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Sample records for radiation transport code

  1. 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

  2. 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)

  3. Validation of comprehensive space radiation transport code

    International Nuclear Information System (INIS)

    Shinn, J.L.; Simonsen, L.C.; Cucinotta, F.A.

    1998-01-01

    The HZETRN code has been developed over the past decade to evaluate the local radiation fields within sensitive materials on spacecraft in the space environment. Most of the more important nuclear and atomic processes are now modeled and evaluation within a complex spacecraft geometry with differing material components, including transition effects across boundaries of dissimilar materials, are included. The atomic/nuclear database and transport procedures have received limited validation in laboratory testing with high energy ion beams. The codes have been applied in design of the SAGE-III instrument resulting in material changes to control injurious neutron production, in the study of the Space Shuttle single event upsets, and in validation with space measurements (particle telescopes, tissue equivalent proportional counters, CR-39) on Shuttle and Mir. The present paper reviews the code development and presents recent results in laboratory and space flight validation

  4. Los Alamos radiation transport code system on desktop computing platforms

    International Nuclear Information System (INIS)

    Briesmeister, J.F.; Brinkley, F.W.; Clark, B.A.; West, J.T.

    1990-01-01

    The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. These codes were originally developed many years ago and have undergone continual improvement. With a large initial effort and continued vigilance, the codes are easily portable from one type of hardware to another. The performance of scientific work-stations (SWS) has evolved to the point that such platforms can be used routinely to perform sophisticated radiation transport calculations. As the personal computer (PC) performance approaches that of the SWS, the hardware options for desk-top radiation transport calculations expands considerably. The current status of the radiation transport codes within the LARTCS is described: MCNP, SABRINA, LAHET, ONEDANT, TWODANT, TWOHEX, and ONELD. Specifically, the authors discuss hardware systems on which the codes run and present code performance comparisons for various machines

  5. Radiation transport phenomena and modeling - part A: Codes

    International Nuclear Information System (INIS)

    Lorence, L.J.

    1997-01-01

    The need to understand how particle radiation (high-energy photons and electrons) from a variety of sources affects materials and electronics has motivated the development of sophisticated computer codes that describe how radiation with energies from 1.0 keV to 100.0 GeV propagates through matter. Predicting radiation transport is the necessary first step in predicting radiation effects. The radiation transport codes that are described here are general-purpose codes capable of analyzing a variety of radiation environments including those produced by nuclear weapons (x-rays, gamma rays, and neutrons), by sources in space (electrons and ions) and by accelerators (x-rays, gamma rays, and electrons). Applications of these codes include the study of radiation effects on electronics, nuclear medicine (imaging and cancer treatment), and industrial processes (food disinfestation, waste sterilization, manufacturing.) The primary focus will be on coupled electron-photon transport codes, with some brief discussion of proton transport. These codes model a radiation cascade in which electrons produce photons and vice versa. This coupling between particles of different types is important for radiation effects. For instance, in an x-ray environment, electrons are produced that drive the response in electronics. In an electron environment, dose due to bremsstrahlung photons can be significant once the source electrons have been stopped

  6. Available computer codes and data for radiation transport analysis

    International Nuclear Information System (INIS)

    Trubey, D.K.; Maskewitz, B.F.; Roussin, R.W.

    1975-01-01

    The Radiation Shielding Information Center (RSIC), sponsored and supported by the Energy Research and Development Administration (ERDA) and the Defense Nuclear Agency (DNA), is a technical institute serving the radiation transport and shielding community. It acquires, selects, stores, retrieves, evaluates, analyzes, synthesizes, and disseminates information on shielding and ionizing radiation transport. The major activities include: (1) operating a computer-based information system and answering inquiries on radiation analysis, (2) collecting, checking out, packaging, and distributing large computer codes, and evaluated and processed data libraries. The data packages include multigroup coupled neutron-gamma-ray cross sections and kerma coefficients, other nuclear data, and radiation transport benchmark problem results

  7. 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

  8. 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

  9. grmonty: A MONTE CARLO CODE FOR RELATIVISTIC RADIATIVE TRANSPORT

    International Nuclear Information System (INIS)

    Dolence, Joshua C.; Gammie, Charles F.; Leung, Po Kin; Moscibrodzka, Monika

    2009-01-01

    We describe a Monte Carlo radiative transport code intended for calculating spectra of hot, optically thin plasmas in full general relativity. The version we describe here is designed to model hot accretion flows in the Kerr metric and therefore incorporates synchrotron emission and absorption, and Compton scattering. The code can be readily generalized, however, to account for other radiative processes and an arbitrary spacetime. We describe a suite of test problems, and demonstrate the expected N -1/2 convergence rate, where N is the number of Monte Carlo samples. Finally, we illustrate the capabilities of the code with a model calculation, a spectrum of the slowly accreting black hole Sgr A* based on data provided by a numerical general relativistic MHD model of the accreting plasma.

  10. 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

  11. Academic Training - The use of Monte Carlo radiation transport codes in radiation physics and dosimetry

    CERN Multimedia

    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...

  12. The use of Monte Carlo radiation transport codes in radiation physics and dosimetry

    CERN Multimedia

    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...

  13. BALTORO a general purpose code for coupling discrete ordinates and Monte-Carlo radiation transport calculations

    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)

  14. Abstracts of digital computer code packages. Assembled by the Radiation Shielding Information Center. [Radiation transport codes

    Energy Technology Data Exchange (ETDEWEB)

    McGill, B.; Maskewitz, B.F.; Anthony, C.M.; Comolander, H.E.; Hendrickson, H.R.

    1976-01-01

    The term ''code package'' is used to describe a miscellaneous grouping of materials which, when interpreted in connection with a digital computer, enables the scientist--user to solve technical problems in the area for which the material was designed. In general, a ''code package'' consists of written material--reports, instructions, flow charts, listings of data, and other useful material and IBM card decks (or, more often, a reel of magnetic tape) on which the source decks, sample problem input (including libraries of data) and the BCD/EBCDIC output listing from the sample problem are written. In addition to the main code, and any available auxiliary routines are also included. The abstract format was chosen to give to a potential code user several criteria for deciding whether or not he wishes to request the code package. (RWR)

  15. Radiation transport phenomena and modeling. Part A: Codes; Part B: Applications with examples

    International Nuclear Information System (INIS)

    Lorence, L.J. Jr.; Beutler, D.E.

    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

  16. A Monte Carlo Code for Relativistic Radiation Transport Around Kerr Black Holes

    Science.gov (United States)

    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.

  17. 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)

  18. Accelerating execution of the integrated TIGER series Monte Carlo radiation transport codes

    Science.gov (United States)

    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).

  19. 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

  20. Computer codes in nuclear safety, radiation transport and dosimetry; Les codes de calcul en radioprotection, radiophysique et dosimetrie

    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.

  1. PBMC: Pre-conditioned Backward Monte Carlo code for radiative transport in planetary atmospheres

    Science.gov (United States)

    García Muñoz, A.; Mills, F. P.

    2017-08-01

    PBMC (Pre-Conditioned Backward Monte Carlo) solves the vector Radiative Transport Equation (vRTE) and can be applied to planetary atmospheres irradiated from above. The code builds the solution by simulating the photon trajectories from the detector towards the radiation source, i.e. in the reverse order of the actual photon displacements. In accounting for the polarization in the sampling of photon propagation directions and pre-conditioning the scattering matrix with information from the scattering matrices of prior (in the BMC integration order) photon collisions, PBMC avoids the unstable and biased solutions of classical BMC algorithms for conservative, optically-thick, strongly-polarizing media such as Rayleigh atmospheres.

  2. The Premar Code for the Monte Carlo Simulation of Radiation Transport In the Atmosphere

    International Nuclear Information System (INIS)

    Cupini, E.; Borgia, M.G.; Premuda, M.

    1997-03-01

    The Montecarlo code PREMAR is described, which allows the user to simulate the radiation transport in the atmosphere, in the ultraviolet-infrared frequency interval. A plan multilayer geometry is at present foreseen by the code, witch albedo possibility at the lower boundary surface. For a given monochromatic point source, the main quantities computed by the code are the absorption spatial distributions of aerosol and molecules, together with the related atmospheric transmittances. Moreover, simulation of of Lidar experiments are foreseen by the code, the source and telescope fields of view being assigned. To build-up the appropriate probability distributions, an input data library is assumed to be read by the code. For this purpose the radiance-transmittance LOWTRAN-7 code has been conveniently adapted as a source of the library so as to exploit the richness of information of the code for a large variety of atmospheric simulations. Results of applications of the PREMAR code are finally presented, with special reference to simulations of Lidar system and radiometer experiments carried out at the Brasimone ENEA Centre by the Environment Department

  3. 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

  4. 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)

  5. GRAVE: An Interactive Geometry Construction and Visualization Software System for the TORT Nuclear Radiation Transport Code

    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

  6. Implementation, capabilities, and benchmarking of Shift, a massively parallel Monte Carlo radiation transport code

    International Nuclear Information System (INIS)

    Pandya, Tara M.; Johnson, Seth R.; Evans, Thomas M.; Davidson, Gregory G.; Hamilton, Steven P.; Godfrey, Andrew T.

    2015-01-01

    This paper discusses the implementation, capabilities, and validation of Shift, a massively parallel Monte Carlo radiation transport package developed and maintained at Oak Ridge National Laboratory. It has been developed to scale well from laptop to small computing clusters to advanced supercomputers. Special features of Shift include hybrid capabilities for variance reduction such as CADIS and FW-CADIS, and advanced parallel decomposition and tally methods optimized for scalability on supercomputing architectures. Shift has been validated and verified against various reactor physics benchmarks and compares well to other state-of-the-art Monte Carlo radiation transport codes such as MCNP5, CE KENO-VI, and OpenMC. Some specific benchmarks used for verification and validation include the CASL VERA criticality test suite and several Westinghouse AP1000 ® problems. These benchmark and scaling studies show promising results

  7. AREVA Developments for an Efficient and Reliable use of Monte Carlo codes for Radiation Transport Applications

    Science.gov (United States)

    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.

  8. AREVA Developments for an Efficient and Reliable use of Monte Carlo codes for Radiation Transport Applications

    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.

  9. TIERCE: A code system for particles and radiation transport in thick targets

    Energy Technology Data Exchange (ETDEWEB)

    Bersillon, O.; Bauge, E.; Borne, F.; Clergeau, J.F.; Collin, M.; Cotten, D.; Delaroche, J.P.; Duarte, H.; Flament, J.L.; Girod, M.; Gosselin, G.; Granier, T.; Hilaire, S.; Morel, P.; Perrier, R.; Romain, P.; Roux, L. [CEA, Bruyeres-le-Chatel (France). Service de Physique Nucleaire

    1997-09-01

    Over the last few years, a great effort at Bruyeres-le-Chatel has been the development of the TIERCE code system for the transport of particles and radiations in complex geometry. The comparison of calculated results with experimental data, either microscopic (double differential spectra, residual nuclide yield...) or macroscopic (energy deposition, neutron leakage...), shows the need to improve the nuclear reaction models used. We present some new developments concerning data required for the evaporation model in the framework of a microscopic approach. 22 refs., 6 figs.

  10. Comparison of Radiation Transport Codes, HZETRN, HETC and FLUKA, Using the 1956 Webber SPE Spectrum

    Science.gov (United States)

    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.; hide

    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.

  11. 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)

  12. The effect on radiation damage of structural material in a hybrid system by using a Monte Carlo radiation transport code

    International Nuclear Information System (INIS)

    Günay, Mehtap; Şarer, Başar; Kasap, Hızır

    2014-01-01

    Highlights: • The effects of some fluids on gas production rates in structural material were investigated. • The MCNPX-2.7.0 Monte Carlo code was used for three-dimensional calculations. • It was found that biggest contribution to gas production rates comes from Fe isotope of the. • The desirable values for 5% SFG-PuO 2 with respect to radiation damage were specified. - Abstract: In this study, the molten salt-heavy metal mixtures 99–95% Li20Sn80-1-5% SFG-Pu, 99–95% Li20Sn80-1-5% SFG-PuF4, 99-95% Li20Sn80-1-5% SFG-PuO2 were used as fluids. The fluids were used in the liquid first-wall, blanket and shield zones of the designed hybrid reactor system. 9Cr2WVTa ferritic steel with the width of 4 cm was used as the structural material. The parameters of radiation damage are proton, deuterium, tritium, He-3 and He-4 gas production rates. In this study, the effects of the selected fluid on the radiation damage, in terms of individual as well as total isotopes in the structural material, were investigated for 30 full power years (FPYs). Three-dimensional analyses were performed using the most recent version of the MCNPX-2.7.0 Monte Carlo radiation transport code and the ENDF/B-VII.0 nuclear data library

  13. Radiation Transport

    Energy Technology Data Exchange (ETDEWEB)

    Urbatsch, Todd James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-06-15

    We present an overview of radiation transport, covering terminology, blackbody raditation, opacities, Boltzmann transport theory, approximations to the transport equation. Next we introduce several transport methods. We present a section on Caseology, observing transport boundary layers. We briefly broach topics of software development, including verification and validation, and we close with a section on high energy-density experiments that highlight and support radiation transport.

  14. 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

  15. C5 Benchmark Problem with Discrete Ordinate Radiation Transport Code DENOVO

    Energy Technology Data Exchange (ETDEWEB)

    Yesilyurt, Gokhan [ORNL; Clarno, Kevin T [ORNL; Evans, Thomas M [ORNL; Davidson, Gregory G [ORNL; Fox, Patricia B [ORNL

    2011-01-01

    The C5 benchmark problem proposed by the Organisation for Economic Co-operation and Development/Nuclear Energy Agency was modeled to examine the capabilities of Denovo, a three-dimensional (3-D) parallel discrete ordinates (S{sub N}) radiation transport code, for problems with no spatial homogenization. Denovo uses state-of-the-art numerical methods to obtain accurate solutions to the Boltzmann transport equation. Problems were run in parallel on Jaguar, a high-performance supercomputer located at Oak Ridge National Laboratory. Both the two-dimensional (2-D) and 3-D configurations were analyzed, and the results were compared with the reference MCNP Monte Carlo calculations. For an additional comparison, SCALE/KENO-V.a Monte Carlo solutions were also included. In addition, a sensitivity analysis was performed for the optimal angular quadrature and mesh resolution for both the 2-D and 3-D infinite lattices of UO{sub 2} fuel pin cells. Denovo was verified with the C5 problem. The effective multiplication factors, pin powers, and assembly powers were found to be in good agreement with the reference MCNP and SCALE/KENO-V.a Monte Carlo calculations.

  16. Development of Parallel Computing Framework to Enhance Radiation Transport Code Capabilities for Rare Isotope Beam Facility Design

    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.

  17. Analysis of the pool critical assembly benchmark using raptor-M3G, a parallel deterministic radiation transport code - 289

    International Nuclear Information System (INIS)

    Fischer, G.A.

    2010-01-01

    The PCA Benchmark is analyzed using RAPTOR-M3G, a parallel SN radiation transport code. A variety of mesh structures, angular quadrature sets, cross section treatments, and reactor dosimetry cross sections are presented. The results show that RAPTOR-M3G is generally suitable for PWR neutron dosimetry applications. (authors)

  18. RADHEAT-V4: a code system to generate multigroup constants and analyze radiation transport for shielding safety evaluation

    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)

  19. Activities of the Radiation Shielding Information Center and a report on codes/data for high energy radiation transport

    International Nuclear Information System (INIS)

    Roussin, R.W.

    1993-01-01

    From the very early days in its history Radiation Shielding Information Center (RSIC) has been involved with high energy radiation transport. The National Aeronautics and Space Administration was an early sponsor of RSIC until the completion of the Apollo Moon Exploration Program. In addition, the intranuclear cascade work of Bertini at Oak Ridge National Laboratory provided valuable resources which were made available through RSIC. Over the years, RSIC has had interactions with many of the developers of high energy radiation transport computing technology and data libraries and has been able to collect and disseminate this technology. The current status of this technology will be reviewed and prospects for new advancements will be examined

  20. Utilization of a photon transport code to investigate radiation therapy treatment planning quantities and techniques

    International Nuclear Information System (INIS)

    Palta, J.R.

    1981-01-01

    A versatile computer program MORSE, based on neutron and photon transport theory has been utilzed to investigate radiation therapy treatment planning quantities and techniques. A multi-energy group representation of transport equation provides a concise approach in utilizing Monte Carlo numerical techniques to multiple radiation therapy treatment planning problems. Central axis total and scattered dose distributions for homogeneous and inhomogeneous water phantoms are calculated and the correction factor for lung and bone inhomogeneities are also evaluated. Results show that Monte Carlo calculations based on multi-energy group tansport theory predict the depth dose distributions that are in good agreement with available experimental data. Central axis depth dose distributions for a bremsstrahlung spectrum from a linear accelerator is also calculated to exhibit the versatility of the computer program in handling multiple radiation therapy problems. A novel approach is undertaken to study the dosimetric properties of brachytherapy sources

  1. A Monte Carlo transport code study of the space radiation environment using FLUKA and ROOT

    CERN Document Server

    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...

  2. 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)

  3. MORSE-CGT Monte Carlo radiation transport code with the capability of the torus geometric treatment

    International Nuclear Information System (INIS)

    Deng Li

    1990-01-01

    The combinatorial geometry package CGT with the capability of the torus geometric treatment is introduced. It is get by developing the combinatorial geometry package CG. The CGT package can be transplanted to those codes which the CG package is being used and makes them also with the capability. The MORSE-CGT code can be used to solve the neutron, gamma-ray or coupled neutron-gamma-ray transport problems and time dependence for both shielding and criticality problems in torus system or system which is produced by arbitrary finite combining torus with torus or other bodies in CG package and it can also be used to design the blanket and compute shielding for TOKAMAK Fusion-Fission Hybrid Reactor

  4. NEACRP comparison of codes for the radiation protection assessment of transportation packages. Solutions to problems 1 - 4

    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

  5. Review of Hybrid (Deterministic/Monte Carlo) Radiation Transport Methods, Codes, and Applications at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Wagner, John C.; Peplow, Douglas E.; Mosher, Scott W.; Evans, Thomas M.

    2010-01-01

    This paper provides a review of the hybrid (Monte Carlo/deterministic) radiation transport methods and codes used at the Oak Ridge National Laboratory and examples of their application for increasing the efficiency of real-world, fixed-source Monte Carlo analyses. The two principal hybrid methods are (1) Consistent Adjoint Driven Importance Sampling (CADIS) for optimization of a localized detector (tally) region (e.g., flux, dose, or reaction rate at a particular location) and (2) Forward Weighted CADIS (FW-CADIS) for optimizing distributions (e.g., mesh tallies over all or part of the problem space) or multiple localized detector regions (e.g., simultaneous optimization of two or more localized tally regions). The two methods have been implemented and automated in both the MAVRIC sequence of SCALE 6 and ADVANTG, a code that works with the MCNP code. As implemented, the methods utilize the results of approximate, fast-running 3-D discrete ordinates transport calculations (with the Denovo code) to generate consistent space- and energy-dependent source and transport (weight windows) biasing parameters. These methods and codes have been applied to many relevant and challenging problems, including calculations of PWR ex-core thermal detector response, dose rates throughout an entire PWR facility, site boundary dose from arrays of commercial spent fuel storage casks, radiation fields for criticality accident alarm system placement, and detector response for special nuclear material detection scenarios and nuclear well-logging tools. Substantial computational speed-ups, generally O(10 2-4 ), have been realized for all applications to date. This paper provides a brief review of the methods, their implementation, results of their application, and current development activities, as well as a considerable list of references for readers seeking more information about the methods and/or their applications.

  6. Review of Hybrid (Deterministic/Monte Carlo) Radiation Transport Methods, Codes, and Applications at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Wagner, John C.; Peplow, Douglas E.; Mosher, Scott W.; Evans, Thomas M.

    2010-01-01

    This paper provides a review of the hybrid (Monte Carlo/deterministic) radiation transport methods and codes used at the Oak Ridge National Laboratory and examples of their application for increasing the efficiency of real-world, fixed-source Monte Carlo analyses. The two principal hybrid methods are (1) Consistent Adjoint Driven Importance Sampling (CADIS) for optimization of a localized detector (tally) region (e.g., flux, dose, or reaction rate at a particular location) and (2) Forward Weighted CADIS (FW-CADIS) for optimizing distributions (e.g., mesh tallies over all or part of the problem space) or multiple localized detector regions (e.g., simultaneous optimization of two or more localized tally regions). The two methods have been implemented and automated in both the MAVRIC sequence of SCALE 6 and ADVANTG, a code that works with the MCNP code. As implemented, the methods utilize the results of approximate, fast-running 3-D discrete ordinates transport calculations (with the Denovo code) to generate consistent space- and energy-dependent source and transport (weight windows) biasing parameters. These methods and codes have been applied to many relevant and challenging problems, including calculations of PWR ex-core thermal detector response, dose rates throughout an entire PWR facility, site boundary dose from arrays of commercial spent fuel storage casks, radiation fields for criticality accident alarm system placement, and detector response for special nuclear material detection scenarios and nuclear well-logging tools. Substantial computational speed-ups, generally O(102-4), have been realized for all applications to date. This paper provides a brief review of the methods, their implementation, results of their application, and current development activities, as well as a considerable list of references for readers seeking more information about the methods and/or their applications.

  7. Review of hybrid (deterministic/Monte Carlo) radiation transport methods, codes, and applications at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Wagner, J.C.; Peplow, D.E.; Mosher, S.W.; Evans, T.M.

    2010-01-01

    This paper provides a review of the hybrid (Monte Carlo/deterministic) radiation transport methods and codes used at the Oak Ridge National Laboratory and examples of their application for increasing the efficiency of real-world, fixed-source Monte Carlo analyses. The two principal hybrid methods are (1) Consistent Adjoint Driven Importance Sampling (CADIS) for optimization of a localized detector (tally) region (e.g., flux, dose, or reaction rate at a particular location) and (2) Forward Weighted CADIS (FW-CADIS) for optimizing distributions (e.g., mesh tallies over all or part of the problem space) or multiple localized detector regions (e.g., simultaneous optimization of two or more localized tally regions). The two methods have been implemented and automated in both the MAVRIC sequence of SCALE 6 and ADVANTG, a code that works with the MCNP code. As implemented, the methods utilize the results of approximate, fast-running 3-D discrete ordinates transport calculations (with the Denovo code) to generate consistent space- and energy-dependent source and transport (weight windows) biasing parameters. These methods and codes have been applied to many relevant and challenging problems, including calculations of PWR ex-core thermal detector response, dose rates throughout an entire PWR facility, site boundary dose from arrays of commercial spent fuel storage casks, radiation fields for criticality accident alarm system placement, and detector response for special nuclear material detection scenarios and nuclear well-logging tools. Substantial computational speed-ups, generally O(10 2-4 ), have been realized for all applications to date. This paper provides a brief review of the methods, their implementation, results of their application, and current development activities, as well as a considerable list of references for readers seeking more information about the methods and/or their applications. (author)

  8. Transport theory and codes

    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

  9. Reactor Dosimetry Applications Using RAPTOR-M3G:. a New Parallel 3-D Radiation Transport Code

    Science.gov (United States)

    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.

  10. Radiation transport and shielding information, computer codes, and nuclear data for use in CTR neutronics research and development

    International Nuclear Information System (INIS)

    Santoro, R.T.; Maskewitz, B.F.; Roussin, R.W.; Trubey, D.K.

    1976-01-01

    The activities of the Radiation Shielding Information Center (RSIC) of the Oak Ridge National Laboratory are being utilized in support of fusion reactor technology. The major activities of RSIC include the operation of a computer-based information storage and retrieval system, the collection, packaging, and distribution of large computer codes, and the compilation and dissemination of processed and evaluated data libraries, with particular emphasis on neutron and gamma-ray cross-section data. The Center has acquired thirteen years of experience in serving fission reactor, weapons, and accelerator shielding research communities, and the extension of its technical base to fusion reactor research represents a logical progression. RSIC is currently working with fusion reactor researchers and contractors in computer code development to provide tested radiation transport and shielding codes and data library packages. Of significant interest to the CTR community are the 100 energy group neutron and 21 energy group gamma-ray coupled cross-section data package (DLC-37) for neutronics studies, a comprehensive 171 energy group neutron and 36 energy group gamma-ray coupled cross-section data base with retrieval programs, including resonance self-shielding, that are tailored to CTR application, and a data base for the generation of energy-dependent atomic displacement and gas production cross sections and heavy-particle-recoil spectra for estimating radiation damage to CTR structural components

  11. Calculation codes in radiation protection, radiation physics and dosimetry

    International Nuclear Information System (INIS)

    2003-01-01

    These scientific days had for objective to draw up the situation of calculation codes of radiation transport, of sources estimation, of radiation doses managements and to draw the future perspectives. (N.C.)

  12. Voxel2MCNP: a framework for modeling, simulation and evaluation of radiation transport scenarios for Monte Carlo codes

    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)

  13. Radiation transport and shielding information, computer codes, and nuclear data for use in CTR neutronics research and development

    International Nuclear Information System (INIS)

    Santoro, R.T.; Maskewitz, B.F.; Roussin, R.W.; Trubey, D.K.

    1976-01-01

    The activities of the Radiation Shielding Information Center (RSIC) of the Oak Ridge National Laboratory are being utilized in support of fusion reactor technology. The major activities of RSIC include the operation of a computer-based information storage and retrieval system, the collection, packaging, and distribution of large computer codes, and the compilation and dissemination of processed and evaluated data libraries, with particular emphasis on neutron and gamma-ray cross-section data. The Center has acquired thirteen years of experience in serving fission reactor, weapons, and accelerator shielding research communities, and the extension of its technical base to fusion reactor research represents a logical progression. RSIC is currently working with fusion reactor researchers and contractors in computer code development to provide tested radiation transport and shielding codes and data library packages. Of significant interest to the CTR community are the 100 energy group neutron and 21 energy group gamma-ray coupled cross-section data package (DLC-37) for neutronics studies, a comprehensive 171 energy group neutron and 36 energy group gamma-ray coupled cross-section data base with retrieval programs, including resonance self-shielding, that are tailored to CTR application, and a data base for the generation of energy-dependent atomic displacement and gas production cross sections and heavy-particle-recoil spectra for estimating radiation damage to CTR structural components. Since 1964, the Center has been involved in the international exchange of information, encouraged and supported by both government and interagency agreements; and to achieve an equally viable and successful program in fusion research, the reciprocal exchange of CTR data and computing technology is encouraged and welcomed

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

    International Nuclear Information System (INIS)

    Raisali, G.R.

    1992-01-01

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

  15. Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code

    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

  16. Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code

    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

  17. Forms of Approximate Radiation Transport

    CERN Document Server

    Brunner, G

    2002-01-01

    Photon radiation transport is described by the Boltzmann equation. Because this equation is difficult to solve, many different approximate forms have been implemented in computer codes. Several of the most common approximations are reviewed, and test problems illustrate the characteristics of each of the approximations. This document is designed as a tutorial so that code users can make an educated choice about which form of approximate radiation transport to use for their particular simulation.

  18. NEACRP comparison of source term codes for the radiation protection assessment of transportation packages

    International Nuclear Information System (INIS)

    Broadhead, B.L.; Locke, H.F.; Avery, A.F.

    1994-01-01

    The results for Problems 5 and 6 of the NEACRP code comparison as submitted by six participating countries are presented in summary. These problems concentrate on the prediction of the neutron and gamma-ray sources arising in fuel after a specified irradiation, the fuel being uranium oxide for problem 5 and a mixture of uranium and plutonium oxides for problem 6. In both problems the predicted neutron sources are in good agreement for all participants. For gamma rays, however, there are differences, largely due to the omission of bremsstrahlung in some calculations

  19. High Energy Transport Code HETC

    International Nuclear Information System (INIS)

    Gabriel, T.A.

    1985-09-01

    The physics contained in the High Energy Transport Code (HETC), in particular the collision models, are discussed. An application using HETC as part of the CALOR code system is also given. 19 refs., 5 figs., 3 tabs

  20. Sub-Transport Layer Coding

    DEFF Research Database (Denmark)

    Hansen, Jonas; Krigslund, Jeppe; Roetter, Daniel Enrique Lucani

    2014-01-01

    Packet losses in wireless networks dramatically curbs the performance of TCP. This paper introduces a simple coding shim that aids IP-layer traffic in lossy environments while being transparent to transport layer protocols. The proposed coding approach enables erasure correction while being...... 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...

  1. Coupling External Radiation Transport Code Results to the GADRAS Detector Response Function

    International Nuclear Information System (INIS)

    Mitchell, Dean J.; Thoreson, Gregory G.; Horne, Steven M.

    2014-01-01

    Simulating gamma spectra is useful for analyzing special nuclear materials. Gamma spectra are influenced not only by the source and the detector, but also by the external, and potentially complex, scattering environment. The scattering environment can make accurate representations of gamma spectra difficult to obtain. By coupling the Monte Carlo Nuclear Particle (MCNP) code with the Gamma Detector Response and Analysis Software (GADRAS) detector response function, gamma spectrum simulations can be computed with a high degree of fidelity even in the presence of a complex scattering environment. Traditionally, GADRAS represents the external scattering environment with empirically derived scattering parameters. By modeling the external scattering environment in MCNP and using the results as input for the GADRAS detector response function, gamma spectra can be obtained with a high degree of fidelity. This method was verified with experimental data obtained in an environment with a significant amount of scattering material. The experiment used both gamma-emitting sources and moderated and bare neutron-emitting sources. The sources were modeled using GADRAS and MCNP in the presence of the external scattering environment, producing accurate representations of the experimental data.

  2. Radiation transport in MEDUSA

    International Nuclear Information System (INIS)

    Rose, S.J.; Evans, R.G.

    1983-09-01

    The transport of energy by X-ray photons has been included in the lD Lagrangian hydrodynamics code, MEDUSA. Calculations of the implosion by 0.53 μm laser irradiation of plastic and glass microballoons of current interest at the Central Laser Facility show that radiation preheats the fill gas and alters the temperature and density profiles during the implosion. A lower maximum gas temperature is obtained and this results, for a DT gas fill, in a greatly reduced neutron yield. (author)

  3. The Premar Code for the Monte Carlo Simulation of Radiation Transport In the Atmosphere; Il codice PREMAR per la simulazione Montecarlo del trasporto della radiazione dell`atmosfera

    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.

  4. RADHEAT-V3, a code system for generating coupled neutron and gamma-ray group constants and analyzing radiation transport

    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.)

  5. Calculation codes in radiation protection, radiation physics and dosimetry; Codes de calcul en radioprotection, radiophysique et dosimetrie

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-07-01

    These scientific days had for objective to draw up the situation of calculation codes of radiation transport, of sources estimation, of radiation doses managements and to draw the future perspectives. (N.C.)

  6. The RAGE radiation-hydrodynamic code

    Energy Technology Data Exchange (ETDEWEB)

    Gittings, Michael; Clover, Michael; Betlach, Thomas; Byrne, Nelson; Ranta, Dale [Science Applications International Corp. MS A-1, 10260 Campus Point Drive, San Diego, CA 92121 (United States); Weaver, Robert; Coker, Robert; Dendy, Edward; Hueckstaedt, Robert; New, Kim; Oakes, W Rob [Los Alamos National Laboratory, MS T087, PO Box 1663, Los Alamos, NM 87545 (United States); Stefan, Ryan [TaylorMade-adidas Golf, 5545 Fermi Court, Carlsbad, CA 92008-7324 (United States)], E-mail: michael.r.clover@saic.com

    2008-10-01

    We describe RAGE, the 'radiation adaptive grid Eulerian' radiation-hydrodynamics code, including its data structures, its parallelization strategy and performance, its hydrodynamic algorithm(s), its (gray) radiation diffusion algorithm, and some of the considerable amount of verification and validation efforts. The hydrodynamics is a basic Godunov solver, to which we have made significant improvements to increase the advection algorithm's robustness and to converge stiffnesses in the equation of state. Similarly, the radiation transport is a basic gray diffusion, but our treatment of the radiation-material coupling, wherein we converge nonlinearities in a novel manner to allow larger timesteps and more robust behavior, can be applied to any multi-group transport algorithm.

