0 and Vitess 3.3a codes were applied to simulate NPD facilities, which could be equipped with different optic devices such as pyrolytic graphite or neutron chopper. So, the Monte Carlo–based codes were used to simulate the NPD facility of the 5 MW Tehran Research Reactor. The simulation results were compared to the experimental data. The theoretical results showed good conformity to experimental data, which indicates acceptable performance of the Vitess 3.3a code in the neutron optic section of calculations. Another extracted result of this work shows that application of neutron chopper instead of monochromator could be efficient to keep neutron flux intensity higher than 106 n/s/cm2 at sample position.
Aviation safety research and transportation/hazard avoidance and elimination
Sonnenschein, C. M.; Dimarzio, C.; Clippinger, D.; Toomey, D.
1976-01-01
Data collected by the Scanning Laser Doppler Velocimeter System (SLDVS) was analyzed to determine the feasibility of the SLDVS for monitoring aircraft wake vortices in an airport environment. Data were collected on atmospheric vortices and analyzed. Over 1600 landings were monitored at Kennedy International Airport and by the end of the test period 95 percent of the runs with large aircraft were producing usable results in real time. The transport was determined in real time and post analysis using algorithms which performed centroids on the highest amplitude in the thresholded spectrum. Making use of other parameters of the spectrum, vortex flow fields were studied along with the time histories of peak velocities and amplitudes. The post analysis of the data was accomplished with a CDC-6700 computer using several programs developed for LDV data analysis.
Energy and Environmental Issues, 1991. Transportation research record
International Nuclear Information System (INIS)
1991-01-01
Partial Contents: Mitigation of Traffic Mortality of Endangered Brown Pelicans on Coastal Bridges; Cooperation Between State Highway and Environmental Agencies in Dealing With Hazardous Waste in the Right-of-Way; Comparison of Intersection Air Quality Models' Ability to Simulate Carbon Monoxide Concentrations in an Urban Area; Model Calculation of Environment-Friendly Traffic Flows in Urban Networks; Sensitivity Analysis for Land Use, Transportation, and Air Quality; Special Events and Carbon Monoxide Violations: TSM, Crowd Control, Economics, and Solutions to Adverse Air Quality Impacts; Mode Split at Large Special Events and Effects on Air Quality; Internal Consistency and Stability of Measurements of Community Reaction to Noise; Impact and Potential Use of Attitude and Other Modifying Variables in Reducing Community Reaction to Noise; Techniques for Aesthetic Design of Freeway Noise Barriers; Effects of Road Surface Texture on Traffic and Vehicle Noise; Electrokinetic Soil Processing in Waste Remediation and Treatment: Synthesis of Available Data; Site Remediation by In Situ Vitrification
Research on Community Structure in Bus Transport Networks
International Nuclear Information System (INIS)
Yang Xuhua; Wang Bo; Sun Youxian
2009-01-01
We abstract the bus transport networks (BTNs) to two kinds of complex networks with space L and space P methods respectively. Using improved community detecting algorithm (PKM agglomerative algorithm), we analyze the community property of two kinds of BTNs graphs. The results show that the BTNs graph described with space L method have obvious community property, but the other kind of BTNs graph described with space P method have not. The reason is that the BTNs graph described with space P method have the intense overlapping community property and general community division algorithms can not identify this kind of community structure. To overcome this problem, we propose a novel community structure called N-depth community and present a corresponding community detecting algorithm, which can detect overlapping community. Applying the novel community structure and detecting algorithm to a BTN evolution model described with space P, whose network property agrees well with real BTNs', we get obvious community property. (general)
The research on natural gas pipeline transportation price formulation method
Directory of Open Access Journals (Sweden)
YU Wenjia
2014-02-01
Full Text Available This paper will introduce a method of natural gas pipeline transportation price on the basis of two-part tariff.Distance,investment and income have been taken into consideration.The total fee is divided into three parts:reservation fee,usage fee and peak-load regulation fee.Because there are different types of users in the natural gas market who show great difference in the continuity and reliability of gas supply,capacity of bearing price,elastic demand and balance use of gas,according to the method,the different types of users can pay reasonable fee.This method not only considers the investment income recovery but also considers the different types of users paying a reasonable fee.We hope the new pricing model can give a reference to the development of China's natural gas industry.