  7. The RAGE radiation-hydrodynamic code

    International Nuclear Information System (INIS)

    Gittings, Michael; Clover, Michael; Betlach, Thomas; Byrne, Nelson; Ranta, Dale; Weaver, Robert; Coker, Robert; Dendy, Edward; Hueckstaedt, Robert; New, Kim; Oakes, W Rob; Stefan, Ryan

    2008-01-01

    We describe RAGE, the 'radiation adaptive grid Eulerian' radiation-hydrodynamics code, including its data structures, its parallelization strategy and performance, its hydrodynamic algorithm(s), its (gray) radiation diffusion algorithm, and some of the considerable amount of verification and validation efforts. The hydrodynamics is a basic Godunov solver, to which we have made significant improvements to increase the advection algorithm's robustness and to converge stiffnesses in the equation of state. Similarly, the radiation transport is a basic gray diffusion, but our treatment of the radiation-material coupling, wherein we converge nonlinearities in a novel manner to allow larger timesteps and more robust behavior, can be applied to any multi-group transport algorithm

  8. 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

  9. CTCN: Colloid transport code -- nuclear

    International Nuclear Information System (INIS)

    Jain, R.

    1993-01-01

    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-algebraic 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-algebraic systems

  10. Introduction to radiation transport

    International Nuclear Information System (INIS)

    Olson, G.L.

    1998-01-01

    This lecture will present time-dependent radiation transport where the radiation is coupled to a static medium, i.e., the material is not in motion. In reality, radiation exerts a pressure on the materials it propagates through and will accelerate the material in the direction of the radiation flow. This fully coupled problem with radiation transport and materials in motion is referred to as radiation-hydrodynamics (or in a shorthand notation: rad-hydro) and is beyond the scope of this lecture

  11. Many-integrated core (MIC) technology for accelerating Monte Carlo simulation of radiation transport: A study based on the code DPM

    Science.gov (United States)

    Rodriguez, M.; Brualla, L.

    2018-04-01

    Monte Carlo simulation of radiation transport is computationally demanding to obtain reasonably low statistical uncertainties of the estimated quantities. Therefore, it can benefit in a large extent from high-performance computing. This work is aimed at assessing the performance of the first generation of the many-integrated core architecture (MIC) Xeon Phi coprocessor with respect to that of a CPU consisting of a double 12-core Xeon processor in Monte Carlo simulation of coupled electron-photonshowers. The comparison was made twofold, first, through a suite of basic tests including parallel versions of the random number generators Mersenne Twister and a modified implementation of RANECU. These tests were addressed to establish a baseline comparison between both devices. Secondly, through the p DPM code developed in this work. p DPM is a parallel version of the Dose Planning Method (DPM) program for fast Monte Carlo simulation of radiation transport in voxelized geometries. A variety of techniques addressed to obtain a large scalability on the Xeon Phi were implemented in p DPM. Maximum scalabilities of 84 . 2 × and 107 . 5 × were obtained in the Xeon Phi for simulations of electron and photon beams, respectively. Nevertheless, in none of the tests involving radiation transport the Xeon Phi performed better than the CPU. The disadvantage of the Xeon Phi with respect to the CPU owes to the low performance of the single core of the former. A single core of the Xeon Phi was more than 10 times less efficient than a single core of the CPU for all radiation transport simulations.

  12. Application of the three-dimensional Oak Ridge transport code

    International Nuclear Information System (INIS)

    Rhoades, W.A.; Childs, R.L.; Emmett, M.B.; Cramer, S.N.

    1984-01-01

    TORT, a 3-d extension of the DOT discrete ordinates transport code is now in production use for studies of radiation penetration into large concrete and masonry structures. This paper discusses certain features of the new code and shows representative results, including comparisons with Monte Carlo calculations

  13. Coupled electron-photon radiation transport

    International Nuclear Information System (INIS)

    Lorence, L.; Kensek, R.P.; Valdez, G.D.; Drumm, C.R.; Fan, W.C.; Powell, J.L.

    2000-01-01

    Massively-parallel computers allow detailed 3D radiation transport simulations to be performed to analyze the response of complex systems to radiation. This has been recently been demonstrated with the coupled electron-photon Monte Carlo code, ITS. To enable such calculations, the combinatorial geometry capability of ITS was improved. For greater geometrical flexibility, a version of ITS is under development that can track particles in CAD geometries. Deterministic radiation transport codes that utilize an unstructured spatial mesh are also being devised. For electron transport, the authors are investigating second-order forms of the transport equations which, when discretized, yield symmetric positive definite matrices. A novel parallelization strategy, simultaneously solving for spatial and angular unknowns, has been applied to the even- and odd-parity forms of the transport equation on a 2D unstructured spatial mesh. Another second-order form, the self-adjoint angular flux transport equation, also shows promise for electron transport

  14. BOT3P5.2, 3D Mesh Generator and Graphical Display of Geometry for Radiation Transport Codes, Display of Results

    International Nuclear Information System (INIS)

    Orsi, Roberto; Bidaud, Adrien

    2007-01-01

    1 - Description of program or function: BOT3P was originally conceived as a set of standard FORTRAN 77 language programs in order to give the users of the DORT and TORT deterministic transport codes some useful diagnostic tools to prepare and check their input data files. Later versions extended the possibility to produce the geometrical, material distribution and fixed neutron source data to other deterministic transport codes such as TWODANT/THREEDANT of the DANTSYS system, PARTISN and, potentially, to any transport code through BOT3P binary output files that can be easily interfaced (see, for example, the Russian two-dimensional (2D) and three-dimensional (3D) discrete ordinates neutron, photon and charged particle transport codes KASKAD-S-2.5 and KATRIN-2.0). As from Version 5.1 BOT3P contained important additions specifically addressed to radiation transport analysis for medical applications. BOT3P-5.2 contains new graphics capabilities. Some of them enable users to select space sub-domains of the total mesh grid in order to improve the zoom simulation of the geometry, both in 2D cuts and in 3D. Moreover the new BOT3P module (PDTM) may improve the interface of BOT3P geometrical models to transport analysis codes. The following programs are included in the BOT3P software package: GGDM, DDM, GGTM, DTM2, DTM3, RVARSCL, COMPARE, MKSRC, CATSM, DTET, and PDTM. The main features of these different programs are described. 2 - Methods: GGDM and GGTM work similarly from the logical point of view. Since the 3D case is more general, the following description refers to GGTM. All the co-ordinate values that characterise the geometrical scheme at the basis of the 3D transport code geometrical and material model are read, sorted and all stored if different from the neighbouring ones more than an input tolerance established by the user. These co-ordinates are always present in the fine-mesh boundary arrays independently of the mesh grid refinement options, because they

  15. The computer code system for reactor radiation shielding in design of nuclear power plant

    International Nuclear Information System (INIS)

    Li Chunhuai; Fu Shouxin; Liu Guilian

    1995-01-01

    The computer code system used in reactor radiation shielding design of nuclear power plant includes the source term codes, discrete ordinate transport codes, Monte Carlo and Albedo Monte Carlo codes, kernel integration codes, optimization code, temperature field code, skyshine code, coupling calculation codes and some processing codes for data libraries. This computer code system has more satisfactory variety of codes and complete sets of data library. It is widely used in reactor radiation shielding design and safety analysis of nuclear power plant and other nuclear facilities

  16. PHITS-a particle and heavy ion transport code system

    International Nuclear Information System (INIS)

    Niita, Koji; Sato, Tatsuhiko; Iwase, Hiroshi; Nose, Hiroyuki; Nakashima, Hiroshi; Sihver, Lembit

    2006-01-01

    The paper presents a summary of the recent development of the multi-purpose Monte Carlo Particle and Heavy Ion Transport code System, PHITS. In particular, we discuss in detail the development of two new models, JAM and JQMD, for high energy particle interactions, incorporated in PHITS, and show comparisons between model calculations and experiments for the validations of these models. The paper presents three applications of the code including spallation neutron source, heavy ion therapy and space radiation. The results and examples shown indicate PHITS has great ability of carrying out the radiation transport analysis of almost all particles including heavy ions within a wide energy range

  17. 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

  18. A multigroup treatment of radiation transport

    International Nuclear Information System (INIS)

    Tahir, N.A.; Laing, E.W.; Nicholas, D.J.

    1980-12-01

    A multi-group radiation package is outlined which will accurately handle radiation transfer problems in laser-produced plasmas. Bremsstrahlung, recombination and line radiation are included as well as fast electron Bremsstrahlung radiation. The entire radiation field is divided into a large number of groups (typically 20), which diffuse radiation energy in real space as well as in energy space, the latter occurring via electron-radiation interaction. Using this model a radiation transport code will be developed to be incorporated into MEDUSA. This modified version of MEDUSA will be used to study radiative preheat effects in laser-compression experiments at the Central Laser Facility, Rutherford Laboratory. The model is also relevant to heavy ion fusion studies. (author)

  19. Code of practice for ionizing radiation

    International Nuclear Information System (INIS)

    Khoo Boo Huat

    1995-01-01

    Prior to 1984, the use of ionizing radiation in Malaysia was governed by the Radioactive Substances Act of 1968. After 1984, its use came under the control of Act 304, called the Atomic Energy Licensing Act 1984. Under powers vested by the Act, the Radiation Protection (Basic Safety Standards) Regulations 1988 were formulated to regulate its use. These Acts do not provide information on proper working procedures. With the publication of the codes of Practice by The Standards and Industrial Research Institute of Malaysia (SIRIM), the users are now able to follow proper guidelines and use ionizing radiation safely and beneficially. This paper discusses the relevant sections in the following codes: 1. Code of Practice for Radiation Protection (Medical X-ray Diagnosis) MS 838:1983. 2. Code of Practice for Safety in Laboratories Part 4: Ionizing radiation MS 1042: Part 4: 1992. (author)

  20. Extension of ANISN and DOT 3.5 transport computer codes to calculate heat generation by radiation and temperature distribution in nuclear reactors

    International Nuclear Information System (INIS)

    Torres, L.M.R.; Gomes, I.C.; Maiorino, J.R.

    1986-01-01

    The ANISN and DOT 3.5 codes solve the transport equation using the discrete ordinate method, in one and two-dimensions, respectively. The objectives of the study were to modify these two codes, frequently used in reactor shielding problems, to include nuclear heating calculations due to the interaction of neutrons and gamma-rays with matter. In order to etermine the temperature distribution, a numerical algorithm was developed using the finite difference method to solve the heat conduction equation, in one and two-dimensions, considering the nuclear heating from neutron and gamma-rays, as the source term. (Author) [pt

  1. 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

  2. Generic programming for deterministic neutron transport codes

    International Nuclear Information System (INIS)

    Plagne, L.; Poncot, A.

    2005-01-01

    This paper discusses the implementation of neutron transport codes via generic programming techniques. Two different Boltzmann equation approximations have been implemented, namely the Sn and SPn methods. This implementation experiment shows that generic programming allows us to improve maintainability and readability of source codes with no performance penalties compared to classical approaches. In the present implementation, matrices and vectors as well as linear algebra algorithms are treated separately from the rest of source code and gathered in a tool library called 'Generic Linear Algebra Solver System' (GLASS). Such a code architecture, based on a linear algebra library, allows us to separate the three different scientific fields involved in transport codes design: numerical analysis, reactor physics and computer science. Our library handles matrices with optional storage policies and thus applies both to Sn code, where the matrix elements are computed on the fly, and to SPn code where stored matrices are used. Thus, using GLASS allows us to share a large fraction of source code between Sn and SPn implementations. Moreover, the GLASS high level of abstraction allows the writing of numerical algorithms in a form which is very close to their textbook descriptions. Hence the GLASS algorithms collection, disconnected from computer science considerations (e.g. storage policy), is very easy to read, to maintain and to extend. (authors)

  3. Parallel thermal radiation transport in two dimensions

    International Nuclear Information System (INIS)

    Smedley-Stevenson, R.P.; Ball, S.R.

    2003-01-01

    This paper describes the distributed memory parallel implementation of a deterministic thermal radiation transport algorithm in a 2-dimensional ALE hydrodynamics code. The parallel algorithm consists of a variety of components which are combined in order to produce a state of the art computational capability, capable of solving large thermal radiation transport problems using Blue-Oak, the 3 Tera-Flop MPP (massive parallel processors) computing facility at AWE (United Kingdom). Particular aspects of the parallel algorithm are described together with examples of the performance on some challenging applications. (author)

  4. Parallel thermal radiation transport in two dimensions

    Energy Technology Data Exchange (ETDEWEB)

    Smedley-Stevenson, R.P.; Ball, S.R. [AWE Aldermaston (United Kingdom)

    2003-07-01

    This paper describes the distributed memory parallel implementation of a deterministic thermal radiation transport algorithm in a 2-dimensional ALE hydrodynamics code. The parallel algorithm consists of a variety of components which are combined in order to produce a state of the art computational capability, capable of solving large thermal radiation transport problems using Blue-Oak, the 3 Tera-Flop MPP (massive parallel processors) computing facility at AWE (United Kingdom). Particular aspects of the parallel algorithm are described together with examples of the performance on some challenging applications. (author)

  5. Coupled geochemical and solute transport code development

    International Nuclear Information System (INIS)

    Morrey, J.R.; Hostetler, C.J.

    1985-01-01

    A number of coupled geochemical hydrologic codes have been reported in the literature. Some of these codes have directly coupled the source-sink term to the solute transport equation. The current consensus seems to be that directly coupling hydrologic transport and chemical models through a series of interdependent differential equations is not feasible for multicomponent problems with complex geochemical processes (e.g., precipitation/dissolution reactions). A two-step process appears to be the required method of coupling codes for problems where a large suite of chemical reactions must be monitored. Two-step structure requires that the source-sink term in the transport equation is supplied by a geochemical code rather than by an analytical expression. We have developed a one-dimensional two-step coupled model designed to calculate relatively complex geochemical equilibria (CTM1D). Our geochemical module implements a Newton-Raphson algorithm to solve heterogeneous geochemical equilibria, involving up to 40 chemical components and 400 aqueous species. The geochemical module was designed to be efficient and compact. A revised version of the MINTEQ Code is used as a parent geochemical code

  6. Intracoin - International Nuclide Transport Code Intercomparison Study

    International Nuclear Information System (INIS)

    1984-09-01

    The purpose of the project is to obtain improved knowledge of the influence of various strategies for radionuclide transport modelling for the safety assessment of final repositories for nuclear waste. This is a report of the first phase of the project which was devoted to a comparison of the numerical accuracy of the computer codes used in the study. The codes can be divided into five groups, namely advection-dispersion models, models including matrix diffusion and chemical effects and finally combined models. The results are presented as comparisons of calculations since the objective of level 1 was code verification. (G.B.)

  7. 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

  8. Deterministic methods in radiation transport

    International Nuclear Information System (INIS)

    Rice, A.F.; Roussin, R.W.

    1992-06-01

    The Seminar on Deterministic Methods in Radiation Transport was held February 4--5, 1992, in Oak Ridge, Tennessee. Eleven presentations were made and the full papers are published in this report, along with three that were submitted but not given orally. These papers represent a good overview of the state of the art in the deterministic solution of radiation transport problems for a variety of applications of current interest to the Radiation Shielding Information Center user community

  9. 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.

  10. Development of particle and heavy ion transport code system

    International Nuclear Information System (INIS)

    Niita, Koji

    2004-01-01

    Particle and heavy ion transport code system (PHITS) is 3 dimension general purpose Monte Carlo simulation codes for description of transport and reaction of particle and heavy ion in materials. It is developed on the basis of NMTC/JAM for design and safety of J-PARC. What is PHITS, it's physical process, physical models and development process of PHITC code are described. For examples of application, evaluation of neutron optics, cancer treatment by heavy particle ray and cosmic radiation are stated. JAM and JQMD model are used as the physical model. Neutron motion in six polar magnetic field and gravitational field, PHITC simulation of trace of C 12 beam and secondary neutron track of small model of cancer treatment device in HIMAC and neutron flux in Space Shuttle are explained. (S.Y.)

  11. 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)

  12. Transport code and nuclear data in intermediate energy region

    International Nuclear Information System (INIS)

    Hasegawa, Akira; Odama, Naomitsu; Maekawa, F.; Ueki, K.; Kosaka, K.; Oyama, Y.

    1998-01-01

    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)

  13. 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

  14. Jetto a free boundary plasma transport code

    International Nuclear Information System (INIS)

    Cenacchi, G.; Taroni, A.

    1988-01-01

    JETTO is a one-and-a-half-dimensional transport code calculating the evolution of plasma parameters in a time dependent axisymmetric MHD equilibrium configuration. A splitting technique gives a consistent solution of coupled equilibrium and transport equations. The plasma boundary is free and defined either by its contact with a limiter (wall) or by a separatrix or by the toroidal magnetic flux. The Grad's approach to the equilibrium problem with adiabatic (or similar) constraints is adopted. This method consists of iterating by alternately solving the Grad-Schluter-Shafranov equation (PDE) and the ODE obtained by averaging the PDE over the magnetic surfaces. The bidimensional equation of the poloidal flux is solved by a finite difference scheme, whereas a Runge-Kutta method is chosen for the averaged equilibrium equation. The 1D transport equations (averaged over the magnetic surfaces) for the electron and ion densities and energies and for the rotational transform are written in terms of a coordinate (ρ) related to the toroidal flux. Impurity transport is also considered, under the hypothesis of coronal equilibrium. The transport equations are solved by an implicit scheme in time and by a finite difference scheme in space. The centering of the source terms and transport coefficients is performed using a Predictor-Corrector scheme. The basic version of the code is described here in detail; input and output parameters are also listed

  15. Transport of radioactivity and radiation

    International Nuclear Information System (INIS)

    De Beer, G.P.

    1988-01-01

    The movement of radioactivity and radiation is of prime importance in a wide variety of fields and the present advanced degree of knowledge of transport mechanisms is due largely to the application of sophisticated computer techniques

  16. A computer code package for electron transport Monte Carlo simulation

    International Nuclear Information System (INIS)

    Popescu, Lucretiu M.

    1999-01-01

    A computer code package was developed for solving various electron transport problems by Monte Carlo simulation. It is based on condensed history Monte Carlo algorithm. In order to get reliable results over wide ranges of electron energies and target atomic numbers, specific techniques of electron transport were implemented such as: Moliere multiscatter angular distributions, Blunck-Leisegang multiscatter energy distribution, sampling of electron-electron and Bremsstrahlung individual interactions. Path-length and lateral displacement corrections algorithms and the module for computing collision, radiative and total restricted stopping powers and ranges of electrons are also included. Comparisons of simulation results with experimental measurements are finally presented. (author)

  17. Permanent phonetic identification code for radiation workers

    International Nuclear Information System (INIS)

    Khatua, R.; Somasundaram, S.; Srivastava, D.N.

    1987-01-01

    This report describes a system of self-checking short and easily memorisable 4-digit 'Permanent Phonetic Radiation Code' (PPRC) using radix 128 for Indians occupationally exposed to radiation, to facilitate entry of all radiation dose data pertaining to an individual in a single record of a file. The logic of PPRC is computer compatible. The necessary computer program has been developed in Health Physics Division for printing the PPRCs in Devanagari script through dot-matrix printers for making it understandable to the majority of the persons concerned. (author)

  18. Prototype demonstration of radiation therapy planning code system

    International Nuclear Information System (INIS)

    Little, R.C.; Adams, K.J.; Estes, G.P.; Hughes, L.S. III; Waters, L.S.

    1996-01-01

    This is the final report of a one-year, Laboratory-Directed Research and Development project at the Los Alamos National Laboratory (LANL). Radiation therapy planning is the process by which a radiation oncologist plans a treatment protocol for a patient preparing to undergo radiation therapy. The objective is to develop a protocol that delivers sufficient radiation dose to the entire tumor volume, while minimizing dose to healthy tissue. Radiation therapy planning, as currently practiced in the field, suffers from inaccuracies made in modeling patient anatomy and radiation transport. This project investigated the ability to automatically model patient-specific, three-dimensional (3-D) geometries in advanced Los Alamos radiation transport codes (such as MCNP), and to efficiently generate accurate radiation dose profiles in these geometries via sophisticated physics modeling. Modem scientific visualization techniques were utilized. The long-term goal is that such a system could be used by a non-expert in a distributed computing environment to help plan the treatment protocol for any candidate radiation source. The improved accuracy offered by such a system promises increased efficacy and reduced costs for this important aspect of health care

  19. 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

  20. 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

  1. Radiation transport calculation methods in BNCT

    International Nuclear Information System (INIS)

    Koivunoro, H.; Seppaelae, T.; Savolainen, S.

    2000-01-01

    Boron neutron capture therapy (BNCT) is used as a radiotherapy for malignant brain tumours. Radiation dose distribution is necessary to determine individually for each patient. Radiation transport and dose distribution calculations in BNCT are more complicated than in conventional radiotherapy. Total dose in BNCT consists of several different dose components. The most important dose component for tumour control is therapeutic boron dose D B . The other dose components are gamma dose D g , incident fast neutron dose D f ast n and nitrogen dose D N . Total dose is a weighted sum of the dose components. Calculation of neutron and photon flux is a complex problem and requires numerical methods, i.e. deterministic or stochastic simulation methods. Deterministic methods are based on the numerical solution of Boltzmann transport equation. Such are discrete ordinates (SN) and spherical harmonics (PN) methods. The stochastic simulation method for calculation of radiation transport is known as Monte Carlo method. In the deterministic methods the spatial geometry is partitioned into mesh elements. In SN method angular integrals of the transport equation are replaced with weighted sums over a set of discrete angular directions. Flux is calculated iteratively for all these mesh elements and for each discrete direction. Discrete ordinates transport codes used in the dosimetric calculations are ANISN, DORT and TORT. In PN method a Legendre expansion for angular flux is used instead of discrete direction fluxes, land the angular dependency comes a property of vector function space itself. Thus, only spatial iterations are required for resulting equations. A novel radiation transport code based on PN method and tree-multigrid technique (TMG) has been developed at VTT (Technical Research Centre of Finland). Monte Carlo method solves the radiation transport by randomly selecting neutrons and photons from a prespecified boundary source and following the histories of selected particles

  2. Radiation transport in numerical astrophysics

    International Nuclear Information System (INIS)

    Lund, C.M.

    1983-02-01

    In this article, we discuss some of the numerical techniques developed by Jim Wilson and co-workers for the calculation of time-dependent radiation flow. Difference equations for multifrequency transport are given for both a discrete-angle representation of radiation transport and a Fick's law-like representation. These methods have the important property that they correctly describe both the streaming and diffusion limits of transport theory in problems where the mean free path divided by characteristic distances varies from much less than one to much greater than one. They are also stable for timesteps comparable to the changes in physical variables, rather than being limited by stability requirements

  3. Code of practice for safety in laboratory - non ionising radiation

    International Nuclear Information System (INIS)

    Ramli Jaya; Mohd Yusof Mohd Ali; Khoo Boo Huat; Khatijah Hashim

    1995-01-01

    The code identifies the non-ionizing radiation encountered in laboratories and the associated hazards. The code is intended as a laboratory standard reference document for general information on safety requirements relating to the usage of non-ionizing radiations in laboratories. The nonionizing radiations cover in this code, namely, are ultraviolet radiation, visible light, radio-frequency radiation, lasers, sound waves and ultrasonic radiation. (author)

  4. Photonuclear Physics in Radiation Transport - II: Implementation

    International Nuclear Information System (INIS)

    White, M.C.; Little, R.C.; Chadwick, M.B.; Young, P.G.; MacFarlane, R.E.

    2003-01-01

    This is the second of two companion papers. The first paper describes model calculations and nuclear data evaluations of photonuclear reactions on isotopes of C, O, Al, Si, Ca, Fe, Cu, Ta, W, and Pb for incident photon energies up to 150 MeV. This paper describes the steps taken to process these files into transport libraries and to update the Monte Carlo N-Particle (MCNP) and MCNPX radiation transport codes to use tabular photonuclear reaction data. The evaluated photonuclear data files are created in the standard evaluated nuclear data file (ENDF) format. These files must be processed by the NJOY data processing system into A Compact ENDF (ACE) files suitable for radiation transport calculations. MCNP and MCNPX have been modified to use these new data in a self-consistent and fully integrated manner. Verification problems were used at each step along the path to check the integrity of the methodology. The resulting methodology and tools provide a comprehensive system for using photonuclear data in radiation transport calculations. Also described are initial validation simulations used to benchmark several of the photonuclear transport tables

  5. 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

  6. Inner Radiation Belt Representation of the Energetic Electron Environment: Model and Data Synthesis Using the Salammbo Radiation Belt Transport Code and Los Alamos Geosynchronous and GPS Energetic Particle Data

    Science.gov (United States)

    Friedel, R. H. W.; Bourdarie, S.; Fennell, J.; Kanekal, S.; Cayton, T. E.

    2004-01-01

    The highly energetic electron environment in the inner magnetosphere (GEO inward) has received a lot of research attention in resent years, as the dynamics of relativistic electron acceleration and transport are not yet fully understood. These electrons can cause deep dielectric charging in any space hardware in the MEO to GEO region. We use a new and novel approach to obtain a global representation of the inner magnetospheric energetic electron environment, which can reproduce the absolute environment (flux) for any spacecraft orbit in that region to within a factor of 2 for the energy range of 100 KeV to 5 MeV electrons, for any levels of magnetospheric activity. We combine the extensive set of inner magnetospheric energetic electron observations available at Los Alamos with the physics based Salammbo transport code, using the data assimilation technique of "nudging". This in effect input in-situ data into the code and allows the diffusion mechanisms in the code to interpolate the data into regions and times of no data availability. We present here details of the methods used, both in the data assimilation process and in the necessary inter-calibration of the input data used. We will present sample runs of the model/data code and compare the results to test spacecraft data not used in the data assimilation process.

  7. Path Toward a Unified Geometry for Radiation Transport

    Science.gov (United States)

    Lee, Kerry

    The Direct Accelerated Geometry for Radiation Analysis and Design (DAGRAD) element of the RadWorks Project under Advanced Exploration Systems (AES) within the Space Technology Mission Directorate (STMD) of NASA will enable new designs and concepts of operation for radiation risk assessment, mitigation and protection. This element is designed to produce a solution that will allow NASA to calculate the transport of space radiation through complex CAD models using the state-of-the-art analytic and Monte Carlo radiation transport codes. Due to the inherent hazard of astronaut and spacecraft exposure to ionizing radiation in low-Earth orbit (LEO) or in deep space, risk analyses must be performed for all crew vehicles and habitats. Incorporating these analyses into the design process can minimize the mass needed solely for radiation protection. Transport of the radiation fields as they pass through shielding and body materials can be simulated using Monte Carlo techniques or described by the Boltzmann equation, which is obtained by balancing changes in particle fluxes as they traverse a small volume of material with the gains and losses caused by atomic and nuclear collisions. Deterministic codes that solve the Boltzmann transport equation, such as HZETRN (high charge and energy transport code developed by NASA LaRC), are generally computationally faster than Monte Carlo codes such as FLUKA, GEANT4, MCNP(X) or PHITS; however, they are currently limited to transport in one dimension, which poorly represents the secondary light ion and neutron radiation fields. NASA currently uses HZETRN space radiation transport software, both because it is computationally efficient and because proven methods have been developed for using this software to analyze complex geometries. Although Monte Carlo codes describe the relevant physics in a fully three-dimensional manner, their computational costs have thus far prevented their widespread use for analysis of complex CAD models, leading

  8. High-Fidelity Kinetics and Radiation Transport for NLTE Hypersonic Flows, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — The modeling of NLTE hypersonic flows combines several disciplines: chemistry, kinetics, radiation transport, fluid mechanics, and surface science. No single code or...

  9. 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

  10. 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.)

  11. ipole: Semianalytic scheme for relativistic polarized radiative transport

    Science.gov (United States)

    Moscibrodzka, Monika; Gammie, Charles F.

    2018-04-01

    ipole is a ray-tracing code for covariant, polarized radiative transport particularly useful for modeling Event Horizon Telescope sources, though may also be used for other relativistic transport problems. The code extends the ibothros scheme for covariant, unpolarized transport using two representations of the polarized radiation field: in the coordinate frame, it parallel transports the coherency tensor, and in the frame of the plasma, it evolves the Stokes parameters under emission, absorption, and Faraday conversion. The transport step is as spacetime- and coordinate- independent as possible; the emission, absorption, and Faraday conversion step is implemented using an analytic solution to the polarized transport equation with constant coefficients. As a result, ipole is stable, efficient, and produces a physically reasonable solution even for a step with high optical depth and Faraday depth.

  12. Multiple component codes based generalized LDPC codes for high-speed optical transport.

    Science.gov (United States)

    Djordjevic, Ivan B; Wang, Ting

    2014-07-14

    A class of generalized low-density parity-check (GLDPC) codes suitable for optical communications is proposed, which consists of multiple local codes. It is shown that Hamming, BCH, and Reed-Muller codes can be used as local codes, and that the maximum a posteriori probability (MAP) decoding of these local codes by Ashikhmin-Lytsin algorithm is feasible in terms of complexity and performance. We demonstrate that record coding gains can be obtained from properly designed GLDPC codes, derived from multiple component codes. We then show that several recently proposed classes of LDPC codes such as convolutional and spatially-coupled codes can be described using the concept of GLDPC coding, which indicates that the GLDPC coding can be used as a unified platform for advanced FEC enabling ultra-high speed optical transport. The proposed class of GLDPC codes is also suitable for code-rate adaption, to adjust the error correction strength depending on the optical channel conditions.

  13. Radiation Shielding Information Center: a source of computer codes and data for fusion neutronics studies

    International Nuclear Information System (INIS)

    McGill, B.L.; Roussin, R.W.; Trubey, D.K.; Maskewitz, B.F.

    1980-01-01

    The Radiation Shielding Information Center (RSIC), established in 1962 to collect, package, analyze, and disseminate information, computer codes, and data in the area of radiation transport related to fission, is now being utilized to support fusion neutronics technology. The major activities include: (1) answering technical inquiries on radiation transport problems, (2) collecting, packaging, testing, and disseminating computing technology and data libraries, and (3) reviewing literature and operating a computer-based information retrieval system containing material pertinent to radiation transport analysis. The computer codes emphasize methods for solving the Boltzmann equation such as the discrete ordinates and Monte Carlo techniques, both of which are widely used in fusion neutronics. The data packages include multigroup coupled neutron-gamma-ray cross sections and kerma coefficients, other nuclear data, and radiation transport benchmark problem results

  14. Deuterons at energies of 10 MeV to 1 TeV: Conversion coefficients for fluence-to-absorbed dose, equivalent dose, effective dose and gray equivalent, calculated using Monte Carlo radiation transport code MCNPX 2.7.C

    International Nuclear Information System (INIS)

    Copeland, K.; Parker, D. E.; Friedberg, W.

    2011-01-01

    Conversion coefficients were calculated for fluence-to-absorbed dose, fluence-to-equivalent dose, fluence-to-effective dose and fluence-to-gray equivalent for isotropic exposure of an adult female and an adult male to deuterons ( 2 H + ) in the energy range 10 MeV -1 TeV (0.01-1000 GeV). Coefficients were calculated using the Monte Carlo transport code MCNPX 2.7.C and BodyBuilder TM 1.3 anthropomorphic phantoms. Phantoms were modified to allow calculation of the effective dose to a Reference Person using tissues and tissue weighting factors from 1990 and 2007 recommendations of the International Commission on Radiological Protection (ICRP) and gray equivalent to selected tissues as recommended by the National Council on Radiation Protection and Measurements. Coefficients for the equivalent and effective dose incorporated a radiation weighting factor of 2. At 15 of 19 energies for which coefficients for the effective dose were calculated, coefficients based on ICRP 1990 and 2007 recommendations differed by < 3 %. The greatest difference, 47 %, occurred at 30 MeV. (authors)

  15. 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.)

  16. 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

  17. Study and application of Dot 3.5 computer code in radiation shielding problems

    International Nuclear Information System (INIS)

    Otto, A.C.; Mendonca, A.G.; Maiorino, J.R.

    1983-01-01

    The application of nuclear transportation code S sub(N), Dot 3.5, to radiation shielding problems is revised. Aiming to study the better available option (convergence scheme, calculation mode), of DOT 3.5 computer code to be applied in radiation shielding problems, a standard model from 'Argonne Code Center' was selected and a combination of several calculation options to evaluate the accuracy of the results and the computational time was used, for then to select the more efficient option. To illustrate the versatility and efficacy in the application of the code for tipical shielding problems, the streaming neutrons calculation along a sodium coolant channel is ilustrated. (E.G.) [pt

  18. Verification of Monte Carlo transport codes by activation experiments

    Energy Technology Data Exchange (ETDEWEB)

    Chetvertkova, Vera

    2012-12-18

    With the increasing energies and intensities of heavy-ion accelerator facilities, the problem of an excessive activation of the accelerator components caused by beam losses becomes more and more important. Numerical experiments using Monte Carlo transport codes are performed in order to assess the levels of activation. The heavy-ion versions of the codes were released approximately a decade ago, therefore the verification is needed to be sure that they give reasonable results. Present work is focused on obtaining the experimental data on activation of the targets by heavy-ion beams. Several experiments were performed at GSI Helmholtzzentrum fuer Schwerionenforschung. The interaction of nitrogen, argon and uranium beams with aluminum targets, as well as interaction of nitrogen and argon beams with copper targets was studied. After the irradiation of the targets by different ion beams from the SIS18 synchrotron at GSI, the γ-spectroscopy analysis was done: the γ-spectra of the residual activity were measured, the radioactive nuclides were identified, their amount and depth distribution were detected. The obtained experimental results were compared with the results of the Monte Carlo simulations using FLUKA, MARS and SHIELD. The discrepancies and agreements between experiment and simulations are pointed out. The origin of discrepancies is discussed. Obtained results allow for a better verification of the Monte Carlo transport codes, and also provide information for their further development. The necessity of the activation studies for accelerator applications is discussed. The limits of applicability of the heavy-ion beam-loss criteria were studied using the FLUKA code. FLUKA-simulations were done to determine the most preferable from the radiation protection point of view materials for use in accelerator components.

  19. Transport and attenuation of radiations

    CERN Document Server

    Nimal, J C

    2003-01-01

    This article treats of the calculation methods used for the dimensioning of the protections against radiations. The method consists in determining for a given point the flux of particles coming from a source at a given time. A strong attenuation (of about some few mu Sv.h sup - sup 1) is in general expected between the source and the areas accessible to the personnel or the public. The calculation has to take into account a huge number of radiation-matter interactions and to solve the integral-differential transport equation which links the particles flux to the source. Several methods exist from the simplified physical model with numerical developments to the more or less precise resolution of the transport equation. These methods allows also the calculation of the uncertainties of equivalent dose rates, heat sources, structure damages using the data covariances (efficient cross-sections, modeling, etc..): 1 - transport equation; 2 - Monte-Carlo method; 3 - semi-numerical methods S sub N; 4 - methods based o...

  20. 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

  1. 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

  2. BLINDAGE: A neutron and gamma-ray transport code for shieldings with the removal-diffusion technique coupled with the point-kernel technique

    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

  3. IPOLE - semi-analytic scheme for relativistic polarized radiative transport

    Science.gov (United States)

    Mościbrodzka, M.; Gammie, C. F.

    2018-03-01

    We describe IPOLE, a new public ray-tracing code for covariant, polarized radiative transport. The code extends the IBOTHROS scheme for covariant, unpolarized transport using two representations of the polarized radiation field: In the coordinate frame, it parallel transports the coherency tensor; in the frame of the plasma it evolves the Stokes parameters under emission, absorption, and Faraday conversion. The transport step is implemented to be as spacetime- and coordinate- independent as possible. The emission, absorption, and Faraday conversion step is implemented using an analytic solution to the polarized transport equation with constant coefficients. As a result, IPOLE is stable, efficient, and produces a physically reasonable solution even for a step with high optical depth and Faraday depth. We show that the code matches analytic results in flat space, and that it produces results that converge to those produced by Dexter's GRTRANS polarized transport code on a complicated model problem. We expect IPOLE will mainly find applications in modelling Event Horizon Telescope sources, but it may also be useful in other relativistic transport problems such as modelling for the IXPE mission.

  4. Treating voxel geometries in radiation protection dosimetry with a patched version of the Monte Carlo codes MCNP and MCNPX.

    Science.gov (United States)

    Burn, K W; Daffara, C; Gualdrini, G; Pierantoni, M; Ferrari, P

    2007-01-01

    The question of Monte Carlo simulation of radiation transport in voxel geometries is addressed. Patched versions of the MCNP and MCNPX codes are developed aimed at transporting radiation both in the standard geometry mode and in the voxel geometry treatment. The patched code reads an unformatted FORTRAN file derived from DICOM format data and uses special subroutines to handle voxel-to-voxel radiation transport. The various phases of the development of the methodology are discussed together with the new input options. Examples are given of employment of the code in internal and external dosimetry and comparisons with results from other groups are reported.

  5. 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

  6. Tritons at energies of 10 MeV to 1 TeV: Conversion coefficients for fluence-to-absorbed dose, equivalent dose, effective dose, and gray equivalent, calculated using Monte Carlo radiation transport code MCNPX 2.7.C

    International Nuclear Information System (INIS)

    Copeland, K.; Parker, D. E.; Friedberg, W.

    2010-01-01

    Conversion coefficients were calculated for fluence-to-absorbed dose, fluence-to-equivalent dose, fluence-to-effective dose and fluence-to-gray equivalent for isotropic exposure of an adult female and an adult male to tritons ( 3 H + ) in the energy range of 10 MeV to 1 TeV (0.01-1000 GeV). Coefficients were calculated using Monte Carlo transport code MCNPX 2.7.C and BodyBuilder TM 1.3 anthropomorphic phantoms. Phantoms were modified to allow calculation of effective dose to a Reference Person using tissues and tissue weighting factors from 1990 and 2007 recommendations of the International Commission on Radiological Protection (ICRP) and calculation of gray equivalent to selected tissues as recommended by the National Council on Radiation Protection and Measurements. At 15 of the 19 energies for which coefficients for effective dose were calculated, coefficients based on ICRP 2007 and 1990 recommendations differed by less than 3%. The greatest difference, 43%, occurred at 30 MeV. Published by Oxford Univ. Press on behalf of the US Government 2010. (authors)

  7. Helions at energies of 10 MeV to 1 TeV: Conversion coefficients for fluence-to-absorbed dose, equivalent dose, effective dose and gray equivalent, calculated using Monte Carlo radiation transport code MCNPX 2.7.C

    International Nuclear Information System (INIS)

    Copeland, K.; Parker, D. E.; Friedberg, W.

    2010-01-01

    Conversion coefficients were calculated for fluence-to-absorbed dose, fluence-to-equivalent dose, fluence-to-effective dose and fluence-to-gray equivalent, for isotropic exposure of an adult male and an adult female to helions ( 3 He 2+ ) in the energy range of 10 MeV to 1 TeV (0.01-1000 GeV). Calculations were performed using Monte Carlo transport code MCNPX 2.7.C and BodyBuilder TM 1.3 anthropomorphic phantoms modified to allow calculation of effective dose using tissues and tissue weighting factors from either the 1990 or 2007 recommendations of the International Commission on Radiological Protection (ICRP), and gray equivalent to selected tissues as recommended by the National Council on Radiation Protection and Measurements. At 15 of the 19 energies for which coefficients for effective dose were calculated, coefficients based on ICRP 2007 and 1990 recommendations differed by less than 2%. The greatest difference, 62%, occurred at 100 MeV. Published by Oxford Univ. Press on behalf of the U.S. Government 2010. (authors)

  8. 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

  9. Simulation of Hamming Coding and Decoding for Microcontroller Radiation Hardening

    OpenAIRE

    Rehab I. Abdul Rahman; Mazhar B. Tayel

    2015-01-01

    This paper presents a method of hardening the 8051 micro-controller, able to assure reliable operation in the presence of bit flips caused by radiation. Aiming at avoiding such faults in the 8051 micro-controller, Hamming code protection was used in its SRAM memory and registers. A VHDL code has been used for this hamming code protection.

  10. Radiation transport Part B: Applications with examples

    International Nuclear Information System (INIS)

    Beutler, D.E.

    1997-01-01

    In the previous sections Len Lorence has described the need, theory, and types of radiation codes that can be applied to model the results of radiation effects tests or working environments for electronics. For the rest of this segment, the author will concentrate on the specific ways the codes can be used to predict device response or analyze radiation test results. Regardless of whether one is predicting responses in a working or test environment, the procedures are virtually the same. The same can be said for the use of 1-, 2-, or 3-dimensional codes and Monte Carlo or discrete ordinates codes. No attempt is made to instruct the student on the specifics of the code. For example, the author will not discuss the details, such as the number of meshes, energy groups, etc. that are appropriate for a discrete ordinates code. For the sake of simplicity, he will restrict himself to the 1-dimensional code CEPXS/ONELD. This code along with a wide variety of other radiation codes can be obtained form the Radiation Safety Information Computational Center (RSICC) for a nominal handling fee

  11. PEREGRINE: An all-particle Monte Carlo code for radiation therapy

    International Nuclear Information System (INIS)

    Hartmann Siantar, C.L.; Chandler, W.P.; Rathkopf, J.A.; Svatos, M.M.; White, R.M.

    1994-09-01

    The goal of radiation therapy is to deliver a lethal dose to the tumor while minimizing the dose to normal tissues. To carry out this task, it is critical to calculate correctly the distribution of dose delivered. Monte Carlo transport methods have the potential to provide more accurate prediction of dose distributions than currently-used methods. PEREGRINE is a new Monte Carlo transport code developed at Lawrence Livermore National Laboratory for the specific purpose of modeling the effects of radiation therapy. PEREGRINE transports neutrons, photons, electrons, positrons, and heavy charged-particles, including protons, deuterons, tritons, helium-3, and alpha particles. This paper describes the PEREGRINE transport code and some preliminary results for clinically relevant materials and radiation sources

  12. Directions in Radiation Transport Modelling

    Directory of Open Access Journals (Sweden)

    P Nicholas Smith

    2016-12-01

    More exciting advances are on the horizon to increase the power of simulation tools. The advent of high performance computers is allowing bigger, higher fidelity models to be created, if the challenges of parallelization and memory management can be met. 3D whole core transport modelling is becoming possible. Uncertainty quantification is improving with large benefits to be gained from more accurate, less pessimistic estimates of uncertainty. Advanced graphical displays allow the user to assimilate and make sense of the vast amounts of data produced by modern modelling tools. Numerical solvers are being developed that use goal-based adaptivity to adjust the nodalisation of the system to provide the optimum scheme to achieve the user requested accuracy on the results, thus removing the need to perform costly convergence studies in space and angle etc. More use is being made of multi-physics methods in which radiation transport is coupled with other phenomena, such as thermal-hydraulics, structural response, fuel performance and/or chemistry in order to better understand their interplay in reactor cores.

  13. Adaptation of radiation shielding code to space environment

    International Nuclear Information System (INIS)

    Okuno, Koichi; Hara, Akihisa

    1992-01-01

    Recently, the trend to the development of space has heightened. To the development of space, many problems are related, and as one of them, there is the protection from cosmic ray. The cosmic ray is the radiation having ultrahigh energy, and there was not the radiation shielding design code that copes with cosmic ray so far. Therefore, the high energy radiation shielding design code for accelerators was improved so as to cope with the peculiarity that cosmic ray possesses. Moreover, the calculation of the radiation dose equivalent rate in the moon base to which the countermeasures against cosmic ray were taken was simulated by using the improved code. As the important countermeasures for the safety protection from radiation, the covering with regolith is carried out, and the effect of regolith was confirmed by using the improved code. Galactic cosmic ray, solar flare particles, radiation belt, the adaptation of the radiation shielding code HERMES to space environment, the improvement of the three-dimensional hadron cascade code HETCKFA-2 and the electromagnetic cascade code EGS 4-KFA, and the cosmic ray simulation are reported. (K.I.)

  14. A New Monte Carlo Neutron Transport Code at UNIST

    International Nuclear Information System (INIS)

    Lee, Hyunsuk; Kong, Chidong; Lee, Deokjung

    2014-01-01

    Monte Carlo neutron transport code named MCS is under development at UNIST for the advanced reactor design and research purpose. This MC code can be used for fixed source calculation and criticality calculation. Continuous energy neutron cross section data and multi-group cross section data can be used for the MC calculation. This paper presents the overview of developed MC code and its calculation results. The real time fixed source calculation ability is also tested in this paper. The calculation results show good agreement with commercial code and experiment. A new Monte Carlo neutron transport code is being developed at UNIST. The MC codes are tested with several benchmark problems: ICSBEP, VENUS-2, and Hoogenboom-Martin benchmark. These benchmarks covers pin geometry to 3-dimensional whole core, and results shows good agreement with reference results

  15. Optix: A Monte Carlo scintillation light transport code

    Energy Technology Data Exchange (ETDEWEB)

    Safari, M.J., E-mail: mjsafari@aut.ac.ir [Department of Energy Engineering and Physics, Amir Kabir University of Technology, PO Box 15875-4413, Tehran (Iran, Islamic Republic of); Afarideh, H. [Department of Energy Engineering and Physics, Amir Kabir University of Technology, PO Box 15875-4413, Tehran (Iran, Islamic Republic of); Ghal-Eh, N. [School of Physics, Damghan University, PO Box 36716-41167, Damghan (Iran, Islamic Republic of); Davani, F. Abbasi [Nuclear Engineering Department, Shahid Beheshti University, PO Box 1983963113, Tehran (Iran, Islamic Republic of)

    2014-02-11

    The paper reports on the capabilities of Monte Carlo scintillation light transport code Optix, which is an extended version of previously introduced code Optics. Optix provides the user a variety of both numerical and graphical outputs with a very simple and user-friendly input structure. A benchmarking strategy has been adopted based on the comparison with experimental results, semi-analytical solutions, and other Monte Carlo simulation codes to verify various aspects of the developed code. Besides, some extensive comparisons have been made against the tracking abilities of general-purpose MCNPX and FLUKA codes. The presented benchmark results for the Optix code exhibit promising agreements. -- Highlights: • Monte Carlo simulation of scintillation light transport in 3D geometry. • Evaluation of angular distribution of detected photons. • Benchmark studies to check the accuracy of Monte Carlo simulations.

  16. 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

  17. Interfacial and Wall Transport Models for SPACE-CAP Code

    International Nuclear Information System (INIS)

    Hong, Soon Joon; Choo, Yeon Joon; Han, Tae Young; Hwang, Su Hyun; Lee, Byung Chul; Choi, Hoon; Ha, Sang Jun

    2009-01-01

    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

  18. 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.

  19. TRIPOLI-4: Monte Carlo transport code functionalities and applications

    International Nuclear Information System (INIS)

    Both, J.P.; Lee, Y.K.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B.

    2003-01-01

    Tripoli-4 is a three dimensional calculations code using the Monte Carlo method to simulate the transport of neutrons, photons, electrons and positrons. This code is used in four application fields: the protection studies, the criticality studies, the core studies and the instrumentation studies. Geometry, cross sections, description of sources, principle. (N.C.)

  20. Non-standard model for electron heat transport for multidimensional hydrodynamic codes

    Energy Technology Data Exchange (ETDEWEB)

    Nicolai, Ph.; Busquet, M.; Schurtz, G. [CEA/DAM-Ile de France, 91 - Bruyeres Le Chatel (France)

    2000-07-01

    In simulations of laser-produced plasma, modeling of heat transport requires an artificial limitation of standard Spitzer-Haerm fluxes. To improve heat conduction processing, we have developed a multidimensional model which accounts for non-local features of heat transport and effects of self-generated magnetic fields. This consistent treatment of both mechanisms has been implemented in a two-dimensional radiation-hydrodynamic code. First results indicate good agreements between simulations and experimental data. (authors)

  1. Non-standard model for electron heat transport for multidimensional hydrodynamic codes

    International Nuclear Information System (INIS)

    Nicolai, Ph.; Busquet, M.; Schurtz, G.

    2000-01-01

    In simulations of laser-produced plasma, modeling of heat transport requires an artificial limitation of standard Spitzer-Haerm fluxes. To improve heat conduction processing, we have developed a multidimensional model which accounts for non-local features of heat transport and effects of self-generated magnetic fields. This consistent treatment of both mechanisms has been implemented in a two-dimensional radiation-hydrodynamic code. First results indicate good agreements between simulations and experimental data. (authors)

  2. Code of practice for radiation protection in veterinary medicine

    International Nuclear Information System (INIS)

    Duffy, J.; Fenton, D.; McGarry, A.; McAllister, H.; Skelly, C

    2002-11-01

    This Code of Practice updates the Code of Practice on Radiation Protection in Veterinary Radiology prepared by the Nuclear Energy Board in June 1989. The Code is designed to give guidance to veterinary surgeons to ensure that they, their employees and members of the public are adequately protected from the hazards of ionising radiation arising from the use of X-ray equipment and radioactive substances in the practice of veterinary medicine. It reflects the regulations as specified in the Radiological Protection Act, 1991, (Ionising Radiation) Order, 2000 (S.I. No. 125 of 2000)

  3. 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

  4. CHMTRNS, Non-Equilibrium Chemical Transport Code

    International Nuclear Information System (INIS)

    Noorishad, J.; Carnahan, C.L.; Benson, L.V.

    1998-01-01

    1 - Description of program or function: CHMTRNS simulates solute transport for steady one-dimensional fluid flow by convection and diffusion or dispersion in a saturated porous medium based on the assumption of local chemical equilibrium. The chemical interactions included in the model are aqueous-phase complexation, solid-phase ion exchange of bare ions and complexes using the surface complexation model, and precipitation or dissolution of solids. The program can simulate the kinetic dissolution or precipitation for calcite and silica as well as irreversible dissolution of glass. Thermodynamic parameters are temperature dependent and are coupled to a companion heat transport simulator; thus, the effects of transient temperature conditions can be considered. Options for oxidation-reduction (redox) and C-13 fractionation as well as non-isothermal conditions are included. 2 - Method of solution: The governing equations for both reactive chemical and heat transport are discretized in time and space. For heat transport, the Crank-Nicolson approximation is used in conjunction with a LU decomposition and backward substitution solution procedure. To deal with the strong nonlinearity of the chemical transport equations, a generalized Newton-Raphson method is used

  5. Benchmark studies of BOUT++ code and TPSMBI code on neutral transport during SMBI

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Y.H. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); Center for Magnetic Fusion Theory, Chinese Academy of Sciences, Hefei 230031 (China); Wang, Z.H., E-mail: zhwang@swip.ac.cn [Southwestern Institute of Physics, Chengdu 610041 (China); Guo, W., E-mail: wfguo@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Center for Magnetic Fusion Theory, Chinese Academy of Sciences, Hefei 230031 (China); Ren, Q.L. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Sun, A.P.; Xu, M.; Wang, A.K. [Southwestern Institute of Physics, Chengdu 610041 (China); Xiang, N. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Center for Magnetic Fusion Theory, Chinese Academy of Sciences, Hefei 230031 (China)

    2017-06-09

    SMBI (supersonic molecule beam injection) plays an important role in tokamak plasma fuelling, density control and ELM mitigation in magnetic confinement plasma physics, which has been widely used in many tokamaks. The trans-neut module of BOUT++ code is the only large-scale parallel 3D fluid code used to simulate the SMBI fueling process, while the TPSMBI (transport of supersonic molecule beam injection) code is a recent developed 1D fluid code of SMBI. In order to find a method to increase SMBI fueling efficiency in H-mode plasma, especially for ITER, it is significant to first verify the codes. The benchmark study between the trans-neut module of BOUT++ code and the TPSMBI code on radial transport dynamics of neutral during SMBI has been first successfully achieved in both slab and cylindrical coordinates. The simulation results from the trans-neut module of BOUT++ code and TPSMBI code are consistent very well with each other. Different upwind schemes have been compared to deal with the sharp gradient front region during the inward propagation of SMBI for the code stability. The influence of the WENO3 (weighted essentially non-oscillatory) and the third order upwind schemes on the benchmark results has also been discussed. - Highlights: • A 1D model of SMBI has developed. • Benchmarks of BOUT++ and TPSMBI codes have first been finished. • The influence of the WENO3 and the third order upwind schemes on the benchmark results has also been discussed.

  6. TRIPOLI-4: Monte Carlo transport code functionalities and applications; TRIPOLI-4: code de transport Monte Carlo fonctionnalites et applications

    Energy Technology Data Exchange (ETDEWEB)

    Both, J P; Lee, Y K; Mazzolo, A; Peneliau, Y; Petit, O; Roesslinger, B [CEA Saclay, Dir. de l' Energie Nucleaire (DEN), Service d' Etudes de Reacteurs et de Modelisation Avancee, 91 - Gif sur Yvette (France)

    2003-07-01

    Tripoli-4 is a three dimensional calculations code using the Monte Carlo method to simulate the transport of neutrons, photons, electrons and positrons. This code is used in four application fields: the protection studies, the criticality studies, the core studies and the instrumentation studies. Geometry, cross sections, description of sources, principle. (N.C.)

  7. 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

  8. Survey of 1 1/2D transport codes

    International Nuclear Information System (INIS)

    Grad, H.

    1978-10-01

    A survey is given of a family of classical transport codes, recently termed ''1 1/2D'', which efficiently and accurately follow the evolution of plasma configurations on a long time scale, following coupled changes in plasma shape and topology with transport (but not wave motion). Codes have been constructed and operated (since 1974) which include various combinations of finite beta, general plasma cross-section and aspect, various topologies (Doublet, tearing, reversed-field mirror) including time dependent transitions in topology resulting from external coil variation and plasma transport, with models including (classical) tensor resistivity and heat flow as well as the adiabatic limiting case

  9. Applications of the Los Alamos High Energy Transport code

    International Nuclear Information System (INIS)

    Waters, L.; Gavron, A.; Prael, R.E.

    1992-01-01

    Simulation codes reliable through a large range of energies are essential to analyze the environment of vehicles and habitats proposed for space exploration. The LAHET monte carlo code has recently been expanded to track high energy hadrons with FLUKA, while retaining the original Los Alamos version of HETC at lower energies. Electrons and photons are transported with EGS4, and an interface to the MCNP monte carlo code is provided to analyze neutrons with kinetic energies less than 20 MeV. These codes are benchmarked by comparison of LAHET/MCNP calculations to data from the Brookhaven experiment E814 participant calorimeter

  10. Method for calculating internal radiation and ventilation with the ADINAT heat-flow code

    International Nuclear Information System (INIS)

    Butkovich, T.R.; Montan, D.N.

    1980-01-01

    One objective of the spent fuel test in Climax Stock granite (SFTC) is to correctly model the thermal transport, and the changes in the stress field and accompanying displacements from the application of the thermal loads. We have chosen the ADINA and ADINAT finite element codes to do these calculations. ADINAT is a heat transfer code compatible to the ADINA displacement and stress analysis code. The heat flow problem encountered at SFTC requires a code with conduction, radiation, and ventilation capabilities, which the present version of ADINAT does not have. We have devised a method for calculating internal radiation and ventilation with the ADINAT code. This method effectively reproduces the results from the TRUMP multi-dimensional finite difference code, which correctly models radiative heat transport between drift surfaces, conductive and convective thermal transport to and through air in the drifts, and mass flow of air in the drifts. The temperature histories for each node in the finite element mesh calculated with ADINAT using this method can be used directly in the ADINA thermal-mechanical calculation

  11. Code of practice of radiation protection in fixed nuclear gauges

    International Nuclear Information System (INIS)

    Eltayeb, M. A. M.

    2012-09-01

    The present work aims at developing and updating a code of practice of radiation protection in fixed nuclear gauges that comply with current international recommendations. The work also intended to evaluate the current radiation protection situation in two selected companies using nuclear gauges in Sudan. A draft of the code is proposed which includes the basic principle of protection such as source construction and gauges radiation monitoring, storage maintenance and leak testing as well as specific issues related to nuclear gauges. The practical part of this study included investigation of radiation protection in the comparisons using nuclear gauges for level detection, to evaluate the level of radiation protection and the compliance to the regulatory authority regulations. The result revealed that the two companies do not have an effective radiation protection program and that can lead to exposure of workers to unnecessary doses. Some recommendations were stated, if implemented they could improve the status of radiation protection in those companies. (Author)

  12. Study of Radiation Shielding Analysis for Low-Intermediate Level Waste Transport Ship

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dohyung; Lee, Unjang; Song, Yangsoo; Kim, Sukhoon; Ko, Jaehoon [Korea Nuclear Engineering and Service Corporation, Seoul (Korea, Republic of)

    2007-07-01

    In Korea, it is planed to transport Low-Intermediate Level Radioactive Waste (LILW) from each nuclear power plant site to Kyongju LILW repository after 2009. Transport through the sea using ship is one of the most prospective ways of LILW transport for current situation in Korea. There are domestic and international regulations for radiation dose limit for radioactive material transport. In this article, radiation shielding analysis for LILW transport ship is performed using 3-D computer simulation code, MCNP. As a result, the thickness and materials for radiation shielding walls next to cargo in the LILW transport ship are determined.

  13. 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.

  14. Survey of radiation protection programmes for transport

    International Nuclear Information System (INIS)

    Lizot, M.T.; Perrin, M.L.; Sert, G.; Lange, F.; Schwarz, G.; Feet, H.J.; Christ, R.; Shaw, K.B.; Hughes, J.S.; Gelder, R.

    2001-07-01

    The survey of radiation protection programmes for transport has been jointly performed by three scientific organisations I.P.S.N. (France), G.R.S. ( Germany), and N.R.P.B. (United kingdom) on behalf of the European Commission and the pertaining documentation summarises the findings and conclusions of the work that was undertaken with the principal objectives to provide guidance on the establishment, implementation and application of radiation protection programmes for the transport of radioactive materials by operators and the assessment and evaluation of such programmes by the competent authority and to review currently existing radiation protection programmes for the transport of radioactive materials. (N.C.)

  15. Experimental transport analysis code system in JT-60

    International Nuclear Information System (INIS)

    Hirayama, Toshio; Shimizu, Katsuhiro; Tani, Keiji; Shirai, Hiroshi; Kikuchi, Mitsuru

    1988-03-01

    Transport analysis codes have been developed in order to study confinement properties related to particle and energy balance in ohmically and neutral beam heated plasmas of JT-60. The analysis procedure is divided into three steps as follows: 1) LOOK ; The shape of the plasma boundary is identified with a fast boundary identification code of FBI by using magnetic data, and flux surfaces are calculated with a MHD equilibrium code of SELENE. The diagnostic data are mapped to flux surfaces for neutral beam heating calculation and/or for radial transport analysis. 2) OFMC ; On the basis of transformed data, an orbit following Monte Carlo code of OFMC calculates both profiles of power deposition and particle source of neutral beam injected into a plasma. 3) SCOOP ; In the last stage, a one dimensional transport code of SCOOP solves particle and energy balance for electron and ion, in order to evaluate transport coefficients as well as global parameters such as energy confinement time and the stored energy. The analysis results are provided to a data bank of DARTS that is used to find an overview of important consideration on confinement with a regression analysis code of RAC. (author)

  16. DIAPHANE: A portable radiation transport library for astrophysical applications

    Science.gov (United States)

    Reed, Darren S.; Dykes, Tim; Cabezón, Rubén; Gheller, Claudio; Mayer, Lucio

    2018-05-01

    One of the most computationally demanding aspects of the hydrodynamical modelingof Astrophysical phenomena is the transport of energy by radiation or relativistic particles. Physical processes involving energy transport are ubiquitous and of capital importance in many scenarios ranging from planet formation to cosmic structure evolution, including explosive events like core collapse supernova or gamma-ray bursts. Moreover, the ability to model and hence understand these processes has often been limited by the approximations and incompleteness in the treatment of radiation and relativistic particles. The DIAPHANE project has focused on developing a portable and scalable library that handles the transport of radiation and particles (in particular neutrinos) independently of the underlying hydrodynamic code. In this work, we present the computational framework and the functionalities of the first version of the DIAPHANE library, which has been successfully ported to three different smoothed-particle hydrodynamic codes, GADGET2, GASOLINE and SPHYNX. We also present validation of different modules solving the equations of radiation and neutrino transport using different numerical schemes.

  17. Implementation of the kinetics in the transport code AZTRAN

    International Nuclear Information System (INIS)

    Duran G, J. A.; Del Valle G, E.; Gomez T, A. M.

    2017-09-01

    This paper shows the implementation of the time dependence in the three-dimensional transport code AZTRAN (AZtlan TRANsport), which belongs to the AZTLAN platform, for the analysis of nuclear reactors (currently under development). The AZTRAN code with this implementation is able to numerically solve the time-dependent transport equation in XYZ geometry, for several energy groups, using the discrete ordinate method S n for the discretization of the angular variable, the nodal method RTN-0 for spatial discretization and method 0 for discretization in time. Initially, the code only solved the neutrons transport equation in steady state, so the implementation of the temporal part was made integrating the neutrons transport equation with respect to time and balance equations corresponding to the concentrations of delayed neutron precursors, for which method 0 was applied. After having directly implemented code kinetics, the improved quasi-static method was implemented, which is a tool for reducing computation time, where the angular flow is factored by the product of two functions called shape function and amplitude function, where the first is calculated for long time steps, called macro-steps and the second is resolved for small time steps called micro-steps. In the new version of AZTRAN several Benchmark problems that were taken from the literature were simulated, the problems used are of two and three dimensions which allowed corroborating the accuracy and stability of the code, showing in general in the reference tests a good behavior. (Author)

  18. Radiation-emitting Electronic Product Codes

    Data.gov (United States)

    U.S. Department of Health & Human Services — This database contains product names and associated information developed by the Center for all products, both medical and non-medical, which emit radiation. It...

  19. Los Alamos neutral particle transport codes: New and enhanced capabilities

    International Nuclear Information System (INIS)

    Alcouffe, R.E.; Baker, R.S.; Brinkley, F.W.; Clark, B.A.; Koch, K.R.; Marr, D.R.

    1992-01-01

    We present new developments in Los Alamos discrete-ordinates transport codes and introduce THREEDANT, the latest in the series of Los Alamos discrete ordinates transport codes. THREEDANT solves the multigroup, neutral-particle transport equation in X-Y-Z and R-Θ-Z geometries. THREEDANT uses computationally efficient algorithms: Diffusion Synthetic Acceleration (DSA) is used to accelerate the convergence of transport iterations, the DSA solution is accelerated using the multigrid technique. THREEDANT runs on a wide range of computers, from scientific workstations to CRAY supercomputers. The algorithms are highly vectorized on CRAY computers. Recently, the THREEDANT transport algorithm was implemented on the massively parallel CM-2 computer, with performance that is comparable to a single-processor CRAY-YMP We present the results of THREEDANT analysis of test problems

  20. Development of a CAD-based neutron transport code with the method of characteristics

    International Nuclear Information System (INIS)

    Chen Zhenping; Wang Dianxi; He Tao; Wang Guozhong; Zheng Huaqing

    2012-01-01

    The main problem determining whether the method of characteristics (MOC) can be used in complicated and highly heterogeneous geometry is how to combine an effective geometry processing method with MOC. In this study, a new idea making use of MCAM, which is a Mutlti-Calculation Automatic Modeling for Neutronics and Radiation Transport program developed by FDS Team, for geometry description and ray tracing of particle transport was brought forward to solve the geometry problem mentioned above. Based on the theory and approach as the foregoing statement, a two dimensional neutron transport code was developed which had been integrated into VisualBUS, developed by FDS Team. Several benchmarks were used to verify the validity of the code and the numerical results were coincident with the reference values very well, which indicated the accuracy and feasibility of the method and the MOC code. (authors)

  1. Code of practice against radiation hazards at PINSTECH

    International Nuclear Information System (INIS)

    Mubarak, M.A.; Javed, M.; Ahmad, S.

    1982-10-01

    It is the radiation safety policy of PAEC/PINSTECH that all radiation exposure should be kept as low as reasonably achievable (ALARA). A code of practice against radiation hazards at PINSTECH was written in 1972 which regulated the conduct of radiation work at PINSTECH. Since the radiation work at PINSTECH has greatly increased, it was considered necessary to revise the code so as to incorporate the new concepts in this field as well as to help meet the present requirements of radiation protection. The procedures set forth in this code are mandatory and in no case should any of them be deviated except under an emergency situation which may be handled according to procedures laid down in a separate manual ''Emergency Procedures at PARR-PINSTECH'' (PINSTECH/HP--19). All those supervising or performing any kind of radiation work are required to study and adhere to these procedures. Copy of this code should be kept in every radiation laboratory for ready reference. (author)

  2. Some factors affecting radiative heat transport in PWR cores

    International Nuclear Information System (INIS)

    Hall, A.N.

    1989-04-01

    This report discusses radiative heat transport in Pressurized Water Reactor cores, using simple models to illustrate basic features of the transport process. Heat transport by conduction and convection is ignored in order to focus attention on the restrictions on radiative heat transport imposed by the geometry of the heat emitting and absorbing structures. The importance of the spacing of the emitting and absorbing structures is emphasised. Steady state temperature distributions are found for models of cores which are uniformly heated by fission product decay. In all of the models, a steady state temperature distribution can only be obtained if the central core temperature is in excess of the melting point of UO 2 . It has recently been reported that the MIMAS computer code, which takes into account radiative heat transport, has been used to model the heat-up of the Three Mile Island-2 reactor core, and the computations indicate that the core could not have reached the melting point of UO 2 at any time or any place. We discuss this result in the light of the calculations presented in this paper. It appears that the predicted stabilisation of the core temperatures at ∼ 2200 0 C may be a consequence of the artificially large spacing between the radial rings employed in the MIMAS code, rather than a result of physical significance. (author)

  3. Particle and heavy ion transport code system; PHITS

    International Nuclear Information System (INIS)

    Niita, Koji

    2004-01-01

    Intermediate and high energy nuclear data are strongly required in design study of many facilities such as accelerator-driven systems, intense pulse spallation neutron sources, and also in medical and space technology. There is, however, few evaluated nuclear data of intermediate and high energy nuclear reactions. Therefore, we have to use some models or systematics for the cross sections, which are essential ingredients of high energy particle and heavy ion transport code to estimate neutron yield, heat deposition and many other quantities of the transport phenomena in materials. We have developed general purpose particle and heavy ion transport Monte Carlo code system, PHITS (Particle and Heavy Ion Transport code System), based on the NMTC/JAM code by the collaboration of Tohoku University, JAERI and RIST. The PHITS has three important ingredients which enable us to calculate (1) high energy nuclear reactions up to 200 GeV, (2) heavy ion collision and its transport in material, (3) low energy neutron transport based on the evaluated nuclear data. In the PHITS, the cross sections of high energy nuclear reactions are obtained by JAM model. JAM (Jet AA Microscopic Transport Model) is a hadronic cascade model, which explicitly treats all established hadronic states including resonances and all hadron-hadron cross sections parametrized based on the resonance model and string model by fitting the available experimental data. The PHITS can describe the transport of heavy ions and their collisions by making use of JQMD and SPAR code. The JQMD (JAERI Quantum Molecular Dynamics) is a simulation code for nucleus nucleus collisions based on the molecular dynamics. The SPAR code is widely used to calculate the stopping powers and ranges for charged particles and heavy ions. The PHITS has included some part of MCNP4C code, by which the transport of low energy neutron, photon and electron based on the evaluated nuclear data can be described. Furthermore, the high energy nuclear

  4. Analysis of radiation field distribution in Yonggwang unit 3 with MCNP code

    International Nuclear Information System (INIS)

    Lee, Cheol Woo; Ha, Wi Ho; Shin, Chang Ho; Kim, Soon Young; Kim, Jong Kyung

    2004-01-01

    Radiation field analysis is performed at the inside of the containment building of nuclear power plant(NPP) using the well-known MCNP code. The target NPP in this study is Yonggwang Unit 3 Cycle 8. In this work, whole transport calculations were done using MCNPX 2.4.0 due to the functional benefits, such as Mesh Tally, that the code provides. The neutron spectra released from the operating reactor core were firstly evaluated as a radiation source term, and then dose distributions in the work areas of the NPP were calculated

  5. Verification of Monte Carlo transport codes by activation experiments

    OpenAIRE

    Chetvertkova, Vera

    2013-01-01

    With the increasing energies and intensities of heavy-ion accelerator facilities, the problem of an excessive activation of the accelerator components caused by beam losses becomes more and more important. Numerical experiments using Monte Carlo transport codes are performed in order to assess the levels of activation. The heavy-ion versions of the codes were released approximately a decade ago, therefore the verification is needed to be sure that they give reasonable results. Present work is...

  6. 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

  7. 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

  8. High energy particle transport code NMTC/JAM

    International Nuclear Information System (INIS)

    Niita, K.; Takada, H.; Meigo, S.; Ikeda, Y.

    2001-01-01

    We have developed a high energy particle transport code NMTC/JAM, which is an upgrade version of NMTC/JAERI97. The available energy range of NMTC/JAM is, in principle, extended to 200 GeV for nucleons and mesons including the high energy nuclear reaction code JAM for the intra-nuclear cascade part. We compare the calculations by NMTC/JAM code with the experimental data of thin and thick targets for proton induced reactions up to several 10 GeV. The results of NMTC/JAM code show excellent agreement with the experimental data. From these code validation, it is concluded that NMTC/JAM is reliable in neutronics optimization study of the high intense spallation neutron utilization facility. (author)

  9. Code system to compute radiation dose in human phantoms

    International Nuclear Information System (INIS)

    Ryman, J.C.; Cristy, M.; Eckerman, K.F.; Davis, J.L.; Tang, J.S.; Kerr, G.D.

    1986-01-01

    Monte Carlo photon transport code and a code using Monte Carlo integration of a point kernel have been revised to incorporate human phantom models for an adult female, juveniles of various ages, and a pregnant female at the end of the first trimester of pregnancy, in addition to the adult male used earlier. An analysis code has been developed for deriving recommended values of specific absorbed fractions of photon energy. The computer code system and calculational method are described, emphasizing recent improvements in methods

  10. 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

  11. 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.

  12. Guide to the declaration procedure and coding system for criteria concerning significant events related to safety, radiation protection or the environment, applicable to basic nuclear installations and the transport of radioactive materials

    International Nuclear Information System (INIS)

    Lacoste, Andre-Claude

    2005-01-01

    This guide notably contains various forms associated with the declaration of significant events, and explanations to fill them in: significant event declaration form for a basic nuclear installation, significant event declaration form for radioactive material transport, significant event report for a basic nuclear installation, significant event report for radioactive material transport, declaration criteria for significant events related to the safety of non-PWR basic nuclear installations, declaration criteria for significant events related to PWR safety, significant events declared further to events resulting in group 1 unavailability and non-compliance with technical operating specifications, declaration criteria for significant events concerning radiation protection for basic nuclear installations, declaration criteria for significant events concerning environmental protection, applicable to basic nuclear installations, and declaration criteria for significant events concerning radioactive material transport

  13. Controlling Energy Radiations of Electromagnetic Waves via Frequency Coding Metamaterials.

    Science.gov (United States)

    Wu, Haotian; Liu, Shuo; Wan, Xiang; Zhang, Lei; Wang, Dan; Li, Lianlin; Cui, Tie Jun

    2017-09-01

    Metamaterials are artificial structures composed of subwavelength unit cells to control electromagnetic (EM) waves. The spatial coding representation of metamaterial has the ability to describe the material in a digital way. The spatial coding metamaterials are typically constructed by unit cells that have similar shapes with fixed functionality. Here, the concept of frequency coding metamaterial is proposed, which achieves different controls of EM energy radiations with a fixed spatial coding pattern when the frequency changes. In this case, not only different phase responses of the unit cells are considered, but also different phase sensitivities are also required. Due to different frequency sensitivities of unit cells, two units with the same phase response at the initial frequency may have different phase responses at higher frequency. To describe the frequency coding property of unit cell, digitalized frequency sensitivity is proposed, in which the units are encoded with digits "0" and "1" to represent the low and high phase sensitivities, respectively. By this merit, two degrees of freedom, spatial coding and frequency coding, are obtained to control the EM energy radiations by a new class of frequency-spatial coding metamaterials. The above concepts and physical phenomena are confirmed by numerical simulations and experiments.

  14. FLUKA A multi-particle transport code (program version 2005)

    CERN Document Server

    Ferrari, A; Fassò, A; Ranft, Johannes

    2005-01-01

    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.

  15. 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

  16. Progress in nuclear well logging modeling using deterministic transport codes

    International Nuclear Information System (INIS)

    Kodeli, I.; Aldama, D.L.; Maucec, M.; Trkov, A.

    2002-01-01

    Further studies in continuation of the work presented in 2001 in Portoroz were performed in order to study and improve the performances, precission and domain of application of the deterministic transport codes with respect to the oil well logging analysis. These codes are in particular expected to complement the Monte Carlo solutions, since they can provide a detailed particle flux distribution in the whole geometry in a very reasonable CPU time. Real-time calculation can be envisaged. The performances of deterministic transport methods were compared to those of the Monte Carlo method. IRTMBA generic benchmark was analysed using the codes MCNP-4C and DORT/TORT. Centric as well as excentric casings were considered using 14 MeV point neutron source and NaI scintillation detectors. Neutron and gamma spectra were compared at two detector positions.(author)

  17. Evaluation of dose equivalent rate distribution in JCO critical accident by radiation transport calculation

    CERN Document Server

    Sakamoto, Y

    2002-01-01

    In the prevention of nuclear disaster, there needs the information on the dose equivalent rate distribution inside and outside the site, and energy spectra. The three dimensional radiation transport calculation code is a useful tool for the site specific detailed analysis with the consideration of facility structures. It is important in the prediction of individual doses in the future countermeasure that the reliability of the evaluation methods of dose equivalent rate distribution and energy spectra by using of Monte Carlo radiation transport calculation code, and the factors which influence the dose equivalent rate distribution outside the site are confirmed. The reliability of radiation transport calculation code and the influence factors of dose equivalent rate distribution were examined through the analyses of critical accident at JCO's uranium processing plant occurred on September 30, 1999. The radiation transport calculations including the burn-up calculations were done by using of the structural info...

  18. 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

  19. Towards a heavy-ion transport capability in the MARS15 Code

    International Nuclear Information System (INIS)

    Mokhov, N.V.; Gudima, K.K.; Mashnik, S.G.; Rakhno, I.L.; Striganov, S.

    2004-01-01

    In order to meet the challenges of new accelerator and space projects and further improve modelling of radiation effects in microscopic objects, heavy-ion interaction and transport physics have been recently incorporated into the MARS15 Monte Carlo code. A brief description of new modules is given in comparison with experimental data. The MARS Monte Carlo code is widely used in numerous accelerator, detector, shielding and cosmic ray applications. The needs of the Relativistic Heavy-Ion Collider, Large Hadron Collider, Rare Isotope Accelerator and NASA projects have recently induced adding heavy-ion interaction and transport physics to the MARS15 code. The key modules of the new implementation are described below along with their comparisons to experimental data.

  20. Assessment of shielding analysis methods, codes, and data for spent fuel transport/storage applications

    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

  1. 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

  2. Code of practice : safe use of ionizing radiation

    International Nuclear Information System (INIS)

    1988-07-01

    Ionizing radiation is used extensively in the field of scientific research. The risk of uncontrolled exposure to both the worker and the environment is ever present. The purpose of this Code is to set out practices considered by the CSIRO Health and Safety Committee to be appropriate for CSIRO staff and, if followed, they will result in appropriate protection for research staff and the environment. The Code does not cover sources of non-ionizing radiation such as microwave ovens, RF generators and laser sources

  3. Colloid transport code-nuclear user`s manual

    Energy Technology Data Exchange (ETDEWEB)

    Jain, R. [New Mexico Univ., Albuquerque, NM (United States)

    1992-04-03

    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.

  4. Simulation of transport in the ignited ITER with 1.5-D predictive code

    International Nuclear Information System (INIS)

    Becker, G.

    1995-01-01

    The confinement in the bulk and scrape-off layer plasmas of the ITER EDA and CDA is investigated with special versions of the 1.5-D BALDUR predictive transport code for the case of peaked density profiles (C υ = 1.0). The code self-consistently computes 2-D equilibria and solves 1-D transport equations with empirical transport coefficients for the ohmic, L and ELMy H mode regimes. Self-sustained steady state thermonuclear burn is demonstrated for up to 500 s. It is shown to be compatible with the strong radiation losses for divertor heat load reduction caused by the seeded impurities iron, neon and argon. The corresponding global and local energy and particle transport are presented. The required radiation corrected energy confinement times of the EDA and CDA are found to be close to 4 s. In the reference cases, the steady state helium fraction is 7%. The fractions of iron, neon and argon needed for the prescribed radiative power loss are given. It is shown that high radiative losses from the confinement zone, mainly by bremsstrahlung, cannot be avoided. The radiation profiles of iron and argon are found to be the same, with two thirds of the total radiation being emitted from closed flux surfaces. Fuel dilution due to iron and argon is small. The neon radiation is more peripheral. But neon is found to cause high fuel dilution. The combined dilution effect by helium and neon conflicts with burn control, self-sustained burn and divertor power reduction. Raising the helium fraction above 10% leads to the same difficulties owing to fuel dilution. The high helium levels of the present EDA design are thus unacceptable. The bootstrap current has only a small impact on the current profile. The sawtooth dominated region is found to cover 35% of the plasma cross-section. Local stability analysis of ideal ballooning modes shows that the plasma is everywhere well below the stability limit. 23 refs, 34 figs, 3 tabs

  5. Solution of charged particle transport equation by Monte-Carlo method in the BRANDZ code system

    International Nuclear Information System (INIS)

    Artamonov, S.N.; Androsenko, P.A.; Androsenko, A.A.

    1992-01-01

    Consideration is given to the issues of Monte-Carlo employment for the solution of charged particle transport equation and its implementation in the BRANDZ code system under the conditions of real 3D geometry and all the data available on radiation-to-matter interaction in multicomponent and multilayer targets. For the solution of implantation problem the results of BRANDZ data comparison with the experiments and calculations by other codes in complexes systems are presented. The results of direct nuclear pumping process simulation for laser-active media by a proton beam are also included. 4 refs.; 7 figs

  6. Nonsteady heat conduction code with radiation boundary conditions

    International Nuclear Information System (INIS)

    Fillo, J.A.; Benenati, R.; Powell, J.

    1975-01-01

    A heat-transfer model for studying the temperature build-up in graphite blankets for fusion reactors is presented. In essence, the computer code developed is for two-dimensional, nonsteady heat conduction in heterogeneous, anisotropic solids with nonuniform internal heating. Thermal radiation as well as bremsstrahlung radiation boundary conditions are included. Numerical calculations are performed for two design options by varying the wall loading, bremsstrahlung, surface layer thickness and thermal conductivity, blanket dimensions, time step and grid size. (auth)

  7. Antiproton annihilation physics annihilation physics in the Monte Carlo particle transport code particle transport code SHIELD-HIT12A

    DEFF Research Database (Denmark)

    Taasti, Vicki Trier; Knudsen, Helge; Holzscheiter, Michael

    2015-01-01

    The Monte Carlo particle transport code SHIELD-HIT12A is designed to simulate therapeutic beams for cancer radiotherapy with fast ions. SHIELD-HIT12A allows creation of antiproton beam kernels for the treatment planning system TRiP98, but first it must be benchmarked against experimental data. An...

  8. 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

  9. 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.

  10. Intercomparison on the usage of computational codes in radiation dosimetry

    International Nuclear Information System (INIS)

    Ilic, R.; Pesic, M.; Pavlovic, R.

    2003-01-01

    SRNA-2KG software package was modified for this work to include necessary input and output data and for predicted voxelized geometry and dosimetry. SRNA is a Monte Carlo code developed for applications in proton transport, radiotherapy and dosimetry. Protons within energy range from 100 keV to 250 MeV with predefined spectra are transported in 3D geometry through material zones confined by planes and second order surfaces or in 3D voxelized geometry. The code can treat proton transport in a few hundred different materials including elements from Z=1 to Z=98. Simulation of proton transport is based on the multiple scattering theory of charged particles and on the model for compound nucleus decay

  11. Code of practice and design principles for portable and transportable radiological protection systems

    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.)

  12. 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

  13. ITS Version 3.0: The Integrated TIGER Series of coupled electron/photon Monte Carlo transport codes

    Energy Technology Data Exchange (ETDEWEB)

    Halbleib, J.A.; Kensek, R.P.; Valdez, G.D.; Mehlhorn, T.A. [Sandia National Labs., Albuquerque, NM (United States); Seltzer, S.M.; Berger, M.J. [National Inst. of Standards and Technology, Gaithersburg, MD (United States). Ionizing Radiation Div.

    1993-06-01

    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.

  14. 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)

  15. Code of practice for radiation protection in nuclear medicine

    International Nuclear Information System (INIS)

    Hamed, M. I.

    2010-05-01

    In aim of this study was to develop a draft for a new code practice for radiation protection in nuclear medicine that meets the current relevant international recommendation. The draft includes the following main fields: methods of radiation protection for workers, patients and public. Also, the principles of safe design of nuclear medicine departments, quality assurance program, proper manipulation of radiation sources including radioactive waste and emergency preparedness and response. The practical part of this study includes inspections of three nuclear medicine departments available in Sudan so as to assess the degree of compliance of those departments with what is stated in this code. The inspection missions have been conducted using a checklist that addresses all items that may affect radiation raincoat issues in addition to per formin area radiation monitoring around the installation of the radioactive sources. The results of this revealed that most of the departments do not have effective radiation protection program which in turn could lead to unnecessary exposure to patients, public and workers. Finally, some recommendations are given that - if implemented - could improve the status of radiation protection in nuclear medicine department. (Author)

  16. Validation of the 3D finite element transport theory code EVENT for shielding applications

    International Nuclear Information System (INIS)

    Warner, Paul; Oliveira, R.E. de

    2000-01-01

    This paper is concerned with the validation of the 3D deterministic neutral-particle transport theory code EVENT for shielding applications. The code is based on the finite element-spherical harmonics (FE-P N ) method which has been extensively developed over the last decade. A general multi-group, anisotropic scattering formalism enables the code to address realistic steady state and time dependent, multi-dimensional coupled neutron/gamma radiation transport problems involving high scattering and deep penetration alike. The powerful geometrical flexibility and competitive computational effort makes the code an attractive tool for shielding applications. In recognition of this, EVENT is currently in the process of being adopted by the UK nuclear industry. The theory behind EVENT is described and its numerical implementation is outlined. Numerical results obtained by the code are compared with predictions of the Monte Carlo code MCBEND and also with the results from benchmark shielding experiments. In particular, results are presented for the ASPIS experimental configuration for both neutron and gamma ray calculations using the BUGLE 96 nuclear data library. (author)

  17. Documentation for TRACE: an interactive beam-transport code

    International Nuclear Information System (INIS)

    Crandall, K.R.; Rusthoi, D.P.

    1985-01-01

    TRACE is an interactive, first-order, beam-dynamics computer program. TRACE includes space-charge forces and mathematical models for a number of beamline elements not commonly found in beam-transport codes, such as permanent-magnet quadrupoles, rf quadrupoles, rf gaps, accelerator columns, and accelerator tanks. TRACE provides an immediate graphic display of calculative results, has a powerful and easy-to-use command procedure, includes eight different types of beam-matching or -fitting capabilities, and contains its own internal HELP package. This report describes the models and equations used for each of the transport elements, the fitting procedures, and the space-charge/emittance calculations, and provides detailed instruction for using the code

  18. Automatic modeling for the monte carlo transport TRIPOLI code

    International Nuclear Information System (INIS)

    Zhang Junjun; Zeng Qin; Wu Yican; Wang Guozhong; FDS Team

    2010-01-01

    TRIPOLI, developed by CEA, France, is Monte Carlo particle transport simulation code. It has been widely applied to nuclear physics, shielding design, evaluation of nuclear safety. However, it is time-consuming and error-prone to manually describe the TRIPOLI input file. This paper implemented bi-directional conversion between CAD model and TRIPOLI model. Its feasibility and efficiency have been demonstrated by several benchmarking examples. (authors)

  19. BALDUR: a one-dimensional plasma transport code

    International Nuclear Information System (INIS)

    Singer, C.E.; Post, D.E.; Mikkelsen, D.R.

    1986-07-01

    The purpose of BALDUR is to calculate the evolution of plasma parameters in an MHD equilibrium which can be approximated by concentric circular flux surfaces. Transport of up to six species of ionized particles, of electron and ion energy, and of poloidal magnetic flux is computed. A wide variety of source terms are calculated including those due to neutral gas, fusion, and auxiliary heating. The code is primarily designed for modeling tokamak plasmas but could be adapted to other toroidal confinement systems

  20. Parallelization of a Monte Carlo particle transport simulation code

    Science.gov (United States)

    Hadjidoukas, P.; Bousis, C.; Emfietzoglou, D.

    2010-05-01

    We have developed a high performance version of the Monte Carlo particle transport simulation code MC4. The original application code, developed in Visual Basic for Applications (VBA) for Microsoft Excel, was first rewritten in the C programming language for improving code portability. Several pseudo-random number generators have been also integrated and studied. The new MC4 version was then parallelized for shared and distributed-memory multiprocessor systems using the Message Passing Interface. Two parallel pseudo-random number generator libraries (SPRNG and DCMT) have been seamlessly integrated. The performance speedup of parallel MC4 has been studied on a variety of parallel computing architectures including an Intel Xeon server with 4 dual-core processors, a Sun cluster consisting of 16 nodes of 2 dual-core AMD Opteron processors and a 200 dual-processor HP cluster. For large problem size, which is limited only by the physical memory of the multiprocessor server, the speedup results are almost linear on all systems. We have validated the parallel implementation against the serial VBA and C implementations using the same random number generator. Our experimental results on the transport and energy loss of electrons in a water medium show that the serial and parallel codes are equivalent in accuracy. The present improvements allow for studying of higher particle energies with the use of more accurate physical models, and improve statistics as more particles tracks can be simulated in low response time.

  1. BERMUDA-1DG: a one-dimensional photon transport code

    International Nuclear Information System (INIS)

    Suzuki, Tomoo; Hasegawa, Akira; Nakashima, Hiroshi; Kaneko, Kunio.

    1984-10-01

    A one-dimensional photon transport code BERMUDA-1DG has been developed for spherical and infinite slab geometries. The purpose of development is to equip the function of gamma rays calculation for the BERMUDA code system, which was developed by 1983 only for neutron transport calculation as a preliminary version. A group constants library has been prepared for 30 nuclides, and it now consists of the 36-group total cross sections and secondary gamma ray yields by the 120-group neutron flux. For the Compton scattering, group-angle transfer matrices are accurately obtained by integrating the Klein-Nishina formula taking into account the energy and scattering angle correlation. The pair production cross sections are now calculated in the code from atomic number and midenergy of each group. To obtain angular flux distribution, the transport equation is solved in the same way as in case of neutron, using the direct integration method in a multigroup model. Both of an independent gamma ray source problem and a neutron-gamma source problem are possible to be solved. This report is written as a user's manual with a brief description of the calculational method. (author)

  2. 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)

  3. 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

  4. Trade and transport of radiation sources

    International Nuclear Information System (INIS)

    1996-01-01

    The guide specifies the obligations pertaining to the trade in and transport of radiation sources and other matters to be taken into account in safety supervision. It also specifies obligations and procedures relating to transfrontier movements of radioactive waste contained in the EU Council Directive 92/3/Euratom. (7 refs.)

  5. LDRD Final Review: Radiation Transport Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Goorley, John Timothy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Morgan, George Lake [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lestone, John Paul [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-06-22

    Both high-fidelity & toy simulations are being used to understand measured signals and improve the Area 11 NDSE diagnostic. We continue to gain more and more confidence in the ability for MCNP to simulate neutron and photon transport from source to radiation detector.

  6. Application of OMEGA Monte Carlo codes for radiation therapy treatment planning

    International Nuclear Information System (INIS)

    Ayyangar, Komanduri M.; Jiang, Steve B.

    1998-01-01

    The accuracy of conventional dose algorithms for radiosurgery treatment planning is limited, due to the inadequate consideration of the lateral radiation transport and the difficulty of acquiring accurate dosimetric data for very small beams. In the present paper, some initial work on the application of Monte Carlo method in radiation treatment planning in general, and in radiosurgery treatment planning in particular, has been presented. Two OMEGA Monte Carlo codes, BEAM and DOSXYZ, are used. The BEAM code is used to simulate the transport of particles in the linac treatment head and radiosurgery collimator. A phase space file is obtained from the BEAM simulation for each collimator size. The DOSXYZ code is used to calculate the dose distribution in the patient's body reconstructed from CT slices using the phase space file as input. The accuracy of OMEGA Monte Carlo simulation for radiosurgery dose calculation is verified by comparing the calculated and measured basic dosimetric data for several radiosurgery beams and a 4 x 4 cm 2 conventional beam. The dose distributions for three clinical cases are calculated using OMEGA codes as the dose engine for an in-house developed radiosurgery treatment planning system. The verification using basic dosimetric data and the dose calculation for clinical cases demonstrate the feasibility of applying OMEGA Monte Carlo code system to radiosurgery treatment planning. (author)

  7. The OpenMC Monte Carlo particle transport code

    International Nuclear Information System (INIS)

    Romano, Paul K.; Forget, Benoit

    2013-01-01

    Highlights: ► An open source Monte Carlo particle transport code, OpenMC, has been developed. ► 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. ► Results on a variety of benchmark problems agree with MCNP5. -- Abstract: A new Monte Carlo code called OpenMC is currently under development at the Massachusetts Institute of Technology as a tool for simulation on high-performance computing platforms. Given that many legacy codes do not scale well on existing and future parallel computer architectures, OpenMC has been developed from scratch with a focus on high performance scalable algorithms as well as modern software design practices. The present work describes the methods used in the OpenMC code and demonstrates the performance and accuracy of the code on a variety of problems.

  8. A Radiation Solver for the National Combustion Code

    Science.gov (United States)

    Sockol, Peter M.

    2015-01-01

    A methodology is given that converts an existing finite volume radiative transfer method that requires input of local absorption coefficients to one that can treat a mixture of combustion gases and compute the coefficients on the fly from the local mixture properties. The Full-spectrum k-distribution method is used to transform the radiative transfer equation (RTE) to an alternate wave number variable, g . The coefficients in the transformed equation are calculated at discrete temperatures and participating species mole fractions that span the values of the problem for each value of g. These results are stored in a table and interpolation is used to find the coefficients at every cell in the field. Finally, the transformed RTE is solved for each g and Gaussian quadrature is used to find the radiant heat flux throughout the field. The present implementation is in an existing cartesian/cylindrical grid radiative transfer code and the local mixture properties are given by a solution of the National Combustion Code (NCC) on the same grid. Based on this work the intention is to apply this method to an existing unstructured grid radiation code which can then be coupled directly to NCC.

  9. 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)

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

    CERN Document Server

    Pinsky, L; Ferrari, A; Sala, P; Carminati, F; Brun, R

    2001-01-01

    This NASA funded project is proceeding to develop a Monte Carlo-based computer simulation of the radiation environment in space. With actual funding only initially in place at the end of May 2000, the study is still in the early stage of development. The general tasks have been identified and personnel have been selected. The code to be assembled will be based upon two major existing software packages. The radiation transport simulation will be accomplished by updating the FLUKA Monte Carlo program, and the user interface will employ the ROOT software being developed at CERN. The end-product will be a Monte Carlo-based code which will complement the existing analytic codes such as BRYNTRN/HZETRN presently used by NASA to evaluate the effects of radiation shielding in space. The planned code will possess the ability to evaluate the radiation environment for spacecraft and habitats in Earth orbit, in interplanetary space, on the lunar surface, or on a planetary surface such as Mars. Furthermore, it will be usef...

  11. Creation and utilization of a World Wide Web based space radiation effects code: SIREST

    Science.gov (United States)

    Singleterry, R. C. Jr; Wilson, J. W.; Shinn, J. L.; Tripathi, R. K.; Thibeault, S. A.; Noor, A. K.; Cucinotta, F. A.; Badavi, F. F.; Chang, C. K.; Qualls, G. D.; hide

    2001-01-01

    In order for humans and electronics to fully and safely operate in the space environment, codes like HZETRN (High Charge and Energy Transport) must be included in any designer's toolbox for design evaluation with respect to radiation damage. Currently, spacecraft designers do not have easy access to accurate radiation codes like HZETRN to evaluate their design for radiation effects on humans and electronics. Today, the World Wide Web is sophisticated enough to support the entire HZETRN code and all of the associated pre and post processing tools. This package is called SIREST (Space Ionizing Radiation Effects and Shielding Tools). There are many advantages to SIREST. The most important advantage is the instant update capability of the web. Another major advantage is the modularity that the web imposes on the code. Right now, the major disadvantage of SIREST will be its modularity inside the designer's system. This mostly comes from the fact that a consistent interface between the designer and the computer system to evaluate the design is incomplete. This, however, is to be solved in the Intelligent Synthesis Environment (ISE) program currently being funded by NASA.

  12. Higher-fidelity yet efficient modeling of radiation energy transport through three-dimensional clouds

    International Nuclear Information System (INIS)

    Hall, M.L.; Davis, A.B.

    2005-01-01

    Accurate modeling of radiative energy transport through cloudy atmospheres is necessary for both climate modeling with GCMs (Global Climate Models) and remote sensing. Previous modeling efforts have taken advantage of extreme aspect ratios (cells that are very wide horizontally) by assuming a 1-D treatment vertically - the Independent Column Approximation (ICA). Recent attempts to resolve radiation transport through the clouds have drastically changed the aspect ratios of the cells, moving them closer to unity, such that the ICA model is no longer valid. We aim to provide a higher-fidelity atmospheric radiation transport model which increases accuracy while maintaining efficiency. To that end, this paper describes the development of an efficient 3-D-capable radiation code that can be easily integrated into cloud resolving models as an alternative to the resident 1-D model. Applications to test cases from the Intercomparison of 3-D Radiation Codes (I3RC) protocol are shown

  13. Aqueous Transport Code Revisions Using Geographic Information Systems

    International Nuclear Information System (INIS)

    Chen, K.F.

    2003-01-01

    STREAM II, developed at the Savannah River Site (SRS) for execution on a personal computer, is an emergency response code that predicts downstream pollutant concentrations for releases from the SRS area to the Savannah River for emergency response management decision making. The STREAM II code consists of pre-processor, calculation, and post-processor modules. The pre-processor module provides a graphical user interface (GUI) for inputting the initial release data. The GUI passes the user specified data to the calculation module that calculates the pollutant concentrations at downstream locations and the transport times. The calculation module of the STREAM II adopts the transport module of the WASP5 code. WASP5 is a US Environmental Protection Agency water quality analysis program that simulates pollutant transport and fate through surface water using a finite difference method to solve the transport equation. The calculated downstream pollutant concentrations and travel times a re passed to the post-processor for display on the computer screen in graphical and tabular forms. To minimize the user's effort in the emergency situation, the required input parameters are limited to the time and date of release, type of release, location of release, amount and duration of release, and the calculation units. The user, however, could only select one of the seventeen predetermined locations. Hence, STREAM II could not be used for situations in which release locations differ from the seventeen predetermined locations. To eliminate this limitation, STREAM II has been revised to allow users to select the release location anywhere along the specified SRS main streams or the Savannah River by mouse-selection from a map displayed on the computer monitor. The required modifications to STREAM II using geographic information systems (GIS) software is discussed in this paper

  14. TART 2000: A Coupled Neutron-Photon, 3-D, Combinatorial Geometry, Time Dependent, Monte Carlo Transport Code

    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

  15. TART 2000 A Coupled Neutron-Photon, 3-D, Combinatorial Geometry, Time Dependent, Monte Carlo Transport Code

    CERN Document Server

    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.

  16. Department of Environmental and Radiation Transport Physics - Overview

    International Nuclear Information System (INIS)

    Woznicka, U.

    2001-01-01

    Full text: We deal with environmental physics and the radiation transport physics, both theoretically and experimentally. Some results find their way to practical applications. Our environmental physics research encompasses hydrogeological problems as well as measurements of trace elements in the atmosphere and in the water. Theoretical (analytical and numerical) and experimental issues of the radiation transport and radiation fields are our main field of research. The interest in radiation transport phenomena is stimulated by their importance for the environmental physics, industrial and nuclear facilities and methods of geophysical. Environmental isotopes and noble gases are used in the investigation of water-bearing geological formations in order to determine the origin and age of groundwater. The papers listed below and three ''Reports on research'' present recent achievements in this field. The gas chromatography methods are used for monitoring the anthropogenic trace gases (SF 6 and freons), which participate in the Earth green-house effect. A very high detection level of SF 6 in water, 0.0028 fg/cm 3 H 2 0, has been reached as required for hydrogeological purposes. A preliminary verification of the SF 6 tracer method for dating young groundwaters by the tritium method has been carried out. We carried on the work on a method of radon measurement in soil in connection with geological conditions. The national seminar ''Radon in Environment'' organized at the INP aroused an interest of Polish scientific centres in that field. The seminar gathered 60 participants who presented 24 oral reports and 8 posters. Within the scope of the radiation transport physics we studied thermal neutron transport in finite hydrogenous media. Advantages and limitations of a Monte Carlo code (MCNP) in thermal neutron transport simulations have been examined by both the analytical solution and the experiment on the INP pulsed neutron generator. An interesting contribution to the

  17. APOLLO-2: An advanced transport code for LWRs

    International Nuclear Information System (INIS)

    Mathonniere, G.

    1995-01-01

    APOLLO-2 is a fully modular code in which each module corresponds to a specific task: access to the cross-sections libraries, creation of isotopes medium or mixtures, geometry definition, self-shielding calculations, computation of multigroup collision probabilities, flux solver, depletion calculations, transport-transport or transport-diffusion equivalence process, SN calculations, etc... Modules communicate exclusively by ''objects'' containing structured data, these objects are identified and handled by user's given names. Among the major improvements offered by APOLLO-2 the modelization of the self-shielding: it is possible now to deal with a great precision, checked versus Montecarlo calculations, a fuel rod divided into several concentric rings. So the total production of Plutonium is quite better estimated than before and its radial distribution may be predicted also with a good accuracy. Thanks to the versatility of the code some reference calculations and routine ones may be compared easily because only one parameter is changed; for example the self-shielding approximations are modified, the libraries or the flux solver being exactly the same. Other interesting features have been introduced in APOLLO-2: the new isotopes JEF.2 are available in 99 and 172 energy groups libraries, the surface leakage model improves the calculation of the control rod efficiency, the flux-current method allows faster calculations, the possibility of an automatic convergence checking during the depletion calculations coupled with fully automatic corrections, heterogeneous diffusion coefficients used for voiding analysis. 17 refs, 1 tab

  18. ATR, Radiation Transport Models in Atmosphere at Various Altitudes

    International Nuclear Information System (INIS)

    1981-01-01

    1 - Description of problem or function: ATR is a user-oriented code for calculating quickly and simply radiation environment problems at all altitudes in the atmosphere. The code is based on parametric models of a comprehensive data base of air transport results which were generated using discrete ordinates transport techniques for infinite homogeneous air. The effects of air-ground interface and non-uniform air density are treated as perturbation corrections on homogeneous air results. ATR includes parametric models for neutrons and secondary gamma rays as a function of space, energy and source- target angle out to angles of 550 g/cm 2 of air. ATR contains parameterizations of infinite medium air transport of neutrons and secondary gamma rays and correction factors for the air-ground interface and high altitude exponential air. It responds to a series of user-oriented commands which specify the source, geometry and print options to output a variety of useful air transport information, including energy-angle dependent fluence, dose, current, and isodose ranges. 2 - Method of solution: The version 3 differs from earlier versions in that version 3 contains the parameterization of the new neutron and secondary gamma rays data base that was calculated using the latest DNA approved cross sections for air. Other improvements to the ATR code include: parameterization and inclusion into ATR of new air- over-ground correction factors, low energy x-rays calculations, new fission source, and new convenience options. 3 - Restrictions on the complexity of the problem: ATR takes approximately 36,000 decimal words of storage. This can be lessened by overlaying different parts of the code

  19. MQRAD, a computer code for synchrotron radiation from quadrupole magnets

    International Nuclear Information System (INIS)

    Morimoto, Teruhisa.

    1984-01-01

    The computer code, MQRAD, is developed for the calculation of the synchrotron radiation from the particles passing through quadrupole magnets at the straight section of the electron-positron colliding machine. This code computes the distributions of photon numbers and photon energies at any given points on the beam orbit. In this code, elements such as the quadrupole magnets and the drift spaces can be divided into many sub-elements in order to obtain the results with good accuracy. The synchrotron radiation produced by inserted quadrupole magnets at the interaction region of the electron-positron collider is one of the main background sources to the detector. The masking system against the synchrotron radiation at TRISTAN is very important because of the relatively high beam energy and the long straight section, which are 30 GeV and 100 meters, respectively. MQRAD has been used to design the masking system of the TOPAZ detector and the result is presented here as an example. (author)

  20. Monte Carlo method in radiation transport problems

    International Nuclear Information System (INIS)

    Dejonghe, G.; Nimal, J.C.; Vergnaud, T.

    1986-11-01

    In neutral radiation transport problems (neutrons, photons), two values are important: the flux in the phase space and the density of particles. To solve the problem with Monte Carlo method leads to, among other things, build a statistical process (called the play) and to provide a numerical value to a variable x (this attribution is called score). Sampling techniques are presented. Play biasing necessity is proved. A biased simulation is made. At last, the current developments (rewriting of programs for instance) are presented due to several reasons: two of them are the vectorial calculation apparition and the photon and neutron transport in vacancy media [fr

  1. Simplified model for radioactive contaminant transport: the TRANSS code

    International Nuclear Information System (INIS)

    Simmons, C.S.; Kincaid, C.T.; Reisenauer, A.E.

    1986-09-01

    A simplified ground-water transport model called TRANSS was devised to estimate the rate of migration of a decaying radionuclide that is subject to sorption governed by a linear isotherm. Transport is modeled as a contaminant mass transmitted along a collection of streamlines constituting a streamtube, which connects a source release zone with an environmental arrival zone. The probability-weighted contaminant arrival distribution along each streamline is represented by an analytical solution of the one-dimensional advection-dispersion equation with constant velocity and dispersion coefficient. The appropriate effective constant velocity for each streamline is based on the exact travel time required to traverse a streamline with a known length. An assumption used in the model to facilitate the mathematical simplification is that transverse dispersion within a streamtube is negligible. Release of contaminant from a source is described in terms of a fraction-remaining curve provided as input information. However, an option included in the code is the calculation of a fraction-remaining curve based on four specialized release models: (1) constant release rate, (2) solubility-controlled release, (3) adsorption-controlled release, and (4) diffusion-controlled release from beneath an infiltration barrier. To apply the code, a user supplies only a certain minimal number of parameters: a probability-weighted list of travel times for streamlines, a local-scale dispersion coefficient, a sorption distribution coefficient, total initial radionuclide inventory, radioactive half-life, a release model choice, and size dimensions of the source. The code is intended to provide scoping estimates of contaminant transport and does not predict the evolution of a concentration distribution in a ground-water flow field. Moreover, the required travel times along streamlines must be obtained from a prior ground-water flow simulation

  2. Radiation transport: Progress report, July 1, 1987-September 30, 1987

    International Nuclear Information System (INIS)

    O'Dell, R.D.; Nagy, A.

    1988-05-01

    Research and development progress in radiation transport for the Los Alamos National Laboratory's Group S-6 for the fourth quarter of FY 87 is reported. Included are unclassified tasks in the areas of Deterministic Radiation Transport, Monte Carlo Radiation Transport, and Cross Sections and Physics. 23 refs., 9 figs

  3. The radiation protection code of practice in teletherapy facilities

    International Nuclear Information System (INIS)

    Fadlalla, N. S. M.

    2010-05-01

    This study aimed to provide a document (code) for the standard practice in teletherapy facilities to be a reference and guide for establishing new teletherapy facilities or mending an existing one, another aim was to evaluated the teletherapy facilities with regard to their compliance to the recommendations and guides mentioned in this document. This document includes: safety specifications for teletherapy equipment, facility planning and shielding design, radiation protection and work practice, quality assurance and personnel requirements and responsibilities. In order to assess the degree of compliance of the two centers in the country with what was stated in the developed document IAEA inspection checklist was utilized and made some radiation measurement were made around the treatment rooms. The results of such inspection mission revealed that the current status of radiation protection in both of inspected centers is almost similar and both are not satisfactory as many of the essential items of radiation protection as stipulated in this document were not followed, which lead to unnecessary, radiation exposure to patients and staff. Finally, some recommendations that may help to improve the status of radiation protection in radiotherapy departments in Sudan are given. (Author)

  4. Improvements to the National Transport Code Collaboration Data Server

    Science.gov (United States)

    Alexander, David A.

    2001-10-01

    The data server of the National Transport Code Colaboration Project provides a universal network interface to interpolated or raw transport data accessible by a universal set of names. Data can be acquired from a local copy of the Iternational Multi-Tokamak (ITER) profile database as well as from TRANSP trees of MDS Plus data systems on the net. Data is provided to the user's network client via a CORBA interface, thus providing stateful data server instances, which have the advantage of remembering the desired interpolation, data set, etc. This paper will review the status and discuss the recent improvements made to the data server, such as the modularization of the data server and the addition of hdf5 and MDS Plus data file writing capability.

  5. History of one family of atmospheric radiative transfer codes

    Science.gov (United States)

    Anderson, Gail P.; Wang, Jinxue; Hoke, Michael L.; Kneizys, F. X.; Chetwynd, James H., Jr.; Rothman, Laurence S.; Kimball, L. M.; McClatchey, Robert A.; Shettle, Eric P.; Clough, Shepard (.; Gallery, William O.; Abreu, Leonard W.; Selby, John E. A.

    1994-12-01

    Beginning in the early 1970's, the then Air Force Cambridge Research Laboratory initiated a program to develop computer-based atmospheric radiative transfer algorithms. The first attempts were translations of graphical procedures described in a 1970 report on The Optical Properties of the Atmosphere, based on empirical transmission functions and effective absorption coefficients derived primarily from controlled laboratory transmittance measurements. The fact that spectrally-averaged atmospheric transmittance (T) does not obey the Beer-Lambert Law (T equals exp(-(sigma) (DOT)(eta) ), where (sigma) is a species absorption cross section, independent of (eta) , the species column amount along the path) at any but the finest spectral resolution was already well known. Band models to describe this gross behavior were developed in the 1950's and 60's. Thus began LOWTRAN, the Low Resolution Transmittance Code, first released in 1972. This limited initial effort has how progressed to a set of codes and related algorithms (including line-of-sight spectral geometry, direct and scattered radiance and irradiance, non-local thermodynamic equilibrium, etc.) that contain thousands of coding lines, hundreds of subroutines, and improved accuracy, efficiency, and, ultimately, accessibility. This review will include LOWTRAN, HITRAN (atlas of high-resolution molecular spectroscopic data), FASCODE (Fast Atmospheric Signature Code), and MODTRAN (Moderate Resolution Transmittance Code), their permutations, validations, and applications, particularly as related to passive remote sensing and energy deposition.

  6. An Implementation of Interfacial Transport Equation into the CUPID code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ik Kyu; Cho, Heong Kyu; Yoon, Han Young; Jeong, Jae Jun

    2009-11-15

    A component scale thermal hydraulic analysis code, CUPID (Component Unstructured Program for Interfacial Dynamics), is being developed for the analysis of components for a nuclear reactor, such as reactor vessel, steam generator, containment, etc. It adopted a three-dimensional, transient, two phase and three-field model. In order to develop the numerical schemes for the three-field model, various numerical schemes have been examined including the SMAS, semi-implicit ICE, SIMPLE. The governing equations for a 2-phase flow are composed of mass, momentum, and energy conservation equations for each phase. These equation sets are closed by the interfacial transfer rate of mass, momentum, and energy. The interfacial transfer of mass, momentum, and energy occurs through the interfacial area, and this area plays an important role in the transfer rate. The flow regime based correlations are used for calculating the interracial area in the traditional style 2-phase flow model. This is dependent upon the flow regime and is limited to the fully developed 2-phase flow region. Its application to the multi-dimensional 2-phase flow has some limitation because it adopts the measured results of 2-phase flow in the 1-dimensional tube. The interfacial area concentration transport equation had been suggested in order to calculate the interfacial area without the interfacial area correlations. The source terms to close the interfacial area transport equation should be further developed for a wide ranger usage of it. In this study, the one group interfacial area concentration transport equation has been implemented into the CUPID code. This interfacial area concentration transport equation can be used instead of the interfacial area concentration correlations for the bubbly flow region.

  7. An Implementation of Interfacial Transport Equation into the CUPID code

    International Nuclear Information System (INIS)

    Park, Ik Kyu; Cho, Heong Kyu; Yoon, Han Young; Jeong, Jae Jun

    2009-11-01

    A component scale thermal hydraulic analysis code, CUPID (Component Unstructured Program for Interfacial Dynamics), is being developed for the analysis of components for a nuclear reactor, such as reactor vessel, steam generator, containment, etc. It adopted a three-dimensional, transient, two phase and three-field model. In order to develop the numerical schemes for the three-field model, various numerical schemes have been examined including the SMAS, semi-implicit ICE, SIMPLE. The governing equations for a 2-phase flow are composed of mass, momentum, and energy conservation equations for each phase. These equation sets are closed by the interfacial transfer rate of mass, momentum, and energy. The interfacial transfer of mass, momentum, and energy occurs through the interfacial area, and this area plays an important role in the transfer rate. The flow regime based correlations are used for calculating the interracial area in the traditional style 2-phase flow model. This is dependent upon the flow regime and is limited to the fully developed 2-phase flow region. Its application to the multi-dimensional 2-phase flow has some limitation because it adopts the measured results of 2-phase flow in the 1-dimensional tube. The interfacial area concentration transport equation had been suggested in order to calculate the interfacial area without the interfacial area correlations. The source terms to close the interfacial area transport equation should be further developed for a wide ranger usage of it. In this study, the one group interfacial area concentration transport equation has been implemented into the CUPID code. This interfacial area concentration transport equation can be used instead of the interfacial area concentration correlations for the bubbly flow region

  8. Fallout radiation protection provided by transportation vehicles

    Energy Technology Data Exchange (ETDEWEB)

    Burson, Z.G.

    1972-10-20

    Fallout radiation protection factors (PF's) were estimated for a variety of civilian transportation vehicles using measurements of the natural terrain radiation as a source. The PF values are below 2 in light vehicles, truck beds, or trailers; from 2.5 to 3 in the cabs of heavy trucks and in a railway guard car; and from 3.0 to 3.5 in the engineer's seat of heavy locomotives. This information can be useful in planning the possible movement of personnel from or through areas contaminated either by a wartime incident or a peacetime accident. The information may also be useful for studying the reduction of exposure to the natural terrestrial radiation environment provided by vehicles.

  9. Graphical-based construction of combinatorial geometries for radiation transport and shielding applications

    International Nuclear Information System (INIS)

    Burns, T.J.

    1992-01-01

    A graphical-based code system is being developed at ORNL to manipulate combinatorial geometries for radiation transport and shielding applications. The current version (basically a combinatorial geometry debugger) consists of two parts: a FORTRAN-based ''view'' generator and a Microsoft Windows application for displaying the geometry. Options and features of both modules are discussed. Examples illustrating the various options available are presented. The potential for utilizing the images produced using the debugger as a visualization tool for the output of the radiation transport codes is discussed as is the future direction of the development

  10. 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)

  11. TU-EF-304-10: Efficient Multiscale Simulation of the Proton Relative Biological Effectiveness (RBE) for DNA Double Strand Break (DSB) Induction and Bio-Effective Dose in the FLUKA Monte Carlo Radiation Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Moskvin, V; Tsiamas, P; Axente, M; Farr, J [St. Jude Children’s Research Hospital, Memphis, TN (United States); Stewart, R [University of Washington, Seattle, WA. (United States)

    2015-06-15

    Purpose: One of the more critical initiating events for reproductive cell death is the creation of a DNA double strand break (DSB). In this study, we present a computationally efficient way to determine spatial variations in the relative biological effectiveness (RBE) of proton therapy beams within the FLUKA Monte Carlo (MC) code. Methods: We used the independently tested Monte Carlo Damage Simulation (MCDS) developed by Stewart and colleagues (Radiat. Res. 176, 587–602 2011) to estimate the RBE for DSB induction of monoenergetic protons, tritium, deuterium, hellium-3, hellium-4 ions and delta-electrons. The dose-weighted (RBE) coefficients were incorporated into FLUKA to determine the equivalent {sup 6}°60Co γ-ray dose for representative proton beams incident on cells in an aerobic and anoxic environment. Results: We found that the proton beam RBE for DSB induction at the tip of the Bragg peak, including primary and secondary particles, is close to 1.2. Furthermore, the RBE increases laterally to the beam axis at the area of Bragg peak. At the distal edge, the RBE is in the range from 1.3–1.4 for cells irradiated under aerobic conditions and may be as large as 1.5–1.8 for cells irradiated under anoxic conditions. Across the plateau region, the recorded RBE for DSB induction is 1.02 for aerobic cells and 1.05 for cells irradiated under anoxic conditions. The contribution to total effective dose from secondary heavy ions decreases with depth and is higher at shallow depths (e.g., at the surface of the skin). Conclusion: Multiscale simulation of the RBE for DSB induction provides useful insights into spatial variations in proton RBE within pristine Bragg peaks. This methodology is potentially useful for the biological optimization of proton therapy for the treatment of cancer. The study highlights the need to incorporate spatial variations in proton RBE into proton therapy treatment plans.

  12. Survey of radiation protection, radiation transport, and shielding information needs of the nuclear power industry. Final report

    International Nuclear Information System (INIS)

    Maskewitz, B.F.; Trubey, D.K.; Roussin, R.W.; McGill, B.L.

    1976-04-01

    The Radiation Shielding Information Center (RSIC) is engaged in a program to seek out, organize, and disseminate information in the area of radiation transport, shielding, and radiation protection. This information consists of published literature, nuclear data, and computer codes and advanced analytical techniques required by ERDA, its contractors, and the nuclear power industry to improve radiation analysis and computing capability. Information generated in this effort becomes a part of the RSIC collection and/or data base. The purpose of this report on project 219-1 is to document the results of the survey of information and computer code needs of the nuclear power industry in the area of radiation analysis and protection

  13. Survey of radiation protection, radiation transport, and shielding information needs of the nuclear power industry. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Maskewitz, B.F.; Trubey, D.K.; Roussin, R.W.; McGill, B.L.

    1976-04-01

    The Radiation Shielding Information Center (RSIC) is engaged in a program to seek out, organize, and disseminate information in the area of radiation transport, shielding, and radiation protection. This information consists of published literature, nuclear data, and computer codes and advanced analytical techniques required by ERDA, its contractors, and the nuclear power industry to improve radiation analysis and computing capability. Information generated in this effort becomes a part of the RSIC collection and/or data base. The purpose of this report on project 219-1 is to document the results of the survey of information and computer code needs of the nuclear power industry in the area of radiation analysis and protection.

  14. Evaluation of coded aperture radiation detectors using a Bayesian approach

    Energy Technology Data Exchange (ETDEWEB)

    Miller, Kyle, E-mail: mille856@andrew.cmu.edu [Auton Lab, The Robotics Institute, Carnegie Mellon University, 5000 Forbes Avenue, Pittsburgh, PA 15213 (United States); Huggins, Peter [Auton Lab, The Robotics Institute, Carnegie Mellon University, 5000 Forbes Avenue, Pittsburgh, PA 15213 (United States); Labov, Simon; Nelson, Karl [Lawrence Livermore National Laboratory, Livermore, CA (United States); Dubrawski, Artur [Auton Lab, The Robotics Institute, Carnegie Mellon University, 5000 Forbes Avenue, Pittsburgh, PA 15213 (United States)

    2016-12-11

    We investigate tradeoffs arising from the use of coded aperture gamma-ray spectrometry to detect and localize sources of harmful radiation in the presence of noisy background. Using an example application scenario of area monitoring and search, we empirically evaluate weakly supervised spectral, spatial, and hybrid spatio-spectral algorithms for scoring individual observations, and two alternative methods of fusing evidence obtained from multiple observations. Results of our experiments confirm the intuition that directional information provided by spectrometers masked with coded aperture enables gains in source localization accuracy, but at the expense of reduced probability of detection. Losses in detection performance can however be to a substantial extent reclaimed by using our new spatial and spatio-spectral scoring methods which rely on realistic assumptions regarding masking and its impact on measured photon distributions.

  15. Code of practice for radiation protection in dentistry (1987)

    International Nuclear Information System (INIS)

    1988-01-01

    This code is intended as a guide to safe practices and lays down detailed requirements for the following protective measures : allocation of responsibility and need for clinical assessment of the indications for radiography; provision of appropriate equipment, film and processing facilities; adoption of procedures to minimise exposure to radiation. This code deals with radiographic procedures commonly used in general and specialist dental practice. These entail : intra-oral radiography: periapical, bitewing and occlusal views; panoramic radiography: intra-oral tube radiography and panoramic tomography; radiography of the complete skull or certain parts of the dento-maxillo-facial region; hand and wrist radiography for the sole purpose of the determination of the bone age

  16. 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

  17. 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.

  18. Transport methods: general. 1. The Analytical Monte Carlo Method for Radiation Transport Calculations

    International Nuclear Information System (INIS)

    Martin, William R.; Brown, Forrest B.

    2001-01-01

    We present an alternative Monte Carlo method for solving the coupled equations of radiation transport and material energy. This method is based on incorporating the analytical solution to the material energy equation directly into the Monte Carlo simulation for the radiation intensity. This method, which we call the Analytical Monte Carlo (AMC) method, differs from the well known Implicit Monte Carlo (IMC) method of Fleck and Cummings because there is no discretization of the material energy equation since it is solved as a by-product of the Monte Carlo simulation of the transport equation. Our method also differs from the method recently proposed by Ahrens and Larsen since they use Monte Carlo to solve both equations, while we are solving only the radiation transport equation with Monte Carlo, albeit with effective sources and cross sections to represent the emission sources. Our method bears some similarity to a method developed and implemented by Carter and Forest nearly three decades ago, but there are substantive differences. We have implemented our method in a simple zero-dimensional Monte Carlo code to test the feasibility of the method, and the preliminary results are very promising, justifying further extension to more realistic geometries. (authors)

  19. Implementation and display of Computer Aided Design (CAD) models in Monte Carlo radiation transport and shielding applications

    International Nuclear Information System (INIS)

    Burns, T.J.

    1994-01-01

    An Xwindow application capable of importing geometric information directly from two Computer Aided Design (CAD) based formats for use in radiation transport and shielding analyses is being developed at ORNL. The application permits the user to graphically view the geometric models imported from the two formats for verification and debugging. Previous models, specifically formatted for the radiation transport and shielding codes can also be imported. Required extensions to the existing combinatorial geometry analysis routines are discussed. Examples illustrating the various options and features which will be implemented in the application are presented. The use of the application as a visualization tool for the output of the radiation transport codes is also discussed

  20. 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)

  1. Review of the Monte Carlo and deterministic codes in radiation protection and dosimetry

    International Nuclear Information System (INIS)

    Tagziria, H.

    2000-02-01

    Modelling a physical system can be carried out either stochastically or deterministically. An example of the former method is the Monte Carlo technique, in which statistically approximate methods are applied to exact models. No transport equation is solved as individual particles are simulated and some specific aspect (tally) of their average behaviour is recorded. The average behaviour of the physical system is then inferred using the central limit theorem. In contrast, deterministic codes use mathematically exact methods that are applied to approximate models to solve the transport equation for the average particle behaviour. The physical system is subdivided in boxes in the phase-space system and particles are followed from one box to the next. The smaller the boxes the better the approximations become. Although the Monte Carlo method has been used for centuries, its more recent manifestation has really emerged from the Manhattan project of the Word War II. Its invention is thought to be mainly due to Metropolis, Ulah (through his interest in poker), Fermi, von Neuman and Richtmeyer. Over the last 20 years or so, the Monte Carlo technique has become a powerful tool in radiation transport. This is due to users taking full advantage of richer cross section data, more powerful computers and Monte Carlo techniques for radiation transport, with high quality physics and better known source spectra. This method is a common sense approach to radiation transport and its success and popularity is quite often also due to necessity, because measurements are not always possible or affordable. In the Monte Carlo method, which is inherently realistic because nature is statistical, a more detailed physics is made possible by isolation of events while rather elaborate geometries can be modelled. Provided that the physics is correct, a simulation is exactly analogous to an experimenter counting particles. In contrast to the deterministic approach, however, a disadvantage of the

  2. The EGS4 Code System: Solution of gamma-ray and electron transport problems

    International Nuclear Information System (INIS)

    Nelson, W.R.; Namito, Yoshihito.

    1990-01-01

    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

  3. The EGS4 Code System: Solution of Gamma-ray and Electron Transport Problems

    Science.gov (United States)

    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.

  4. 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)

  5. Radiation transport. Progress report, April 1-December 31, 1983

    International Nuclear Information System (INIS)

    O'Dell, R.D.

    1984-10-01

    Research and development progress in radiation transport by the Los Alamos National Laboratory's Group X-6 for the last nine months of CY 83 is reported. Included are unclassified tasks in the areas of Fission Reactor Neutronics, Deterministic Transport Methods, Monte Carlo Radiation Transport, and Cross Sections and Physics

  6. 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

  7. 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

  8. A dual-sided coded-aperture radiation detection system

    International Nuclear Information System (INIS)

    Penny, R.D.; Hood, W.E.; Polichar, R.M.; Cardone, F.H.; Chavez, L.G.; Grubbs, S.G.; Huntley, B.P.; Kuharski, R.A.; Shyffer, R.T.; Fabris, L.; Ziock, K.P.; Labov, S.E.; Nelson, K.

    2011-01-01

    We report the development of a large-area, mobile, coded-aperture radiation imaging system for localizing compact radioactive sources in three dimensions while rejecting distributed background. The 3D Stand-Off Radiation Detection System (SORDS-3D) has been tested at speeds up to 95 km/h and has detected and located sources in the millicurie range at distances of over 100 m. Radiation data are imaged to a geospatially mapped world grid with a nominal 1.25- to 2.5-m pixel pitch at distances out to 120 m on either side of the platform. Source elevation is also extracted. Imaged radiation alarms are superimposed on a side-facing video log that can be played back for direct localization of sources in buildings in urban environments. The system utilizes a 37-element array of 5x5x50 cm 3 cesium-iodide (sodium) detectors. Scintillation light is collected by a pair of photomultiplier tubes placed at either end of each detector, with the detectors achieving an energy resolution of 6.15% FWHM (662 keV) and a position resolution along their length of 5 cm FWHM. The imaging system generates a dual-sided two-dimensional image allowing users to efficiently survey a large area. Imaged radiation data and raw spectra are forwarded to the RadioNuclide Analysis Kit (RNAK), developed by our collaborators, for isotope ID. An intuitive real-time display aids users in performing searches. Detector calibration is dynamically maintained by monitoring the potassium-40 peak and digitally adjusting individual detector gains. We have recently realized improvements, both in isotope identification and in distinguishing compact sources from background, through the installation of optimal-filter reconstruction kernels.

  9. Study and application of the ANISN and DOT 3.5 codes to problems in nuclear radiation shielding

    International Nuclear Information System (INIS)

    Otto, A.C.

    1983-01-01

    The application of the Sn transport codes ANISN and DOT 3.5 to problems in radiation shielding is reviewed. In addition, a large array of codes involved in radiation shielding calculations is described and applied in this work. The ANISN and DOT 3.5 codes solve the multigroup transport equation in plane, cylindrical and spherical geometries, the first in one dimension and the second in two dimensions, by using the Sn approximation and were designed to solve coupled neutron-photon transport problems commonly found in reactor shielding calculations. In this work the numerical methods used in these codes are reviewed and their basic application to deep-penetration and void problems is discussed. Benchmark problems are solved by employing the array of codes previously mentioned. In particular, the ability of the ISOFLUXO program coupled to the DOT 3.5 code of mapping contours of regions with approximately the same scalar fluxes is illustrated, showing that they can be efficiently used in shielding analysis. (Author) [pt

  10. Discrete-ordinates electron transport calculations using standard neutron transport codes

    International Nuclear Information System (INIS)

    Morel, J.E.

    1979-01-01

    The primary purpose of this work was to develop a method for using standard neutron transport codes to perform electron transport calculations. The method is to develop approximate electron cross sections which are sufficiently well-behaved to be treated with standard S/sub n/ methods, but which nonetheless yield flux solutions which are very similar to the exact solutions. The main advantage of this approach is that, once the approximate cross sections are constructed, their multigroup Legendre expansion coefficients can be calculated and input to any standard S/sub n/ code. Discrete-ordinates calculations were performed to determine the accuracy of the flux solutions for problems corresponding to 1.0-MeV electrons incident upon slabs of aluminum and gold. All S/sub n/ calculations were compared with similar calculations performed with an electron Monte Carlo code, considered to be exact. In all cases, the discrete-ordinates solutions for integral flux quantities (i.e., scalar flux, energy deposition profiles, etc.) are generally in agreement with the Monte Carlo solutions to within approximately 5% or less. The central conclusion is that integral electron flux quantities can be efficiently and accurately calculated using standard S/sub n/ codes in conjunction with approximate cross sections. Furthermore, if group structures and approximate cross section construction are optimized, accurate differential flux energy spectra may also be obtainable without having to use an inordinately large number of energy groups. 1 figure

  11. 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

  12. Ethical codes. Fig leaf argument, ballast or cultural element for radiation protection?

    International Nuclear Information System (INIS)

    Gellermann, Rainer

    2014-01-01

    The international association for radiation protection (IRPA) adopted in May 2004 a Code of Ethics in order to allow their members to hold an adequate professional level of ethical line of action. Based on this code of ethics the professional body of radiation protection (Fachverband fuer Strahlenschutz) has developed its own ethical code and adopted in 2005.

  13. Rare event simulation in radiation transport

    International Nuclear Information System (INIS)

    Kollman, C.

    1993-10-01

    This dissertation studies methods for estimating extremely small probabilities by Monte Carlo simulation. Problems in radiation transport typically involve estimating very rare events or the expected value of a random variable which is with overwhelming probability equal to zero. These problems often have high dimensional state spaces and irregular geometries so that analytic solutions are not possible. Monte Carlo simulation must be used to estimate the radiation dosage being transported to a particular location. If the area is well shielded the probability of any one particular particle getting through is very small. Because of the large number of particles involved, even a tiny fraction penetrating the shield may represent an unacceptable level of radiation. It therefore becomes critical to be able to accurately estimate this extremely small probability. Importance sampling is a well known technique for improving the efficiency of rare event calculations. Here, a new set of probabilities is used in the simulation runs. The results are multiple by the likelihood ratio between the true and simulated probabilities so as to keep the estimator unbiased. The variance of the resulting estimator is very sensitive to which new set of transition probabilities are chosen. It is shown that a zero variance estimator does exist, but that its computation requires exact knowledge of the solution. A simple random walk with an associated killing model for the scatter of neutrons is introduced. Large deviation results for optimal importance sampling in random walks are extended to the case where killing is present. An adaptive ''learning'' algorithm for implementing importance sampling is given for more general Markov chain models of neutron scatter. For finite state spaces this algorithm is shown to give with probability one, a sequence of estimates converging exponentially fast to the true solution

  14. 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.

  15. 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

  16. Monte Carlo radiation transport: A revolution in science

    International Nuclear Information System (INIS)

    Hendricks, J.

    1993-01-01

    When Enrico Fermi, Stan Ulam, Nicholas Metropolis, John von Neuman, and Robert Richtmyer invented the Monte Carlo method fifty years ago, little could they imagine the far-flung consequences, the international applications, and the revolution in science epitomized by their abstract mathematical method. The Monte Carlo method is used in a wide variety of fields to solve exact computational models approximately by statistical sampling. It is an alternative to traditional physics modeling methods which solve approximate computational models exactly by deterministic methods. Modern computers and improved methods, such as variance reduction, have enhanced the method to the point of enabling a true predictive capability in areas such as radiation or particle transport. This predictive capability has contributed to a radical change in the way science is done: design and understanding come from computations built upon experiments rather than being limited to experiments, and the computer codes doing the computations have become the repository for physics knowledge. The MCNP Monte Carlo computer code effort at Los Alamos is an example of this revolution. Physicians unfamiliar with physics details can design cancer treatments using physics buried in the MCNP computer code. Hazardous environments and hypothetical accidents can be explored. Many other fields, from underground oil well exploration to aerospace, from physics research to energy production, from safety to bulk materials processing, benefit from MCNP, the Monte Carlo method, and the revolution in science

  17. Code of Nursing Practice for Staff Exposed to Ionizing Radiation (1984)

    International Nuclear Information System (INIS)

    1984-01-01

    This Code, published by the National Health and Medical Research Council and intended for nurses and auxiliary staff provides general guidance on radiation protection. The Code is supplementary to radiation control legislation relating to the use of ionizing radiation in medical practice. The principles established by the recommendations of the International Commission on Radiological Protection (ICRP) have been taken into account. (NEA) [fr

  18. Modelling plastic scintillator response to gamma rays using light transport incorporated FLUKA code

    Energy Technology Data Exchange (ETDEWEB)

    Ranjbar Kohan, M. [Physics Department, Tafresh University, Tafresh (Iran, Islamic Republic of); Etaati, G.R. [Department of Nuclear Engineering and Physics, Amir Kabir University of Technology, Tehran (Iran, Islamic Republic of); Ghal-Eh, N., E-mail: ghal-eh@du.ac.ir [School of Physics, Damghan University, Damghan (Iran, Islamic Republic of); Safari, M.J. [Department of Energy Engineering, Sharif University of Technology, Tehran (Iran, Islamic Republic of); Afarideh, H. [Department of Nuclear Engineering and Physics, Amir Kabir University of Technology, Tehran (Iran, Islamic Republic of); Asadi, E. [Department of Physics, Payam-e-Noor University, Tehran (Iran, Islamic Republic of)

    2012-05-15

    The response function of NE102 plastic scintillator to gamma rays has been simulated using a joint FLUKA+PHOTRACK Monte Carlo code. The multi-purpose particle transport code, FLUKA, has been responsible for gamma transport whilst the light transport code, PHOTRACK, has simulated the transport of scintillation photons through scintillator and lightguide. The simulation results of plastic scintillator with/without light guides of different surface coverings have been successfully verified with experiments. - Highlights: Black-Right-Pointing-Pointer A multi-purpose code (FLUKA) and a light transport code (PHOTRACK) have been linked. Black-Right-Pointing-Pointer The hybrid code has been used to generate the response function of an NE102 scintillator. Black-Right-Pointing-Pointer The simulated response functions exhibit a good agreement with experimental data.

  19. Transoptr-a second order beam transport design code with automatic internal optimization and general constraints

    International Nuclear Information System (INIS)

    Heighway, E.A.

    1980-07-01

    A second order beam transport design code with parametric optimization is described. The code analyzes the transport of charged particle beams through a user defined magnet system. The magnet system parameters are varied (within user defined limits) until the properties of the transported beam and/or the system transport matrix match those properties requested by the user. The code uses matrix formalism to represent the transport elements and optimization is achieved using the variable metric method. Any constraints that can be expressed algebraically may be included by the user as part of his design. Instruction in the use of the program is given. (auth)

  20. 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

  1. 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)

  2. MULTI2D - a computer code for two-dimensional radiation hydrodynamics

    Science.gov (United States)

    Ramis, R.; Meyer-ter-Vehn, J.; Ramírez, J.

    2009-06-01

    Simulation of radiation hydrodynamics in two spatial dimensions is developed, having in mind, in particular, target design for indirectly driven inertial confinement energy (IFE) and the interpretation of related experiments. Intense radiation pulses by laser or particle beams heat high-Z target configurations of different geometries and lead to a regime which is optically thick in some regions and optically thin in others. A diffusion description is inadequate in this situation. A new numerical code has been developed which describes hydrodynamics in two spatial dimensions (cylindrical R-Z geometry) and radiation transport along rays in three dimensions with the 4 π solid angle discretized in direction. Matter moves on a non-structured mesh composed of trilateral and quadrilateral elements. Radiation flux of a given direction enters on two (one) sides of a triangle and leaves on the opposite side(s) in proportion to the viewing angles depending on the geometry. This scheme allows to propagate sharply edged beams without ray tracing, though at the price of some lateral diffusion. The algorithm treats correctly both the optically thin and optically thick regimes. A symmetric semi-implicit (SSI) method is used to guarantee numerical stability. Program summaryProgram title: MULTI2D Catalogue identifier: AECV_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AECV_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 151 098 No. of bytes in distributed program, including test data, etc.: 889 622 Distribution format: tar.gz Programming language: C Computer: PC (32 bits architecture) Operating system: Linux/Unix RAM: 2 Mbytes Word size: 32 bits Classification: 19.7 External routines: X-window standard library (libX11.so) and corresponding heading files (X11/*.h) are

  3. Vectorization and parallelization of Monte-Carlo programs for calculation of radiation transport

    International Nuclear Information System (INIS)

    Seidel, R.

    1995-01-01

    The versatile MCNP-3B Monte-Carlo code written in FORTRAN77, for simulation of the radiation transport of neutral particles, has been subjected to vectorization and parallelization of essential parts, without touching its versatility. Vectorization is not dependent on a specific computer. Several sample tasks have been selected in order to test the vectorized MCNP-3B code in comparison to the scalar MNCP-3B code. The samples are a representative example of the 3-D calculations to be performed for simulation of radiation transport in neutron and reactor physics. (1) 4πneutron detector. (2) High-energy calorimeter. (3) PROTEUS benchmark (conversion rates and neutron multiplication factors for the HCLWR (High Conversion Light Water Reactor)). (orig./HP) [de

  4. Rare Event Simulation in Radiation Transport

    Science.gov (United States)

    Kollman, Craig

    This dissertation studies methods for estimating extremely small probabilities by Monte Carlo simulation. Problems in radiation transport typically involve estimating very rare events or the expected value of a random variable which is with overwhelming probability equal to zero. These problems often have high dimensional state spaces and irregular geometries so that analytic solutions are not possible. Monte Carlo simulation must be used to estimate the radiation dosage being transported to a particular location. If the area is well shielded the probability of any one particular particle getting through is very small. Because of the large number of particles involved, even a tiny fraction penetrating the shield may represent an unacceptable level of radiation. It therefore becomes critical to be able to accurately estimate this extremely small probability. Importance sampling is a well known technique for improving the efficiency of rare event calculations. Here, a new set of probabilities is used in the simulation runs. The results are multiplied by the likelihood ratio between the true and simulated probabilities so as to keep our estimator unbiased. The variance of the resulting estimator is very sensitive to which new set of transition probabilities are chosen. It is shown that a zero variance estimator does exist, but that its computation requires exact knowledge of the solution. A simple random walk with an associated killing model for the scatter of neutrons is introduced. Large deviation results for optimal importance sampling in random walks are extended to the case where killing is present. An adaptive "learning" algorithm for implementing importance sampling is given for more general Markov chain models of neutron scatter. For finite state spaces this algorithm is shown to give, with probability one, a sequence of estimates converging exponentially fast to the true solution. In the final chapter, an attempt to generalize this algorithm to a continuous

  5. Study on transport safety of refresh MOX fuel. Radiation dose from package hypothetically submerged into sea

    International Nuclear Information System (INIS)

    Tsumune, Daisuke; Suzuki; Hiroshi; Saegusa, Toshiari; Maruyama, Koki; Ito, Chihiro; Watabe, Naoto

    1999-01-01

    The sea transport of fresh MOX fuel from Europe to Japan is under planning. For the structure and equipment of transport ships for fresh MOX fuels, there is a special safety standard called the INF Code of IMO (International Maritime Organization). For transport of radioactive materials, there is a safety standard stipulated in Regulations for the Safe Transport of Radioactive Material issued by IAEA (International Atomic Energy Agency). Under those code and standard, fresh MOX fuel will be transported safely on the sea. However, a dose assessment has been made by assuming that a fresh MOX fuel package might be sunk into the sea by unexpected reasons. In the both cases for a package sunk at the coastal region and for that sunk at the ocean, the evaluated result of the dose equivalent by radiation exposure to the public are far below the dose equivalent limit of the ICRP recommendation (1 mSv/year). (author)

  6. Current status of radiation transport tools for proliferation and terrorism prevention

    International Nuclear Information System (INIS)

    Sale, K.E.

    2004-01-01

    We present the current status and future plans for the set of calculational tools and data bases developed and maintained at LLNL. The calculational tools include the Monte Carlo codes TART and COG as well as the deterministic code ARDRA. In addition to these codes presently in use there is a major development effort for a new massively parallel transport code. An important part of the capability we're developing is a sophisticated user interface, based on a commercial 3-D modeling product, to improve the model development process. A major part of this user interface tool is being developed by Strela under the Nuclear Cities Initiative. Strela has developed a hub-and-spoke technology for code input interconversions (between COG, TART and MCNP) and will produce the plug-ins that extend the capabilities of the 3-D modeler for use as a radiation transport input generator. The major advantages of this approach are the built-in user interface for 3-D modeling and the ability to read a large variety of CAD-file formats. In addition to supporting our current radiation transport codes and developing new capabilities we are working on some nuclear data needs for homeland security. These projects are carried out and the Lawrence Berkeley National Laboratory 88' cyclotron and at the Institute for Nuclear Research of the Nation Academy of Science of Ukraine under and STCU contract. (author)

  7. Physical Processes and Applications of the Monte Carlo Radiative Energy Deposition (MRED) Code

    Science.gov (United States)

    Reed, Robert A.; Weller, Robert A.; Mendenhall, Marcus H.; Fleetwood, Daniel M.; Warren, Kevin M.; Sierawski, Brian D.; King, Michael P.; Schrimpf, Ronald D.; Auden, Elizabeth C.

    2015-08-01

    MRED is a Python-language scriptable computer application that simulates radiation transport. It is the computational engine for the on-line tool CRÈME-MC. MRED is based on c++ code from Geant4 with additional Fortran components to simulate electron transport and nuclear reactions with high precision. We provide a detailed description of the structure of MRED and the implementation of the simulation of physical processes used to simulate radiation effects in electronic devices and circuits. Extensive discussion and references are provided that illustrate the validation of models used to implement specific simulations of relevant physical processes. Several applications of MRED are summarized that demonstrate its ability to predict and describe basic physical phenomena associated with irradiation of electronic circuits and devices. These include effects from single particle radiation (including both direct ionization and indirect ionization effects), dose enhancement effects, and displacement damage effects. MRED simulations have also helped to identify new single event upset mechanisms not previously observed by experiment, but since confirmed, including upsets due to muons and energetic electrons.

  8. A 3D Monte Carlo code for plasma transport in island divertors

    International Nuclear Information System (INIS)

    Feng, Y.; Sardei, F.; Kisslinger, J.; Grigull, P.

    1997-01-01

    A fully 3D self-consistent Monte Carlo code EMC3 (edge Monte Carlo 3D) for modelling the plasma transport in island divertors has been developed. In a first step, the code solves a simplified version of the 3D time-independent plasma fluid equations. Coupled to the neutral transport code EIRENE, the EMC3 code has been used to study the particle, energy and neutral transport in W7-AS island divertor configurations. First results are compared with data from different diagnostics (Langmuir probes, H α cameras and thermography). (orig.)

  9. Development of a tracer transport option for the NAPSAC fracture network computer code

    International Nuclear Information System (INIS)

    Herbert, A.W.

    1990-06-01

    The Napsac computer code predicts groundwater flow through fractured rock using a direct fracture network approach. This paper describes the development of a tracer transport algorithm for the NAPSAC code. A very efficient particle-following approach is used enabling tracer transport to be predicted through large fracture networks. The new algorithm is tested against three test examples. These demonstrations confirm the accuracy of the code for simple networks, where there is an analytical solution to the transport problem, and illustrates the use of the computer code on a more realistic problem. (author)

  10. 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.)

  11. Library system for a one dimensional tokamak transport code: (LIBJT60), 1

    International Nuclear Information System (INIS)

    Hirayama, Toshio

    1982-12-01

    A library system is developed to control and manage huge programs in terms of FORTRAN source. It is applied to widely used one dimensional tokamak transport codes (LIBJT60), which have been developed in the Division of Large Tokamak Development. The structure of data and program in the transport code turn out to be flexible enough to respond to various demands and this gigantic code frame work can be decomposed into groups of a compact code with a specific function. Some editing support tools for programming and debugging are also developed to save programming work. By applying this library system, users can obtain a code whose functions can be efficiently developed. (author)

  12. Statistics of Monte Carlo methods used in radiation transport calculation

    International Nuclear Information System (INIS)

    Datta, D.

    2009-01-01

    Radiation transport calculation can be carried out by using either deterministic or statistical methods. Radiation transport calculation based on statistical methods is basic theme of the Monte Carlo methods. The aim of this lecture is to describe the fundamental statistics required to build the foundations of Monte Carlo technique for radiation transport calculation. Lecture note is organized in the following way. Section (1) will describe the introduction of Basic Monte Carlo and its classification towards the respective field. Section (2) will describe the random sampling methods, a key component of Monte Carlo radiation transport calculation, Section (3) will provide the statistical uncertainty of Monte Carlo estimates, Section (4) will describe in brief the importance of variance reduction techniques while sampling particles such as photon, or neutron in the process of radiation transport

  13. CODE's new solar radiation pressure model for GNSS orbit determination

    Science.gov (United States)

    Arnold, D.; Meindl, M.; Beutler, G.; Dach, R.; Schaer, S.; Lutz, S.; Prange, L.; Sośnica, K.; Mervart, L.; Jäggi, A.

    2015-08-01

    The Empirical CODE Orbit Model (ECOM) of the Center for Orbit Determination in Europe (CODE), which was developed in the early 1990s, is widely used in the International GNSS Service (IGS) community. For a rather long time, spurious spectral lines are known to exist in geophysical parameters, in particular in the Earth Rotation Parameters (ERPs) and in the estimated geocenter coordinates, which could recently be attributed to the ECOM. These effects grew creepingly with the increasing influence of the GLONASS system in recent years in the CODE analysis, which is based on a rigorous combination of GPS and GLONASS since May 2003. In a first step we show that the problems associated with the ECOM are to the largest extent caused by the GLONASS, which was reaching full deployment by the end of 2011. GPS-only, GLONASS-only, and combined GPS/GLONASS solutions using the observations in the years 2009-2011 of a global network of 92 combined GPS/GLONASS receivers were analyzed for this purpose. In a second step we review direct solar radiation pressure (SRP) models for GNSS satellites. We demonstrate that only even-order short-period harmonic perturbations acting along the direction Sun-satellite occur for GPS and GLONASS satellites, and only odd-order perturbations acting along the direction perpendicular to both, the vector Sun-satellite and the spacecraft's solar panel axis. Based on this insight we assess in the third step the performance of four candidate orbit models for the future ECOM. The geocenter coordinates, the ERP differences w. r. t. the IERS 08 C04 series of ERPs, the misclosures for the midnight epochs of the daily orbital arcs, and scale parameters of Helmert transformations for station coordinates serve as quality criteria. The old and updated ECOM are validated in addition with satellite laser ranging (SLR) observations and by comparing the orbits to those of the IGS and other analysis centers. Based on all tests, we present a new extended ECOM which

  14. Internal radiation dose calculations with the INREM II computer code

    International Nuclear Information System (INIS)

    Dunning, D.E. Jr.; Killough, G.G.

    1978-01-01

    A computer code, INREM II, was developed to calculate the internal radiation dose equivalent to organs of man which results from the intake of a radionuclide by inhalation or ingestion. Deposition and removal of radioactivity from the respiratory tract is represented by the Internal Commission on Radiological Protection Task Group Lung Model. A four-segment catenary model of the gastrointestinal tract is used to estimate movement of radioactive material that is ingested, or swallowed after being cleared from the respiratory tract. Retention of radioactivity in other organs is specified by linear combinations of decaying exponential functions. The formation and decay of radioactive daughters is treated explicitly, with each radionuclide in the decay chain having its own uptake and retention parameters, as supplied by the user. The dose equivalent to a target organ is computed as the sum of contributions from each source organ in which radioactivity is assumed to be situated. This calculation utilizes a matrix of dosimetric S-factors (rem/μCi-day) supplied by the user for the particular choice of source and target organs. Output permits the evaluation of components of dose from cross-irradiations when penetrating radiations are present. INREM II has been utilized with current radioactive decay data and metabolic models to produce extensive tabulations of dose conversion factors for a reference adult for approximately 150 radionuclides of interest in environmental assessments of light-water-reactor fuel cycles. These dose conversion factors represent the 50-year dose commitment per microcurie intake of a given radionuclide for 22target organs including contributions from specified source organs and surplus activity in the rest of the body. These tabulations are particularly significant in their consistent use of contemporary models and data and in the detail of documentation

  15. Renormalization-group approach to nonlinear radiation-transport problems

    International Nuclear Information System (INIS)

    Chapline, G.F.

    1980-01-01

    A Monte Carlo method is derived for solving nonlinear radiation-transport problems that allows one to average over the effects of many photon absorptions and emissions at frequencies where the opacity is large. This method should allow one to treat radiation-transport problems with large optical depths, e.g., line-transport problems, with little increase in computational effort over that which is required for optically thin problems

  16. HELIOS-CR - A 1-D radiation-magnetohydrodynamics code with inline atomic kinetics modeling

    International Nuclear Information System (INIS)

    MacFarlane, J.J.; Golovkin, I.E.; Woodruff, P.R.

    2006-01-01

    HELIOS-CR is a user-oriented 1D radiation-magnetohydrodynamics code to simulate the dynamic evolution of laser-produced plasmas and z-pinch plasmas. It includes an in-line collisional-radiative (CR) model for computing non-LTE atomic level populations at each time step of the hydrodynamics simulation. HELIOS-CR has been designed for ease of use, and is well-suited for experimentalists, as well as graduate and undergraduate student researchers. The energy equations employed include models for laser energy deposition, radiation from external sources, and high-current discharges. Radiative transport can be calculated using either a multi-frequency flux-limited diffusion model, or a multi-frequency, multi-angle short characteristics model. HELIOS-CR supports the use of SESAME equation of state (EOS) tables, PROPACEOS EOS/multi-group opacity data tables, and non-LTE plasma properties computed using the inline CR modeling. Time-, space-, and frequency-dependent results from HELIOS-CR calculations are readily displayed with the HydroPLOT graphics tool. In addition, the results of HELIOS simulations can be post-processed using the SPECT3D Imaging and Spectral Analysis Suite to generate images and spectra that can be directly compared with experimental measurements. The HELIOS-CR package runs on Windows, Linux, and Mac OSX platforms, and includes online documentation. We will discuss the major features of HELIOS-CR, and present example results from simulations

  17. Transport of infrared radiation in cuboidal clouds

    Science.gov (United States)

    Harshvardhan, MR.; Weinman, J. A.; Davies, R.

    1981-01-01

    The transport of infrared radiation in a single cuboidal cloud is modeled using a variable azimuth two-stream approximation. Computations are made at 10 microns for a Deirmendjian (1969) C-1 water cloud where the single scattering albedo is equal to 0.638 and the asymmetry parameter is 0.865. The results indicate that the emittance of the top face of the model cloud is always less than that for a plane parallel cloud of the same optical depth. The hemispheric flux escaping from the cloud top possesses a gradient from the center to the edges which are warmer when the cloud is over warmer ground. Cooling rate calculations in the 8-13.6 micron region demonstrate that there is cooling out of the sides of the cloud at all levels even when there is heating of the core from the ground below. The radiances exiting from model cuboidal clouds are computed by path integration over the source function obtained with the two-stream approximation. Results indicate that the brightness temperature measured from finite clouds will overestimate the cloud-top temperature.

  18. 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

  19. Aerosol sampling and Transport Efficiency Calculation (ASTEC) and application to surtsey/DCH aerosol sampling system: Code version 1.0: Code description and user's manual

    International Nuclear Information System (INIS)

    Yamano, N.; Brockmann, J.E.

    1989-05-01

    This report describes the features and use of the Aerosol Sampling and Transport Efficiency Calculation (ASTEC) Code. The ASTEC code has been developed to assess aerosol transport efficiency source term experiments at Sandia National Laboratories. This code also has broad application for aerosol sampling and transport efficiency calculations in general as well as for aerosol transport considerations in nuclear reactor safety issues. 32 refs., 31 figs., 7 tabs

  20. BOT3P: a mesh generation software package for the transport analysis codes Dort, Tort, Twodant, Threedant and MCNP

    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)

  1. The adaptive collision source method for discrete ordinates radiation transport

    International Nuclear Information System (INIS)

    Walters, William J.; Haghighat, Alireza

    2017-01-01

    Highlights: • A new adaptive quadrature method to solve the discrete ordinates transport equation. • The adaptive collision source (ACS) method splits the flux into n’th collided components. • Uncollided flux requires high quadrature; this is lowered with number of collisions. • ACS automatically applies appropriate quadrature order each collided component. • The adaptive quadrature is 1.5–4 times more efficient than uniform quadrature. - Abstract: A novel collision source method has been developed to solve the Linear Boltzmann Equation (LBE) more efficiently by adaptation of the angular quadrature order. The angular adaptation method is unique in that the flux from each scattering source iteration is obtained, with potentially a different quadrature order used for each. Traditionally, the flux from every iteration is combined, with the same quadrature applied to the combined flux. Since the scattering process tends to distribute the radiation more evenly over angles (i.e., make it more isotropic), the quadrature requirements generally decrease with each iteration. This method allows for an optimal use of processing power, by using a high order quadrature for the first iterations that need it, before shifting to lower order quadratures for the remaining iterations. This is essentially an extension of the first collision source method, and is referred to as the adaptive collision source (ACS) method. The ACS methodology has been implemented in the 3-D, parallel, multigroup discrete ordinates code TITAN. This code was tested on a several simple and complex fixed-source problems. The ACS implementation in TITAN has shown a reduction in computation time by a factor of 1.5–4 on the fixed-source test problems, for the same desired level of accuracy, as compared to the standard TITAN code.

  2. Regional Atmospheric Transport Code for Hanford Emission Tracking, Version 2 (RATCHET2)

    International Nuclear Information System (INIS)

    Ramsdell, James V.; Rishel, Jeremy P.

    2006-01-01

    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.

  3. 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.

  4. Implementation and testing of a multivariate inverse radiation transport solver

    International Nuclear Information System (INIS)

    Mattingly, John; Mitchell, Dean J.

    2012-01-01

    Detection, identification, and characterization of special nuclear materials (SNM) all face the same basic challenge: to varying degrees, each must infer the presence, composition, and configuration of the SNM by analyzing a set of measured radiation signatures. Solutions to this problem implement inverse radiation transport methods. Given a set of measured radiation signatures, inverse radiation transport estimates properties of the source terms and transport media that are consistent with those signatures. This paper describes one implementation of a multivariate inverse radiation transport solver. The solver simultaneously analyzes gamma spectrometry and neutron multiplicity measurements to fit a one-dimensional radiation transport model with variable layer thicknesses using nonlinear regression. The solver's essential components are described, and its performance is illustrated by application to benchmark experiments conducted with plutonium metal. - Highlights: ► Inverse problems, specifically applied to identifying and characterizing radiation sources . ► Radiation transport. ► Analysis of gamma spectroscopy and neutron multiplicity counting measurements. ► Experimental testing of the inverse solver against measurements of plutonium.

  5. HZETRN radiation transport validation using balloon-based experimental data

    Science.gov (United States)

    Warner, James E.; Norman, Ryan B.; Blattnig, Steve R.

    2018-05-01

    The deterministic radiation transport code HZETRN (High charge (Z) and Energy TRaNsport) was developed by NASA to study the effects of cosmic radiation on astronauts and instrumentation shielded by various materials. This work presents an analysis of computed differential flux from HZETRN compared with measurement data from three balloon-based experiments over a range of atmospheric depths, particle types, and energies. Model uncertainties were quantified using an interval-based validation metric that takes into account measurement uncertainty both in the flux and the energy at which it was measured. Average uncertainty metrics were computed for the entire dataset as well as subsets of the measurements (by experiment, particle type, energy, etc.) to reveal any specific trends of systematic over- or under-prediction by HZETRN. The distribution of individual model uncertainties was also investigated to study the range and dispersion of errors beyond just single scalar and interval metrics. The differential fluxes from HZETRN were generally well-correlated with balloon-based measurements; the median relative model difference across the entire dataset was determined to be 30%. The distribution of model uncertainties, however, revealed that the range of errors was relatively broad, with approximately 30% of the uncertainties exceeding ± 40%. The distribution also indicated that HZETRN systematically under-predicts the measurement dataset as a whole, with approximately 80% of the relative uncertainties having negative values. Instances of systematic bias for subsets of the data were also observed, including a significant underestimation of alpha particles and protons for energies below 2.5 GeV/u. Muons were found to be systematically over-predicted at atmospheric depths deeper than 50 g/cm2 but under-predicted for shallower depths. Furthermore, a systematic under-prediction of alpha particles and protons was observed below the geomagnetic cutoff, suggesting that

  6. Highly radiative plasmas for local transport studies and power and particle handling in reactor regimes

    International Nuclear Information System (INIS)

    Hill, K.W.; Bell, M.G.; Budny, R.

    1999-01-01

    To study the applicability of artificially enhanced impurity radiation for mitigation of the plasma-limiter interaction in reactor regimes, krypton and xenon gases were injected into TFTR supershots and high-l i plasmas. At neutral beam injection (NBI) powers P B ≥ 30 MW, carbon influxes (blooms) were suppressed, leading to improved energy confinement and neutron production in both D and DT plasmas, and the highest DT fusion energy production (7.6 MJ) in a TFTR pulse. Comparisons of the measured radiated power profiles with predictions of the MIST impurity transport code have guided studies of highly-radiative plasmas in ITER. The response of the electron and ion temperatures to greatly increased radiative losses from the electrons was used to study thermal transport mechanisms. (author)

  7. Highly radiative plasmas for local transport studies and power and particle handling in reactor regimes

    International Nuclear Information System (INIS)

    Hill, K.W.; Bell, M.G.; Budny, R.

    2001-01-01

    To study the applicability of artificially enhanced impurity radiation for mitigation of the plasma-limiter interaction in reactor regimes, krypton and xenon gases were injected into TFTR supershots and high-l i plasmas. At neutral beam injection (NBI) powers P B ≤30MW, carbon influxes (blooms) were suppressed, leading to improved energy confinement and neutron production in both D and DT plasmas, and the highest DT fusion energy production (7.6 MJ) in a TFTR pulse. Comparisons of the measured radiated power profiles with predictions of the MIST impurity transport code have guided studies of highly-radiative plasmas in ITER. The response of the electron and ion temperatures to greatly increased radiative losses from the electrons was used to study thermal transport mechanisms. (author)

  8. Highly Radiative Plasmas for Local Transport Studies and Power and Particle Handling in Reactor Regimes

    International Nuclear Information System (INIS)

    Bell, M.G.; Bell, R.E.; Budny, R.; Bush, C.E.; Hill, K.W.

    1998-01-01

    To study the applicability of artificially enhanced impurity radiation for mitigation of the plasma-limiter interaction in reactor regimes, krypton and xenon gases were injected into the Tokamak Fusion Test Reactor (TFTR) supershots and high-l(subscript) plasmas. At neutral beam injection (NBI) powers P(subscript B) greater than or equal to 30 MW, carbon influxes (blooms) were suppressed, leading to improved energy confinement and neutron production in both deuterium (D) and deuterium-tritium (DT) plasmas, and the highest DT fusion energy production (7.6 MJ) in a TFTR pulse. Comparisons of the measured radiated power profiles with predictions of the MIST impurity transport code have guided studies of highly-radiative plasmas in the International Thermonuclear Experimental Reactor (ITER). The response of the electron and ion temperatures to greatly increased radiative losses from the electrons was used to study thermal transport mechanisms

  9. Recommendations for computer code selection of a flow and transport code to be used in undisturbed vadose zone calculations for TWRS immobilized wastes environmental analyses

    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

  10. 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)

  11. Verification and Validation of the Tritium Transport Code TMAP7

    International Nuclear Information System (INIS)

    Longhurst, Glen R.; Ambrosek, James

    2005-01-01

    The TMAP code has been upgraded to version 7, which includes radioactive decay along with many features implemented in prior versions. Pursuant to acceptance and release for distribution, the code was exercised in a variety of problem types to demonstrate that it provides results in agreement with theoretical results for cases where those are available. It has also been used to model certain experimental results. In this paper, the capabilities of the TMAP7 code are demonstrated by presenting some of the results from the verification and validation process

  12. A Photon Free Method to Solve Radiation Transport Equations

    International Nuclear Information System (INIS)

    Chang, B

    2006-01-01

    The multi-group discrete-ordinate equations of radiation transfer is solved for the first time by Newton's method. It is a photon free method because the photon variables are eliminated from the radiation equations to yield a N group XN direction smaller but equivalent system of equations. The smaller set of equations can be solved more efficiently than the original set of equations. Newton's method is more stable than the Semi-implicit Linear method currently used by conventional radiation codes

  13. Code of practice for the safe use of ionizing radiation in secondary schools (1986)

    International Nuclear Information System (INIS)

    1987-01-01

    The code of practice is intended for schools and indicates the basic philosophy behind the current approach to the control of hazards associated with the use of ionizing radiation. The purpose of this code is to provide guidance on safe and proper practices in the use of radiation. It covers modes of radiation exposure, shielding, dose limits, responsibility, general rules, x-ray generators, general control of radioactive sources, sealed sources and unsealed sources

  14. Review of the Monte Carlo and deterministic codes in radiation protection and dosimetry

    Energy Technology Data Exchange (ETDEWEB)

    Tagziria, H

    2000-02-01

    Modelling a physical system can be carried out either stochastically or deterministically. An example of the former method is the Monte Carlo technique, in which statistically approximate methods are applied to exact models. No transport equation is solved as individual particles are simulated and some specific aspect (tally) of their average behaviour is recorded. The average behaviour of the physical system is then inferred using the central limit theorem. In contrast, deterministic codes use mathematically exact methods that are applied to approximate models to solve the transport equation for the average particle behaviour. The physical system is subdivided in boxes in the phase-space system and particles are followed from one box to the next. The smaller the boxes the better the approximations become. Although the Monte Carlo method has been used for centuries, its more recent manifestation has really emerged from the Manhattan project of the Word War II. Its invention is thought to be mainly due to Metropolis, Ulah (through his interest in poker), Fermi, von Neuman andRichtmeyer. Over the last 20 years or so, the Monte Carlo technique has become a powerful tool in radiation transport. This is due to users taking full advantage of richer cross section data, more powerful computers and Monte Carlo techniques for radiation transport, with high quality physics and better known source spectra. This method is a common sense approach to radiation transport and its success and popularity is quite often also due to necessity, because measurements are not always possible or affordable. In the Monte Carlo method, which is inherently realistic because nature is statistical, a more detailed physics is made possible by isolation of events while rather elaborate geometries can be modelled. Provided that the physics is correct, a simulation is exactly analogous to an experimenter counting particles. In contrast to the deterministic approach, however, a disadvantage of the

  15. Integrated transport code system for a multicomponent plasma in a gas dynamic trap

    International Nuclear Information System (INIS)

    Anikeev, A.V.; Karpushov, A.N.; Noak, K.; Strogalova, S.L.

    2000-01-01

    This report is focused on the development of the theoretical and numerical models of multicomponent high-β plasma confinement and transport in the gas-dynamic trap (GDT). In order to simulate the plasma behavior in the GDT as well as that in the GDT-based neutron source the Integrated Transport Code System is developed from existing stand-alone codes calculating the target plasma, the fast ions and the neutral gas in the GDT. The code system considers the full dependence of the transport phenomena on space, time, energy and angle variables as well as the interactions between the particle fields [ru

  16. Radiation doses from the transport of radioactive materials

    International Nuclear Information System (INIS)

    Shaw, K.B.; Holyoak, B.

    1983-01-01

    A summary is given of a study on radiation exposure resulting from the transport of radioactive materials within the United Kingdom. It was concluded that the transport of technetium generators for hospital use accounts for about 49% of the occupational exposure for the normal transport of radioactive materials. Other isotopes for medical and industrial use contribute about 38% of the occupational exposure and the remainder can be attributed to transportation as a result of the nuclear fuel cycle including the transport of irradiated nuclear fuel. The occupational collective dose for all modes of transport is estimated at 1 man Sv y -1 . (UK)

  17. Ethical codes. Fig leaf argument, ballast or cultural element for radiation protection?; Ethik-Codes. Feigenblatt, Ballast oder Kulturelement fuer den Strahlenschutz?

    Energy Technology Data Exchange (ETDEWEB)

    Gellermann, Rainer [Nuclear Control and Consulting GmbH, Braunschweig (Germany)

    2014-07-01

    The international association for radiation protection (IRPA) adopted in May 2004 a Code of Ethics in order to allow their members to hold an adequate professional level of ethical line of action. Based on this code of ethics the professional body of radiation protection (Fachverband fuer Strahlenschutz) has developed its own ethical code and adopted in 2005.

  18. The Monte Carlo photoionization and moving-mesh radiation hydrodynamics code CMACIONIZE

    Science.gov (United States)

    Vandenbroucke, B.; Wood, K.

    2018-04-01

    We present the public Monte Carlo photoionization and moving-mesh radiation hydrodynamics code CMACIONIZE, which can be used to simulate the self-consistent evolution of HII regions surrounding young O and B stars, or other sources of ionizing radiation. The code combines a Monte Carlo photoionization algorithm that uses a complex mix of hydrogen, helium and several coolants in order to self-consistently solve for the ionization and temperature balance at any given type, with a standard first order hydrodynamics scheme. The code can be run as a post-processing tool to get the line emission from an existing simulation snapshot, but can also be used to run full radiation hydrodynamical simulations. Both the radiation transfer and the hydrodynamics are implemented in a general way that is independent of the grid structure that is used to discretize the system, allowing it to be run both as a standard fixed grid code, but also as a moving-mesh code.

  19. Radiation safety in sea transport of radioactive material in Japan

    International Nuclear Information System (INIS)

    Odano, N.; Yanagi, H.

    2004-01-01

    Radiation safety for sea transport of radioactive material in Japan has been discussed based on records of the exposed dose of sea transport workers and measured data of dose rate equivalents distribution inboard exclusive radioactive material shipping vessels. Recent surveyed records of the exposed doses of workers who engaged in sea transport operation indicate that exposed doses of transport workers are significantly low. Measured distribution of the exposed dose equivalents inboard those vessels indicates that dose rate equivalents inside those vessels are lower than levels regulated by the transport regulations of Japan. These facts clarify that radiation safety of inboard environment and handling of transport casks in sea transport of radioactive material in Japan are assured

  20. Radiation safety in sea transport of radioactive material in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Odano, N. [National Maritime Research Inst., Tokyo (Japan); Yanagi, H. [Nuclear Fuel Transport Co., Ltd., Tokyo (Japan)

    2004-07-01

    Radiation safety for sea transport of radioactive material in Japan has been discussed based on records of the exposed dose of sea transport workers and measured data of dose rate equivalents distribution inboard exclusive radioactive material shipping vessels. Recent surveyed records of the exposed doses of workers who engaged in sea transport operation indicate that exposed doses of transport workers are significantly low. Measured distribution of the exposed dose equivalents inboard those vessels indicates that dose rate equivalents inside those vessels are lower than levels regulated by the transport regulations of Japan. These facts clarify that radiation safety of inboard environment and handling of transport casks in sea transport of radioactive material in Japan are assured.

  1. The one-dimensional transport codes MAKOKOT. Presentation and directions for use

    International Nuclear Information System (INIS)

    Capes, H.; Mercier, C.; Morera, J.P.

    1986-06-01

    In this note are presented the different one-dimensional evolution codes available to date under the generic name MAKOKOT. They are six principal codes: - TRANS: for ion and electron transport; -NEUTRE: for neutrals; -IMPUR: for impurities; -ECRH: for electron cyclotron resonance; -DENT: for sawtooth modelling and analysis; -BILAN: for global verification of conservation. One supplementary code is added which is an impurity evolution code; it takes in account, in 1-D geometry, the buffer zone generated by the limiter between the hot plasma and the wall. An abundant bibliography is given. A comprehensive manner of using is given which underlines the use versatility of these codes [fr

  2. Methodology for a thermal analysis of a proposed SFR transport cask with the thermal code SYRTHES

    International Nuclear Information System (INIS)

    Peniguel, C.; Rupp, I.; Schneider, J. P.

    2010-01-01

    Fast reactors with liquid metal coolant have received a renewed interest owing to the need of a more efficient usage of the primary uranium resources, and they are one of the proposal for the next Generation IV. In the framework of the 2006 French law on sustainable management of radioactive materials and waste, an evaluation of the industrial perspectives of minor actinides transmutation advantages and drawbacks in Generation IV fast spectrum reactors system is requested for 2012. The CEA is in charge of studying the global problem, but on some aspects, EDF is interested to do its own exploratory studies. Among other points, transport is seen as important for the nuclear industry, to link points of production and treatment. Nuclear fuel is generally transported in thick walled rail or truck casks. These packages are designed to provide confinement, shielding and criticality protection during normal and severe transport conditions. Heat generated within the fuel (and a contribution of solar heating) makes the package becoming quite hot, but one must demonstrate that the cladding temperature does not exceed a long term temperature limit during normal transport. This paper presents a thermal study done on a package in which 9 SFR assemblies are included. Each of them is of hexagonal shape and contains 271 fuel pins. The approach followed for these calculations is to rely on an explicit representation of all pins. For these calculations a 2D analysis is performed thanks to the thermal code SYRTHES. Conduction is solved thanks to a finite element method, while thermal radiation is handled through a radiosity approach. The main aim of this paper is to present a possible numerical methodology to handle the thermal problem. (authors)

  3. Monte Carlo 2000 Conference : Advanced Monte Carlo for Radiation Physics, Particle Transport Simulation and Applications

    CERN Document Server

    Baräo, Fernando; Nakagawa, Masayuki; Távora, Luis; Vaz, Pedro

    2001-01-01

    This book focusses on the state of the art of Monte Carlo methods in radiation physics and particle transport simulation and applications, the latter involving in particular, the use and development of electron--gamma, neutron--gamma and hadronic codes. Besides the basic theory and the methods employed, special attention is paid to algorithm development for modeling, and the analysis of experiments and measurements in a variety of fields ranging from particle to medical physics.

  4. 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)

  5. Study on MPI/OpenMP hybrid parallelism for Monte Carlo neutron transport code

    International Nuclear Information System (INIS)

    Liang Jingang; Xu Qi; Wang Kan; Liu Shiwen

    2013-01-01

    Parallel programming with mixed mode of messages-passing and shared-memory has several advantages when used in Monte Carlo neutron transport code, such as fitting hardware of distributed-shared clusters, economizing memory demand of Monte Carlo transport, improving parallel performance, and so on. MPI/OpenMP hybrid parallelism was implemented based on a one dimension Monte Carlo neutron transport code. Some critical factors affecting the parallel performance were analyzed and solutions were proposed for several problems such as contention access, lock contention and false sharing. After optimization the code was tested finally. It is shown that the hybrid parallel code can reach good performance just as pure MPI parallel program, while it saves a lot of memory usage at the same time. Therefore hybrid parallel is efficient for achieving large-scale parallel of Monte Carlo neutron transport. (authors)

  6. 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

  7. Intense radiative heat transport across a nano-scale gap

    International Nuclear Information System (INIS)

    Budaev, Bair V.; Ghafari, Amin; Bogy, David B.

    2016-01-01

    In this paper, we analyze the radiative heat transport in layered structures. The analysis is based on our prior description of the spectrum of thermally excited waves in systems with a heat flux. The developed method correctly predicts results for all known special cases for both large and closing gaps. Numerical examples demonstrate the applicability of our approach to the calculation of the radiative heat transport coefficient across various layered structures.

  8. THREEDANT: A code to perform three-dimensional, neutral particle transport calculations

    International Nuclear Information System (INIS)

    Alcouffe, R.E.

    1994-01-01

    The THREEDANT code solves the three-dimensional neutral particle transport equation in its first order, multigroup, discrate ordinate form. The code allows an unlimited number of groups (depending upon the cross section set), angular quadrature up to S-100, and unlimited Pn order again depending upon the cross section set. The code has three options for spatial differencing, diamond with set-to-zero fixup, adaptive weighted diamond, and linear modal. The geometry options are XYZ and RZΘ with a special XYZ option based upon a volume fraction method. This allows objects or bodies of any shape to be modelled as input which gives the code as much geometric description flexibility as the Monte Carlo code MCNP. The transport equation is solved by source iteration accelerated by the DSA method. Both inner and outer iterations are so accelerated. Some results are presented which demonstrate the effectiveness of these techniques. The code is available on several types of computing platforms

  9. An investigation of fission models for high-energy radiation transport calculations

    International Nuclear Information System (INIS)

    Armstrong, T.W.; Cloth, P.; Filges, D.; Neef, R.D.

    1983-07-01

    An investigation of high-energy fission models for use in the HETC code has been made. The validation work has been directed checking the accuracy of the high-energy radiation transport computer code HETC to investigate the appropriate model for routine calculations, particularly for spallation neutron source applications. Model calculations are given in terms of neutron production, fission fragment energy release, and residual nuclei production for high-energy protons incident on thin uranium targets. The effect of the fission models on neutron production from thick uranium targets is also shown. (orig.)

  10. On the derivation of vector radiative transfer equation for polarized radiative transport in graded index media

    International Nuclear Information System (INIS)

    Zhao, J.M.; Tan, J.Y.; Liu, L.H.

    2012-01-01

    Light transport in graded index media follows a curved trajectory determined by Fermat's principle. Besides the effect of variation of the refractive index on the transport of radiative intensity, the curved ray trajectory will induce geometrical effects on the transport of polarization ellipse. This paper presents a complete derivation of vector radiative transfer equation for polarized radiation transport in absorption, emission and scattering graded index media. The derivation is based on the analysis of the conserved quantities for polarized light transport along curved trajectory and a novel approach. The obtained transfer equation can be considered as a generalization of the classic vector radiative transfer equation that is only valid for uniform refractive index media. Several variant forms of the transport equation are also presented, which include the form for Stokes parameters defined with a fixed reference and the Eulerian forms in the ray coordinate and in several common orthogonal coordinate systems.

  11. MCNP: a general Monte Carlo code for neutron and photon transport

    International Nuclear Information System (INIS)

    1978-07-01

    The general-purpose Monte Carlo code MCNP can be used for neutron, photon, or coupled neutron--photon transport. 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 (such as ENDF/B-IV) are accounted for. 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. Standard optional variance reduction schemes include geometry splitting and Russian roulette, the exponential transformation, energy splitting, forced collisions in designated cells, flux estimates at point detectors, track-length estimators, and source biasing. The standard output of MCNP includes two-way current as a function of energy, time, and angle with the normal, across any subset of bounding surfaces in the problem. Fluxes across any set of bounding surfaces are available as a function of time and energy. Similarly, the flux at designated points and the average flux in a cell (track length per unit volume) are standard tallies. Reactions such as fissions or absorptions may be obtained in a subset of geometric cells. The heating tallies give the energy deposition per starting particle. In addition, particles may be flagged when they cross specified surfaces or enter designated cells, and the contributions of these flagged particles to certain of the tallies are listed separately. All quantities printed out have their relative errors listed also. 11 figures, 27 tables

  12. General-purpose Monte Carlo codes for neutron and photon transport calculations. MVP version 3

    International Nuclear Information System (INIS)

    Nagaya, Yasunobu

    2017-01-01

    JAEA has developed a general-purpose neutron/photon transport Monte Carlo code MVP. This paper describes the recent development of the MVP code and reviews the basic features and capabilities. In addition, capabilities implemented in Version 3 are also described. (author)

  13. Implementation of the kinetics in the transport code AZTRAN; Implementacion de la cinetica en el codigo de transporte AZTRAN

    Energy Technology Data Exchange (ETDEWEB)

    Duran G, J. A.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, San Pedro Zacatenco, 07738 Ciudad de Mexico (Mexico); Gomez T, A. M., E-mail: redfield1290@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2017-09-15

    This paper shows the implementation of the time dependence in the three-dimensional transport code AZTRAN (AZtlan TRANsport), which belongs to the AZTLAN platform, for the analysis of nuclear reactors (currently under development). The AZTRAN code with this implementation is able to numerically solve the time-dependent transport equation in XYZ geometry, for several energy groups, using the discrete ordinate method S{sub n} for the discretization of the angular variable, the nodal method RTN-0 for spatial discretization and method 0 for discretization in time. Initially, the code only solved the neutrons transport equation in steady state, so the implementation of the temporal part was made integrating the neutrons transport equation with respect to time and balance equations corresponding to the concentrations of delayed neutron precursors, for which method 0 was applied. After having directly implemented code kinetics, the improved quasi-static method was implemented, which is a tool for reducing computation time, where the angular flow is factored by the product of two functions called shape function and amplitude function, where the first is calculated for long time steps, called macro-steps and the second is resolved for small time steps called micro-steps. In the new version of AZTRAN several Benchmark problems that were taken from the literature were simulated, the problems used are of two and three dimensions which allowed corroborating the accuracy and stability of the code, showing in general in the reference tests a good behavior. (Author)

  14. Simplified calculation method for radiation dose under normal condition of transport

    International Nuclear Information System (INIS)

    Watabe, N.; Ozaki, S.; Sato, K.; Sugahara, A.

    1993-01-01

    In order to estimate radiation dose during transportation of radioactive materials, the following computer codes are available: RADTRAN, INTERTRAN, J-TRAN. Because these codes consist of functions for estimating doses not only under normal conditions but also in the case of accidents, when nuclei may leak and spread into the environment by air diffusion, the user needs to have special knowledge and experience. In this presentation, we describe how, with a view to preparing a method by which a person in charge of transportation can calculate doses in normal conditions, the main parameters upon which the value of doses depends were extracted and the dose for a unit of transportation was estimated. (J.P.N.)

  15. A simplified computer code based on point Kernel theory for calculating radiation dose in packages of radioactive material

    International Nuclear Information System (INIS)

    1986-03-01

    A study on radiation dose control in packages of radioactive waste from nuclear facilities, hospitals and industries, such as sources of Ra-226, Co-60, Ir-192 and Cs-137, is presented. The MAPA and MAPAM computer codes, based on point Kernel theory for calculating doses of several source-shielding type configurations, aiming to assure the safe transport conditions for these sources, was developed. The validation of the code for point sources, using the values provided by NCRP, for the thickness of lead and concrete shieldings, limiting the dose at 100 Mrem/hr for several distances from the source to the detector, was carried out. The validation for non point sources was carried out, measuring experimentally radiation dose from packages developed by Brazilian CNEN/S.P. for removing the sources. (M.C.K.) [pt

  16. Nonrelativistic grey Sn-transport radiative-shock solutions

    International Nuclear Information System (INIS)

    Ferguson, J. M.; Morel, J. E.; Lowrie, R. B.

    2017-01-01

    We present semi-analytic radiative-shock solutions in which grey Sn-transport is used to model the radiation, and we include both constant cross sections and cross sections that depend on temperature and density. These new solutions solve for a variable Eddington factor (VEF) across the shock domain, which allows for interesting physics not seen before in radiative-shock solutions. Comparisons are made with the grey nonequilibrium-diffusion radiative-shock solutions of Lowrie and Edwards [1], which assumed that the Eddington factor is constant across the shock domain. It is our experience that the local Mach number is monotonic when producing nonequilibrium-diffusion solutions, but that this monotonicity may disappear while integrating the precursor region to produce Sn-transport solutions. For temperature- and density-dependent cross sections we show evidence of a spike in the VEF in the far upstream portion of the radiative-shock precursor. We show evidence of an adaptation zone in the precursor region, adjacent to the embedded hydrodynamic shock, as conjectured by Drake [2, 3], and also confirm his expectation that the precursor temperatures adjacent to the Zel’dovich spike take values that are greater than the downstream post-shock equilibrium temperature. We also show evidence that the radiation energy density can be nonmonotonic under the Zel’dovich spike, which is indicative of anti-diffusive radiation flow as predicted by McClarren and Drake [4]. We compare the angle dependence of the radiation flow for the Sn-transport and nonequilibriumdiffusion radiation solutions, and show that there are considerable differences in the radiation flow between these models across the shock structure. Lastly, we analyze the radiation flow to understand the cause of the adaptation zone, as well as the structure of the Sn-transport radiation-intensity solutions across the shock structure.

  17. Coherent Synchrotron Radiation A Simulation Code Based on the Non-Linear Extension of the Operator Splitting Method

    CERN Document Server

    Dattoli, Giuseppe

    2005-01-01

    The coherent synchrotron radiation (CSR) is one of the main problems limiting the performance of high intensity electron accelerators. A code devoted to the analysis of this type of problems should be fast and reliable: conditions that are usually hardly achieved at the same time. In the past, codes based on Lie algebraic techniques have been very efficient to treat transport problem in accelerators. The extension of these method to the non-linear case is ideally suited to treat CSR instability problems. We report on the development of a numerical code, based on the solution of the Vlasov equation, with the inclusion of non-linear contribution due to wake field effects. The proposed solution method exploits an algebraic technique, using exponential operators implemented numerically in C++. We show that the integration procedure is capable of reproducing the onset of an instability and effects associated with bunching mechanisms leading to the growth of the instability itself. In addition, parametric studies a...

  18. Transport simulations of a density limit in radiation-dominated tokamak discharges: profile effects

    International Nuclear Information System (INIS)

    Stotler, D.P.

    1988-01-01

    The density limit observed in tokamak experiments is thought to be due to a radiative collapse of the current channel. A transport code coupled with a magnetohydrodynamic (MHD) equilibrium routine is used to determine the detailed, self-consistent evolution of the plasma profiles in tokamak discharges with radiated power close to or equaling the input power. The present work is confined to Ohmic discharges in steady state. It is found that the shape of the density profile can have a significant impact on the variation of the maximum electron density with plasma current. Analytic calculations confirm this result

  19. Transport simulations of a density limit in radiation-dominated tokamak discharges: Profile effects

    International Nuclear Information System (INIS)

    Stotler, D.P.

    1988-06-01

    The density limit observed in tokamak experiments is thought to be due to a radiative collapse of the current channel. A transport code coupled with an MHD equilibrium routine is used to determine the detailed, self-consistent evolution of the plasma profiles in tokamak discharges with radiated power close to or equalling the input power. The present work is confined to ohmic discharges in steady state. It is found that the shape of the density profile can have a significant impact on the variation of the maximum electron density with plasma current. Analytic calculations confirm this result. 41 refs., 9 figs

  20. 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

  1. 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.

  2. Asymptotic diffusion limit of cell temperature discretisation schemes for thermal radiation transport

    International Nuclear Information System (INIS)

    Smedley-Stevenson, Richard P.; McClarren, Ryan G.

    2015-01-01

    This paper attempts to unify the asymptotic diffusion limit analysis of thermal radiation transport schemes, for a linear-discontinuous representation of the material temperature reconstructed from cell centred temperature unknowns, in a process known as ‘source tilting’. The asymptotic limits of both Monte Carlo (continuous in space) and deterministic approaches (based on linear-discontinuous finite elements) for solving the transport equation are investigated in slab geometry. The resulting discrete diffusion equations are found to have nonphysical terms that are proportional to any cell-edge discontinuity in the temperature representation. Based on this analysis it is possible to design accurate schemes for representing the material temperature, for coupling thermal radiation transport codes to a cell centred representation of internal energy favoured by ALE (arbitrary Lagrange–Eulerian) hydrodynamics schemes

  3. Asymptotic diffusion limit of cell temperature discretisation schemes for thermal radiation transport

    Energy Technology Data Exchange (ETDEWEB)

    Smedley-Stevenson, Richard P., E-mail: richard.smedley-stevenson@awe.co.uk [AWE PLC, Aldermaston, Reading, Berkshire, RG7 4PR (United Kingdom); Department of Earth Science and Engineering, Imperial College London, SW7 2AZ (United Kingdom); McClarren, Ryan G., E-mail: rmcclarren@ne.tamu.edu [Department of Nuclear Engineering, Texas A & M University, College Station, TX 77843-3133 (United States)

    2015-04-01

    This paper attempts to unify the asymptotic diffusion limit analysis of thermal radiation transport schemes, for a linear-discontinuous representation of the material temperature reconstructed from cell centred temperature unknowns, in a process known as ‘source tilting’. The asymptotic limits of both Monte Carlo (continuous in space) and deterministic approaches (based on linear-discontinuous finite elements) for solving the transport equation are investigated in slab geometry. The resulting discrete diffusion equations are found to have nonphysical terms that are proportional to any cell-edge discontinuity in the temperature representation. Based on this analysis it is possible to design accurate schemes for representing the material temperature, for coupling thermal radiation transport codes to a cell centred representation of internal energy favoured by ALE (arbitrary Lagrange–Eulerian) hydrodynamics schemes.

  4. CFRX, a one-and-a-quarter-dimensional transport code for field-reversed configuration studies

    International Nuclear Information System (INIS)

    Hsiao Mingyuan

    1989-01-01

    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. A typical example of the code results is also given. (orig.)

  5. HETFIS: High-Energy Nucleon-Meson Transport Code with Fission

    International Nuclear Information System (INIS)

    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

  6. MIGFRAC - a code for modelling of radionuclide transport in fracture media

    International Nuclear Information System (INIS)

    Satyanarayana, S.V.M.; Mohankumar, N.; Sasidhar, P.

    2002-05-01

    Radionuclides migrate through diffusion process from radioactive waste disposal facilities into fractures present in the host rock. The transport phenomenon is aided by the circulating ground waters. To model the transport of radionuclides in the charnockite rock formations present at Kalpakkam, a numerical code - MIGFRAC has been developed at SHINE Group, IGCAR. The code has been subjected to rigorous tests and the results of the build up of radionuclide concentrations are validated with a test case up to a distance of 100 meter along the fracture. The report discusses the model, code features and the results obtained up to a distance of 400 meter are presented. (author)

  7. A one-dimensional transport code for the simulation of D-T burning tokamak plasma

    International Nuclear Information System (INIS)

    Tone, Tatsuzo; Maki, Koichi; Kasai, Masao; Nishida, Hidetsugu

    1980-11-01

    A one-dimensional transport code for D-T burning tokamak plasma has been developed, which simulates the spatial behavior of fuel ions(D, T), alpha particles, impurities, temperatures of ions and electrons, plasma current, neutrals, heating of alpha and injected beam particles. The basic transport equations are represented by one generalized equation so that the improvement of models and the addition of new equations may be easily made. A model of burn control using a variable toroidal field ripple is employed. This report describes in detail the simulation model, numerical method and the usage of the code. Some typical examples to which the code has been applied are presented. (author)

  8. Discontinuous Galerkin for the Radiative Transport Equation

    KAUST Repository

    Guermond, Jean-Luc; Kanschat, Guido; Ragusa, Jean C.

    2013-01-01

    This note presents some recent results regarding the approximation of the linear radiative transfer equation using discontinuous Galerkin methods. The locking effect occurring in the diffusion limit with the upwind numerical flux is investigated and a correction technique is proposed.

  9. Discontinuous Galerkin for the Radiative Transport Equation

    KAUST Repository

    Guermond, Jean-Luc

    2013-10-11

    This note presents some recent results regarding the approximation of the linear radiative transfer equation using discontinuous Galerkin methods. The locking effect occurring in the diffusion limit with the upwind numerical flux is investigated and a correction technique is proposed.

  10. Code of practice for conducting radiation work at PINSTECH (revised 1992)

    International Nuclear Information System (INIS)

    Aslam, M.; Atta, M.A.; Orfi, S.D.

    1992-02-01

    The primary objective of this code is to achieve standard of radiation protection and safety set by Pakistan Nuclear Safety and Radiation Protection (PNSRP) ordinance 1984 and PNSRP regulations 1990. Secondary objective remains to make all best efforts to implement latest ICRP recommendations. The revised code of practice sets forth the objective of adequate system radiological safety of radiation workers, environment and general public. The code provides the guidance to persons and authorities who are responsible for the protection of workers and those who are concerned with the planning and management of personnel monitoring services. The procedures set forth in the code are mandatory and in no case should any of them be deviated under normal conditions. All those supervising and performing any kind of radiation work are required to study and adhere to those procedures and shell make all possible efforts to keep the exposure as low as reasonably achievable (ALARA), social and economic factor being taken into account. (author)

  11. Monte Carlo closure for moment-based transport schemes in general relativistic radiation hydrodynamic simulations

    Science.gov (United States)

    Foucart, Francois

    2018-04-01

    General relativistic radiation hydrodynamic simulations are necessary to accurately model a number of astrophysical systems involving black holes and neutron stars. Photon transport plays a crucial role in radiatively dominated accretion discs, while neutrino transport is critical to core-collapse supernovae and to the modelling of electromagnetic transients and nucleosynthesis in neutron star mergers. However, evolving the full Boltzmann equations of radiative transport is extremely expensive. Here, we describe the implementation in the general relativistic SPEC code of a cheaper radiation hydrodynamic method that theoretically converges to a solution of Boltzmann's equation in the limit of infinite numerical resources. The algorithm is based on a grey two-moment scheme, in which we evolve the energy density and momentum density of the radiation. Two-moment schemes require a closure that fills in missing information about the energy spectrum and higher order moments of the radiation. Instead of the approximate analytical closure currently used in core-collapse and merger simulations, we complement the two-moment scheme with a low-accuracy Monte Carlo evolution. The Monte Carlo results can provide any or all of the missing information in the evolution of the moments, as desired by the user. As a first test of our methods, we study a set of idealized problems demonstrating that our algorithm performs significantly better than existing analytical closures. We also discuss the current limitations of our method, in particular open questions regarding the stability of the fully coupled scheme.

  12. 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)

  13. 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.

  14. MATADOR (Methods for the Analysis of Transport And Deposition Of Radionuclides) code description and User's Manual

    International Nuclear Information System (INIS)

    Avci, H.I.; Raghuram, S.; Baybutt, P.

    1985-04-01

    A new computer code called MATADOR (Methods for the Analysis of Transport And Deposition Of Radionuclides) has been developed to replace the CORRAL-2 computer code which was written for the Reactor Safety Study (WASH-1400). This report is a User's Manual for MATADOR. MATADOR is intended for use in system risk studies to analyze radionuclide transport and deposition in reactor containments. The principal output of the code is information on the timing and magnitude of radionuclide releases to the environment as a result of severely degraded core accidents. MATADOR considers the transport of radionuclides through the containment and their removal by natural deposition and by engineered safety systems such as sprays. It is capable of analyzing the behavior of radionuclides existing either as vapors or aerosols in the containment. The code requires input data on the source terms into the containment, the geometry of the containment, and thermal-hydraulic conditions in the containment

  15. A vectorized Monte Carlo code for modeling photon transport in SPECT

    International Nuclear Information System (INIS)

    Smith, M.F.; Floyd, C.E. Jr.; Jaszczak, R.J.

    1993-01-01

    A vectorized Monte Carlo computer code has been developed for modeling photon transport in single photon emission computed tomography (SPECT). The code models photon transport in a uniform attenuating region and photon detection by a gamma camera. It is adapted from a history-based Monte Carlo code in which photon history data are stored in scalar variables and photon histories are computed sequentially. The vectorized code is written in FORTRAN77 and uses an event-based algorithm in which photon history data are stored in arrays and photon history computations are performed within DO loops. The indices of the DO loops range over the number of photon histories, and these loops may take advantage of the vector processing unit of our Stellar GS1000 computer for pipelined computations. Without the use of the vector processor the event-based code is faster than the history-based code because of numerical optimization performed during conversion to the event-based algorithm. When only the detection of unscattered photons is modeled, the event-based code executes 5.1 times faster with the use of the vector processor than without; when the detection of scattered and unscattered photons is modeled the speed increase is a factor of 2.9. Vectorization is a valuable way to increase the performance of Monte Carlo code for modeling photon transport in SPECT

  16. 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)

  17. 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.

  18. 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)

  19. Code of practice on radiation protection in the mining and milling of radioactive ores 1987

    Energy Technology Data Exchange (ETDEWEB)

    1987-01-01

    This Code was formulated under provisions of the Environment Protection (Nuclear Codes) Act 1978 in close consultation with the Governments of the States and the Northern Territory. It is a major revision of the Code of Practice on Radiation Protection in the Mining and Milling of Radioactive Ores (1980), incorporating changes flowing from advances in internationally agreed radiation protection philosophy, and experience gained in Australia in uranium mining and milling operations and the extraction of monazite from mineral sands. The Code specifies the standards, practices, procedures, and measures to prevent or limit risk to employees and to the public from uranium mining and milling, mineral sands operations and extraction of radioactive ores. To assist the industry in meeting the requirements and responsibilities imposed by the Code, guidelines to the former Code will be reviewed and, if appropriate, revised. New guidelines to assist compliance with changed requirements will also be prepared. The Act provides for the revision of codes of practice. The Code of Practice on Radiation Protection in the Mining and Milling of Radioactive Ores (1987) will be reviewed from time to time and revised if necessary to ensure that the highest standards of radiation protection in the mining and milling of radioactive ores are maintained.

  20. Code of practice on radiation protection in the mining and milling of radioactive ores 1987

    International Nuclear Information System (INIS)

    1987-01-01

    This Code was formulated under provisions of the Environment Protection (Nuclear Codes) Act 1978 in close consultation with the Governments of the States and the Northern Territory. It is a major revision of the Code of Practice on Radiation Protection in the Mining and Milling of Radioactive Ores (1980), incorporating changes flowing from advances in internationally agreed radiation protection philosophy, and experience gained in Australia in uranium mining and milling operations and the extraction of monazite from mineral sands. The Code specifies the standards, practices, procedures, and measures to prevent or limit risk to employees and to the public from uranium mining and milling, mineral sands operations and extraction of radioactive ores. To assist the industry in meeting the requirements and responsibilities imposed by the Code, guidelines to the former Code will be reviewed and, if appropriate, revised. New guidelines to assist compliance with changed requirements will also be prepared. The Act provides for the revision of codes of practice. The Code of Practice on Radiation Protection in the Mining and Milling of Radioactive Ores (1987) will be reviewed from time to time and revised if necessary to ensure that the highest standards of radiation protection in the mining and milling of radioactive ores are maintained

  1. 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

  2. Current status of radiation transport tools for proliferation and terrorism prevention

    International Nuclear Information System (INIS)

    Sale, K.E.

    2004-01-01

    Full text: We will present the current status and future plans for the set of calculational tools and databases developed and maintained at LLNL. The calculational tools include the Monte Carlo codes TART 1) and COG 2) as well as the deterministic code ARDRA 3) . In addition to these codes we use currently there is a major development effort for a new massively parallel transport code. An important part of the capability we're developing is a sophisticated user interface, based on a commercial 3-D modeling product, to improve the model development process. A major part of this user interface tool is being developed by Strela 4) under the Nuclear Cities Initiative. Strela has developed a hub-and-spoke technology for code input interconversions (between COG, TART and MCNP) and will produce the plug-ins that extend the capabilities of the 3-D modeler for use as a radiation transport input generator. The major advantages of this approach are the built-in user interface for 3-D modeling and the ability to read a large variety of CAD-file formats. In addition to supporting our current radiation transport codes and developing new capabilities we are working on some nuclear data needs for homeland security. These projects are carried out and the Lawrence Berkeley National Laboratory 88' cyclotron and at the Institute for Nuclear Research of the Nation Academy of Science of Ukraine under and STCU contract. Reference: 1. http://www.llnl.gov/cullen1/mc/htm; 2. http://www-phys.llnl.gov/N_Div/COG/ETR/ETR_9306.html; 3. http://www.llnl.gov/CASC/asciturb/talks/PPT-HTML.131596/tsld030.htm; 4. http://strela.snz.ru/

  3. Atmospheric radiative transfer modeling: a summary of the AER codes

    Energy Technology Data Exchange (ETDEWEB)

    Clough, S.A. [Atmospheric and Environmental Research (AER) Inc., 131 Hartwell Avenue, Lexington, MA 02421-3126 (United States); Shephard, M.W. [Atmospheric and Environmental Research (AER) Inc., 131 Hartwell Avenue, Lexington, MA 02421-3126 (United States)]. E-mail: mshephar@aer.com; Mlawer, E.J. [Atmospheric and Environmental Research (AER) Inc., 131 Hartwell Avenue, Lexington, MA 02421-3126 (United States); Delamere, J.S. [Atmospheric and Environmental Research (AER) Inc., 131 Hartwell Avenue, Lexington, MA 02421-3126 (United States); Iacono, M.J. [Atmospheric and Environmental Research (AER) Inc., 131 Hartwell Avenue, Lexington, MA 02421-3126 (United States); Cady-Pereira, K. [Atmospheric and Environmental Research (AER) Inc., 131 Hartwell Avenue, Lexington, MA 02421-3126 (United States); Boukabara, S. [Atmospheric and Environmental Research (AER) Inc., 131 Hartwell Avenue, Lexington, MA 02421-3126 (United States); Brown, P.D. [Atmospheric and Environmental Research (AER) Inc., 131 Hartwell Avenue, Lexington, MA 02421-3126 (United States)

    2005-03-01

    The radiative transfer models developed at AER are being used extensively for a wide range of applications in the atmospheric sciences. This communication is intended to provide a coherent summary of the various radiative transfer models and associated databases publicly available from AER (http://www.rtweb.aer.com). Among the communities using the models are the remote sensing community (e.g. TES, IASI), the numerical weather prediction community (e.g. ECMWF, NCEP GFS, WRF, MM5), and the climate community (e.g. ECHAM5). Included in this communication is a description of the central features and recent updates for the following models: the line-by-line radiative transfer model (LBLRTM); the line file creation program (LNFL); the longwave and shortwave rapid radiative transfer models, RRTM{sub L}W and RRTM{sub S}W; the Monochromatic Radiative Transfer Model (MonoRTM); the MT{sub C}KD Continuum; and the Kurucz Solar Source Function. LBLRTM and the associated line parameter database (e.g. HITRAN 2000 with 2001 updates) play a central role in the suite of models. The physics adopted for LBLRTM has been extensively analyzed in the context of closure experiments involving the evaluation of the model inputs (e.g. atmospheric state), spectral radiative measurements and the spectral model output. The rapid radiative transfer models are then developed and evaluated using the validated LBLRTM model.

  4. Overview. Department of Environmental and Radiation Transport Physics. Section 6

    Energy Technology Data Exchange (ETDEWEB)

    Loskiewicz, J. [Institute of Nuclear Physics, Cracow (Poland)

    1995-12-31

    Research activities in the Department of Environmental and Radiation Transport Physics are carried out by three Laboratories: Laboratory of Environmental Physics, Laboratory of Neutron Transport Physics and Laboratory of Physics and Modeling of Radiation Transport. The researches provided in 1994 cover: tracer transport and flows in porous media, studies on pollution in atmospheric air, physics of molecular phenomena in chromatographic detectors, studies on neutron transport in heterogenous media, studies on evaluation of neutron cross-section in the thermal region, studies on theory and utilization of neural network in data evaluation, numerical modelling of particle cascades for particle accelerator shielding purpose. In this section the description of mentioned activities as well as the information about personnel employed in the Department, papers and reports published in 1994, contribution to conferences and grants is also given.

  5. Overview. Department of Environmental and Radiation Transport Physics. Section 6

    Energy Technology Data Exchange (ETDEWEB)

    Loskiewicz, J [Institute of Nuclear Physics, Cracow (Poland)

    1996-12-31

    Research activities in the Department of Environmental and Radiation Transport Physics are carried out by three Laboratories: Laboratory of Environmental Physics, Laboratory of Neutron Transport Physics and Laboratory of Physics and Modeling of Radiation Transport. The researches provided in 1994 cover: tracer transport and flows in porous media, studies on pollution in atmospheric air, physics of molecular phenomena in chromatographic detectors, studies on neutron transport in heterogenous media, studies on evaluation of neutron cross-section in the thermal region, studies on theory and utilization of neural network in data evaluation, numerical modelling of particle cascades for particle accelerator shielding purpose. In this section the description of mentioned activities as well as the information about personnel employed in the Department, papers and reports published in 1994, contribution to conferences and grants is also given.

  6. 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

  7. Automatic modeling for the Monte Carlo transport code Geant4

    International Nuclear Information System (INIS)

    Nie Fanzhi; Hu Liqin; Wang Guozhong; Wang Dianxi; Wu Yican; Wang Dong; Long Pengcheng; FDS Team

    2015-01-01

    Geant4 is a widely used Monte Carlo transport simulation package. Its geometry models could be described in Geometry Description Markup Language (GDML), but it is time-consuming and error-prone to describe the geometry models manually. This study implemented the conversion between computer-aided design (CAD) geometry models and GDML models. This method has been Studied based on Multi-Physics Coupling Analysis Modeling Program (MCAM). The tests, including FDS-Ⅱ model, demonstrated its accuracy and feasibility. (authors)

  8. Radiation heat transfer model for the SCDAP code

    International Nuclear Information System (INIS)

    Sohal, M.S.

    1984-01-01

    A radiation heat transfer model has been developed for severe fuel damage analysis which accounts for anisotropic effects of reflected radiation. The model simplifies the view factor calculation which results in significant savings in computational cost with little loss of accuracy. Radiation heat transfer rates calculated by the isotropic and anisotropic models compare reasonably well with those calculated by other models. The model is applied to an experimental nuclear rod bundle during a slow boiloff of the coolant liquid, a situation encountered during a loss of coolant accident with severe fuel damage. At lower temperatures and also lower temperature gradients in the core, the anisotropic effect was not found to be significant

  9. Radiation protection studies for medical particle accelerators using FLUKA Monte Carlo code

    International Nuclear Information System (INIS)

    Infantino, Angelo; Mostacci, Domiziano; Cicoria, Gianfranco; Lucconi, Giulia; Pancaldi, Davide; Vichi, Sara; Zagni, Federico; Marengo, Mario

    2017-01-01

    Radiation protection (RP) in the use of medical cyclotrons involves many aspects both in the routine use and for the decommissioning of a site. Guidelines for site planning and installation, as well as for RP assessment, are given in international documents; however, the latter typically offer analytic methods of calculation of shielding and materials activation, in approximate or idealised geometry set-ups. The availability of Monte Carlo (MC) codes with accurate up-to-date libraries for transport and interaction of neutrons and charged particles at energies below 250 MeV, together with the continuously increasing power of modern computers, makes the systematic use of simulations with realistic geometries possible, yielding equipment and site-specific evaluation of the source terms, shielding requirements and all quantities relevant to RP at the same time. In this work, the well-known FLUKA MC code was used to simulate different aspects of RP in the use of biomedical accelerators, particularly for the production of medical radioisotopes. In the context of the Young Professionals Award, held at the IRPA 14 conference, only a part of the complete work is presented. In particular, the simulation of the GE PETtrace cyclotron (16.5 MeV) installed at S. Orsola-Malpighi University Hospital evaluated the effective dose distribution around the equipment; the effective number of neutrons produced per incident proton and their spectral distribution; the activation of the structure of the cyclotron and the vault walls; the activation of the ambient air, in particular the production of "4"1Ar. The simulations were validated, in terms of physical and transport parameters to be used at the energy range of interest, through an extensive measurement campaign of the neutron environmental dose equivalent using a rem-counter and TLD dosemeters. The validated model was then used in the design and the licensing request of a new Positron Emission Tomography facility. (authors)

  10. User manual for version 4.3 of the Tripoli-4 Monte-Carlo method particle transport computer code

    International Nuclear Information System (INIS)

    Both, J.P.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B.

    2003-01-01

    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 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)

  11. Radiological emergency: road map for radiation accident victim transport

    International Nuclear Information System (INIS)

    Costa, V.S.G.; Alcantara, Y.P.; Lima, C.M.A.; Silva, F. C. A. da

    2017-01-01

    During a radiological or nuclear emergency, a number of necessary actions are taken, both within the radiation protection of individuals and the environment, involving many institutions and highly specialized personnel. Among them it is possible to emphasize the air transportation of radiation accident victims.The procedures and measures for the safe transport of these radiation accident victims are generally the responsibility of the armed forces, specifically the Aeronautics, with the action denominated 'Aeromedical Military Evacuation of Radiation Accident Victims'. The experience with the Radiological Accident of Goiânia demonstrated the importance of adequate preparation and response during a radiological emergency and the need for procedures and measures with regard to the transport of radiation victims are clearly defined and clearly presented for the effectiveness of the actions. This work presents the necessary actions for the transport of radiation accident victim during a radiological emergency, through the road map technique, which has been widely used in scientific technical area to facilitate understanding and show the way to be followed to reach the proposed objectives

  12. Transport of cobalt-60 industrial radiation sources

    Science.gov (United States)

    Kunstadt, Peter; Gibson, Wayne

    This paper will deal with safety aspects of the handling of Cobalt-60, the most widely used industrial radio-isotope. Cobalt-60 is a man-made radioisotope of Cobalt-59, a naturally occurring non radioactive element, that is made to order for radiation therapy and a wide range of industrial processing applications including sterilization of medical disposables, food irradiation, etc.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-02-01

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

  14. Radiation protection for human exploration of the moon and mars: Application of the mash code system

    International Nuclear Information System (INIS)

    Johnson, J.O.; Santoro, R.T.; Drischler, J.D.; Barnes, J.M.

    1992-01-01

    The Monte Carlo Adjoint Shielding code system -- MASH, developed for the Department of Defense for calculating radiation protection factors for armored vehicles against neutron and gamma radiation, has been used to assess the dose from reactor radiation to an occupant in a habitat on Mars. The capability of MASH to reproduce measured data is summarized to demonstrate the accuracy of the code. The estimation of the radiation environment in an idealized reactor-habitat model is reported to illustrate the merits of the adjoint Monte Carlo procedure for space related studies. The reactor radiation dose for different reactor-habitat surface configurations to a habitat occupant is compared with the natural radiation dose acquired during a 500-day Mars mission

  15. bhlight: GENERAL RELATIVISTIC RADIATION MAGNETOHYDRODYNAMICS WITH MONTE CARLO TRANSPORT

    International Nuclear Information System (INIS)

    Ryan, B. R.; Gammie, C. F.; Dolence, J. C.

    2015-01-01

    We present bhlight, a numerical scheme for solving the equations of general relativistic radiation magnetohydrodynamics using a direct Monte Carlo solution of the frequency-dependent radiative transport equation. bhlight is designed to evolve black hole accretion flows at intermediate accretion rate, in the regime between the classical radiatively efficient disk and the radiatively inefficient accretion flow (RIAF), in which global radiative effects play a sub-dominant but non-negligible role in disk dynamics. We describe the governing equations, numerical method, idiosyncrasies of our implementation, and a suite of test and convergence results. We also describe example applications to radiative Bondi accretion and to a slowly accreting Kerr black hole in axisymmetry

  16. The application of the Monte Carlo code FLUKA in radiation protection studies for the large hadron collider

    International Nuclear Information System (INIS)

    Battistoni, Giuseppe; Broggi, Francesco; Brugger, Markus

    2010-01-01

    The multi-purpose particle interaction and transport code FLUKA is integral part of all radiation protection studies for the design and operation of the Large Hadron Collider (LHC) at CERN. It is one of the very few codes available for this type of calculations which is capable to calculate in one and the same simulation proton-proton and heavy ion collisions at LHC energies as well as the entire hadronic and electromagnetic particle cascade initiated by secondary particles in detectors and beam-line components from TeV energies down to energies of thermal neutrons. The present paper reviews these capabilities of FLUKA in giving details of relevant physics models along with examples of radiation protection studies for the LHC such as shielding studies for underground areas occupied by personnel during LHC operation and the simulation of induced radioactivity around beam loss points. Integral part of the FLUKA development is a careful benchmarking of specific models as well as the code performance in complex, real life applications which is demonstrated with examples of studies relevant to radiation protection at the LHC. (author)

  17. Quasilinear simulation of auroral kilometric radiation by a relativistic Fokker-Planck code

    International Nuclear Information System (INIS)

    Matsuda, Y.

    1991-01-01

    An intense terrestrial radiation called the auroral kilometric radiation (AKR) is believed to be generated by cyclotron maser instability. We study a quasilinear evolution of this instability by means of a two-dimensional relativistic Fokker-Planck code which treats waves and distributions self-consistently, including radiation loss and electron source and sink. We compare the distributions and wave amplitude with spacecraft observations to elucidate physical processes involved. 3 refs., 1 fig

  18. Radiation transport in statistically inhomogeneous rocks

    International Nuclear Information System (INIS)

    Lukhminskij, B.E.

    1975-01-01

    A study has been made of radiation transfer in statistically inhomogeneous rocks. Account has been taken of the statistical character of rock composition through randomization of density. Formulas are summarized for sigma-distribution, homogeneous density, the Simpson and Cauchy distributions. Consideration is given to the statistics of mean square ranges in a medium, simulated by the jump Markov random function. A quantitative criterion of rock heterogeneity is proposed

  19. Verification of the network flow and transport/distributed velocity (NWFT/DVM) computer code

    International Nuclear Information System (INIS)

    Duda, L.E.

    1984-05-01

    The Network Flow and Transport/Distributed Velocity Method (NWFT/DVM) computer code was developed primarily to fulfill a need for a computationally efficient ground-water flow and contaminant transport capability for use in risk analyses where, quite frequently, large numbers of calculations are required. It is a semi-analytic, quasi-two-dimensional network code that simulates ground-water flow and the transport of dissolved species (radionuclides) in a saturated porous medium. The development of this code was carried out under a program funded by the US Nuclear Regulatory Commission (NRC) to develop a methodology for assessing the risk from disposal of radioactive wastes in deep geologic formations (FIN: A-1192 and A-1266). In support to the methodology development program, the NRC has funded a separate Maintenance of Computer Programs Project (FIN: A-1166) to ensure that the codes developed under A-1192 or A-1266 remain consistent with current operating systems, are as error-free as possible, and have up-to-date documentations for reference by the NRC staff. Part of this effort would include verification and validation tests to assure that a code correctly performs the operations specified and/or is representing the processes or system for which it is intended. This document contains four verification problems for the NWFT/DVM computer code. Two of these problems are analytical verifications of NWFT/DVM where results are compared to analytical solutions. The other two are code-to-code verifications where results from NWFT/DVM are compared to those of another computer code. In all cases NWFT/DVM showed good agreement with both the analytical solutions and the results from the other code

  20. Integrated Radiation Transport and Nuclear Fuel Performance for Assembly-Level Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin T [ORNL; Hamilton, Steven P [ORNL; Philip, Bobby [ORNL; Berrill, Mark A [ORNL; Sampath, Rahul S [ORNL; Allu, Srikanth [ORNL; Pugmire, Dave [ORNL; Dilts, Gary [Los Alamos National Laboratory (LANL); Banfield, James E [ORNL

    2012-02-01

    The Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) is focused on predicting the temperature and strain within a nuclear fuel assembly to evaluate the performance and safety of existing and advanced nuclear fuel bundles within existing and advanced nuclear reactors. AMPFuel was extended to include an integrated nuclear fuel assembly capability for (one-way) coupled radiation transport and nuclear fuel assembly thermo-mechanics. This capability is the initial step toward incorporating an improved predictive nuclear fuel assembly modeling capability to accurately account for source-terms and boundary conditions of traditional (single-pin) nuclear fuel performance simulation, such as the neutron flux distribution, coolant conditions, and assembly mechanical stresses. A novel scheme is introduced for transferring the power distribution from the Scale/Denovo (Denovo) radiation transport code (structured, Cartesian mesh with smeared materials within each cell) to AMPFuel (unstructured, hexagonal mesh with a single material within each cell), allowing the use of a relatively coarse spatial mesh (10 million elements) for the radiation transport and a fine spatial mesh (3.3 billion elements) for thermo-mechanics with very little loss of accuracy. In addition, a new nuclear fuel-specific preconditioner was developed to account for the high aspect ratio of each fuel pin (12 feet axially, but 1 4 inches in diameter) with many individual fuel regions (pellets). With this novel capability, AMPFuel was used to model an entire 17 17 pressurized water reactor fuel assembly with many of the features resolved in three dimensions (for thermo-mechanics and/or neutronics), including the fuel, gap, and cladding of each of the 264 fuel pins; the 25 guide tubes; the top and bottom structural regions; and the upper and lower (neutron) reflector regions. The final, full assembly calculation was executed on Jaguar using 40,000 cores in under 10 hours to model over 162

  1. Verification and Validation of The Tritium Transport Code TMAP7

    International Nuclear Information System (INIS)

    Glen R. Longhurst; James Ambrosek

    2004-01-01

    The TMAP Code was written at the Idaho National Engineering and Environmental Laboratory in the late 1980s as a tool for safety analysis of systems involving tritium. Since then it has been upgraded several times and has been used in numerous applications including experiments supporting fusion safety, predictions for advanced systems such as the International Thermonuclear Experimental Reactor (ITER), and estimates involving tritium production technologies. Its most recent upgrade to TMAP7 was accomplished in response to several needs. Prior versions had the capacity to deal with only a single trap for diffusing gaseous species in solid structures. TMAP7 includes up to three separate traps and up to 10 diffusing species. The original code had difficulty dealing with heteronuclear molecule formation such as HD and DT. That has been removed. Under pre-specified boundary enclosure conditions and solution-law dependent diffusion boundary conditions, such as Sieverts' law, TMAP7 automatically generates heteronuclear molecular partial pressures when solubilities and partial pressures of the homonuclear molecular species are provided for law-dependent diffusion boundary conditions. A further sophistication is the addition of non-diffusing surface species. Atoms such as oxygen or nitrogen or formation of hydroxyl radicals on metal surfaces are sometimes important in molecule formation with diffusing hydrogen isotopes but do not, themselves, diffuse appreciably in the material. TMAP7 will accommodate up to 30 such surface species, allowing the user to specify relationships between those surface concentrations and partial pressures of gaseous species above the surfaces or to form them dynamically by combining diffusion species or other surface species. Additionally, TMAP7 allows the user to include a surface binding energy and an adsorption barrier energy and includes asymmetrical diffusion between the surface sites and regular diffusion sites in the bulk. All of the

  2. The use of the SRIM code for calculation of radiation damage induced by neutrons

    Science.gov (United States)

    Mohammadi, A.; Hamidi, S.; Asadabad, Mohsen Asadi

    2017-12-01

    Materials subjected to neutron irradiation will being evolve to structural changes by the displacement cascades initiated by nuclear reaction. This study discusses a methodology to compute primary knock-on atoms or PKAs information that lead to radiation damage. A program AMTRACK has been developed for assessing of the PKAs information. This software determines the specifications of recoil atoms (using PTRAC card of MCNPX code) and also the kinematics of interactions. The deterministic method was used for verification of the results of (MCNPX+AMTRACK). The SRIM (formely TRIM) code is capable to compute neutron radiation damage. The PKAs information was extracted by AMTRACK program, which can be used as an input of SRIM codes for systematic analysis of primary radiation damage. Then the Bushehr Nuclear Power Plant (BNPP) radiation damage on reactor pressure vessel is calculated.

  3. 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)

  4. ITS version 5.0 :the integrated TIGER series of coupled electron/Photon monte carlo transport codes with CAD geometry.

    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.

  5. Deciphering free-radical code of radiation effects

    International Nuclear Information System (INIS)

    Volovyk, S.; Bazyka, D.; Loganovsky, K.; Bebeshko, V.

    2007-01-01

    Complete text of publication follows. Objective: Ionizing radiation is fundamental environmental factor for life origin and evolution. Free radicals, primordial 'sea' for life conceiving and existence, induced by cosmic and terrestrial background radiation, are evolutionally archetypal, ubiquitous, and omnipotent in physiological- pathophysiological dichotomy. Classical free-radical paradigm in radiation biology and medicine, focused in essence on oxidative damage, needs new conceptualization and generalization. Methods: Suggested novel insights into free radicals dual immanent nature and functions in organism systems are based on original concepts of radicals dynamic charge transfer (CT) - redox ambivalence (interactional nucleo-, electro-, and ambiphilicity spectrum); pertinent chemical reactivity and selectivity delocalization model; physiological functional ambivalence and complementarity, and dynamic free-radical homeostasis. Results: Subtle perturbations in radicals CT spatiotemporal homeodynamics, in responsive signaling / controlling networks, concomitant alterations in genes expression, transcription, and apoptosis, redox control of mitochondrial ET chain, telomere/telomerase balance, DNA CT, circadian clock, hemispheric biochemical dominance/accentuation, including alteration of nitric oxide-superoxide complementarity, membranes permeability, neurotransmission pattern, synaptic circuitry, etc under radiation exposure have more fundamental impact on organism systems (especially CNS and CVS) deterioration than simple radicals inflicted oxidative (nitrosative) damage of cellular constituents. Conclusions: This novel conceptualization of free-radical paradigm constitutes new dimension in deciphering molecular mechanisms of radiation effects on subtle borderline norm-pathology and continuity-discontinuity dichotomy in organisms systems disorders - CT(redox)omics, which involves investigation of CT, redox, and spin states of free radicals, DNA bases

  6. An inverse method for radiation transport

    Energy Technology Data Exchange (ETDEWEB)

    Favorite, J. A. (Jeffrey A.); Sanchez, R. (Richard)

    2004-01-01

    Adjoint functions have been used with forward functions to compute gradients in implicit (iterative) solution methods for inverse problems in optical tomography, geoscience, thermal science, and other fields, but only once has this approach been used for inverse solutions to the Boltzmann transport equation. In this paper, this approach is used to develop an inverse method that requires only angle-independent flux measurements, rather than angle-dependent measurements as was done previously. The method is applied to a simplified form of the transport equation that does not include scattering. The resulting procedure uses measured values of gamma-ray fluxes of discrete, characteristic energies to determine interface locations in a multilayer shield. The method was implemented with a Newton-Raphson optimization algorithm, and it worked very well in numerical one-dimensional spherical test cases. A more sophisticated optimization method would better exploit the potential of the inverse method.

  7. Verification of three dimensional triangular prismatic discrete ordinates transport code ENSEMBLE-TRIZ by comparison with Monte Carlo code GMVP

    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)

  8. Radiative transport-based frequency-domain fluorescence tomography

    International Nuclear Information System (INIS)

    Joshi, Amit; Rasmussen, John C; Sevick-Muraca, Eva M; Wareing, Todd A; McGhee, John

    2008-01-01

    We report the development of radiative transport model-based fluorescence optical tomography from frequency-domain boundary measurements. The coupled radiative transport model for describing NIR fluorescence propagation in tissue is solved by a novel software based on the established Attila(TM) particle transport simulation platform. The proposed scheme enables the prediction of fluorescence measurements with non-contact sources and detectors at a minimal computational cost. An adjoint transport solution-based fluorescence tomography algorithm is implemented on dual grids to efficiently assemble the measurement sensitivity Jacobian matrix. Finally, we demonstrate fluorescence tomography on a realistic computational mouse model to locate nM to μM fluorophore concentration distributions in simulated mouse organs

  9. Induced Compton scattering effects in radiation transport approximations

    International Nuclear Information System (INIS)

    Gibson, D.R. Jr.

    1982-01-01

    In this thesis the method of characteristics is used to solve radiation transport problems with induced Compton scattering effects included. The methods used to date have only addressed problems in which either induced Compton scattering is ignored, or problems in which linear scattering is ignored. Also, problems which include both induced Compton scattering and spatial effects have not been considered previously. The introduction of induced scattering into the radiation transport equation results in a quadratic nonlinearity. Methods are developed to solve problems in which both linear and nonlinear Compton scattering are important. Solutions to scattering problems are found for a variety of initial photon energy distributions

  10. Induced Compton-scattering effects in radiation-transport approximations

    International Nuclear Information System (INIS)

    Gibson, D.R. Jr.

    1982-02-01

    The method of characteristics is used to solve radiation transport problems with induced Compton scattering effects included. The methods used to date have only addressed problems in which either induced Compton scattering is ignored, or problems in which linear scattering is ignored. Also, problems which include both induced Compton scattering and spatial effects have not been considered previously. The introduction of induced scattering into the radiation transport equation results in a quadratic nonlinearity. Methods are developed to solve problems in which both linear and nonlinear Compton scattering are important. Solutions to scattering problems are found for a variety of initial photon energy distributions

  11. 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

  12. Neutron secondary-particle production cross sections and their incorporation into Monte-Carlo transport codes

    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

  13. METHES: A Monte Carlo collision code for the simulation of electron transport in low temperature plasmas

    Science.gov (United States)

    Rabie, M.; Franck, C. M.

    2016-06-01

    We present a freely available MATLAB code for the simulation of electron transport in arbitrary gas mixtures in the presence of uniform electric fields. For steady-state electron transport, the program provides the transport coefficients, reaction rates and the electron energy distribution function. The program uses established Monte Carlo techniques and is compatible with the electron scattering cross section files from the open-access Plasma Data Exchange Project LXCat. The code is written in object-oriented design, allowing the tracing and visualization of the spatiotemporal evolution of electron swarms and the temporal development of the mean energy and the electron number due to attachment and/or ionization processes. We benchmark our code with well-known model gases as well as the real gases argon, N2, O2, CF4, SF6 and mixtures of N2 and O2.

  14. GPU - Accelerated Monte Carlo electron transport methods: development and application for radiation dose calculations using 6 GPU cards

    International Nuclear Information System (INIS)

    Su, L.; Du, X.; Liu, T.; Xu, X. G.

    2013-01-01

    An electron-photon coupled Monte Carlo code ARCHER - Accelerated Radiation-transport Computations in Heterogeneous EnviRonments - is being developed at Rensselaer Polytechnic Institute as a software test-bed for emerging heterogeneous high performance computers that utilize accelerators such as GPUs (Graphics Processing Units). This paper presents the preliminary code development and the testing involving radiation dose related problems. In particular, the paper discusses the electron transport simulations using the class-II condensed history method. The considered electron energy ranges from a few hundreds of keV to 30 MeV. As for photon part, photoelectric effect, Compton scattering and pair production were simulated. Voxelized geometry was supported. A serial CPU (Central Processing Unit)code was first written in C++. The code was then transplanted to the GPU using the CUDA C 5.0 standards. The hardware involved a desktop PC with an Intel Xeon X5660 CPU and six NVIDIA Tesla M2090 GPUs. The code was tested for a case of 20 MeV electron beam incident perpendicularly on a water-aluminum-water phantom. The depth and later dose profiles were found to agree with results obtained from well tested MC codes. Using six GPU cards, 6*10 6 electron histories were simulated within 2 seconds. In comparison, the same case running the EGSnrc and MCNPX codes required 1645 seconds and 9213 seconds, respectively. On-going work continues to test the code for different medical applications such as radiotherapy and brachytherapy. (authors)

  15. Radiation transport around Kerr black holes

    Science.gov (United States)

    Schnittman, Jeremy David

    This Thesis describes the basic framework of a relativistic ray-tracing code for analyzing accretion processes around Kerr black holes. We begin in Chapter 1 with a brief historical summary of the major advances in black hole astrophysics over the past few decades. In Chapter 2 we present a detailed description of the ray-tracing code, which can be used to calculate the transfer function between the plane of the accretion disk and the detector plane, an important tool for modeling relativistically broadened emission lines. Observations from the Rossi X-Ray Timing Explorer have shown the existence of high frequency quasi-periodic oscillations (HFQPOs) in a number of black hole binary systems. In Chapter 3, we employ a simple "hot spot" model to explain the position and amplitude of these HFQPO peaks. The power spectrum of the periodic X-ray light curve consists of multiple peaks located at integral combinations of the black hole coordinate frequencies, with the relative amplitude of each peak determined by the orbital inclination, eccentricity, and hot spot arc length. In Chapter 4, we introduce additional features to the model to explain the broadening of the QPO peaks as well as the damping of higher frequency harmonics in the power spectrum. The complete model is used to fit the power spectra observed in XTE J1550-564, giving confidence limits on each of the model parameters. In Chapter 5 we present a description of the structure of a relativistic alpha- disk around a Kerr black hole. Given the surface temperature of the disk, the observed spectrum is calculated using the transfer function mentioned above. The features of this modified thermal spectrum may be used to infer the physical properties of the accretion disk and the central black hole. In Chapter 6 we develop a Monte Carlo code to calculate the detailed propagation of photons from a hot spot emitter scattering through a corona surrounding the black hole. The coronal scattering has two major observable

  16. Mesh generation and energy group condensation studies for the jaguar deterministic transport code

    International Nuclear Information System (INIS)

    Kennedy, R. A.; Watson, A. M.; Iwueke, C. I.; Edwards, E. J.

    2012-01-01

    The deterministic transport code Jaguar is introduced, and the modeling process for Jaguar is demonstrated using a two-dimensional assembly model of the Hoogenboom-Martin Performance Benchmark Problem. This single assembly model is being used to test and analyze optimal modeling methodologies and techniques for Jaguar. This paper focuses on spatial mesh generation and energy condensation techniques. In this summary, the models and processes are defined as well as thermal flux solution comparisons with the Monte Carlo code MC21. (authors)

  17. 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

  18. Mesh generation and energy group condensation studies for the jaguar deterministic transport code

    Energy Technology Data Exchange (ETDEWEB)

    Kennedy, R. A.; Watson, A. M.; Iwueke, C. I.; Edwards, E. J. [Knolls Atomic Power Laboratory, Bechtel Marine Propulsion Corporation, P.O. Box 1072, Schenectady, NY 12301-1072 (United States)

    2012-07-01

    The deterministic transport code Jaguar is introduced, and the modeling process for Jaguar is demonstrated using a two-dimensional assembly model of the Hoogenboom-Martin Performance Benchmark Problem. This single assembly model is being used to test and analyze optimal modeling methodologies and techniques for Jaguar. This paper focuses on spatial mesh generation and energy condensation techniques. In this summary, the models and processes are defined as well as thermal flux solution comparisons with the Monte Carlo code MC21. (authors)

  19. 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)

  20. Systems guide to MCNP (Monte Carlo Neutron and Photon Transport Code)

    International Nuclear Information System (INIS)

    Kirk, B.L.; West, J.T.

    1984-06-01

    The subject of this report is the implementation of the Los Alamos National Laboratory Monte Carlo Neutron and Photon Transport Code - Version 3 (MCNP) on the different types of computer systems, especially the IBM MVS system. The report supplements the documentation of the RSIC computer code package CCC-200/MCNP. Details of the procedure to follow in executing MCNP on the IBM computers, either in batch mode or interactive mode, are provided

  1. Radiation exposure during air and ground transportation

    International Nuclear Information System (INIS)

    Hsu, P.C.; Weng, P.S.

    1976-01-01

    The results of a one year study program of radiation exposure experienced on both domestic and international flights of the China Airline and the Far East Airline in the Pacific, Southeast Asia and Taiwan areas and on trains and buses on Taiwan island are reported. CaSO 4 :Dy thermoluminescent dosimeters were used. It has been shown that transit exposures may amount to 10 times that on the ground with an altitude varying from 3,050 to 12,200 m. (U.K.)

  2. Regulatory practices of radiation safety of SNF transportation in Russia

    International Nuclear Information System (INIS)

    Kuryndina, Lidia; Kuryndin, Anton; Stroganov, Anatoly

    2008-01-01

    This paper overviews current regulatory practices for the assurance of nuclear and radiation safety during railway transportation of SNF on the territory of Russian Federation from NPPs to longterm-storage of reprocessing sites. The legal and regulatory requirements (mostly compliant with IAEA ST-1), licensing procedure for NM transportation are discussed. The current procedure does not require a regulatory approval for each particular shipment if the SNF fully comply with the Rosatom's branch standard and is transported in approved casks. It has been demonstrated that SNF packages compliant with the branch standard, which is knowingly provide sufficient safety margin, will conform to the federal level regulations. The regulatory approval is required if a particular shipment does not comply with the branch standard. In this case, the shipment can be approved only after regulatory review of Applicant's documents to demonstrate that the shipment still conformant to the higher level (federal) regulations. The regulatory review frequently needs a full calculation test of the radiation safety assurance. This test can take a lot of time. That's why the special calculation tools were created in SEC NRS. These tools aimed for precision calculation of the radiation safety parameters by SNF transportation use preliminary calculated Green's functions. Such approach allows quickly simulate any source distribution and optimize spent fuel assemblies placement in cask due to the transport equation property of linearity relatively the source. The short description of calculation tools are presented. Also, the paper discusses foreseen implications related to transportation of mixed-oxide SNF. (author)

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

    CERN Document Server

    Ilic, R D; Stankovic, S J

    2002-01-01

    This paper describes new Monte Carlo codes for proton transport simulations in complex geometrical forms and in materials of different composition. The SRNA codes were developed for three dimensional (3D) dose distribution calculation in proton therapy and dosimetry. The model of these codes is based on the theory of proton multiple scattering and a simple model of compound nucleus decay. The developed package consists of two codes: SRNA-2KG and SRNA-VOX. The first code simulates proton transport in combined geometry that can be described by planes and second order surfaces. The second one uses the voxelized geometry of material zones and is specifically adopted for the application of patient computer tomography data. Transition probabilities for both codes are given by the SRNADAT program. In this paper, we will present the models and algorithms of our programs, as well as the results of the numerical experiments we have carried out applying them, along with the results of proton transport simulation obtaine...

  4. 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

  5. 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)

  6. 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.

  7. 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.

  8. Design of sampling tools for Monte Carlo particle transport code JMCT

    International Nuclear Information System (INIS)

    Shangguan Danhua; Li Gang; Zhang Baoyin; Deng Li

    2012-01-01

    A class of sampling tools for general Monte Carlo particle transport code JMCT is designed. Two ways are provided to sample from distributions. One is the utilization of special sampling methods for special distribution; the other is the utilization of general sampling methods for arbitrary discrete distribution and one-dimensional continuous distribution on a finite interval. Some open source codes are included in the general sampling method for the maximum convenience of users. The sampling results show sampling correctly from distribution which are popular in particle transport can be achieved with these tools, and the user's convenience can be assured. (authors)

  9. Metropol: A computer code for the simulation of transport of contaminants with groundwater

    International Nuclear Information System (INIS)

    Sauter, F.J.; Hassanizadeh, S.M.; Leijnse, A.; Glasbergen, P.; Slot, A.F.M.

    1990-01-01

    In this report a description is given of the computer code Metropol. This code simulates the three-dimensional flow of groundwater with varying density and the simultaneous transport of contaminants in low concentration and is based on the finite element method. The basic equations for groundwater flow and transport are described as well as the mathematical techniques used to solve these equations. Pre-processing facilities for mesh generation and post-processing facilities such as particle tracking are also discussed. This work was part of the Community Mirage project Second phase, research area Calculation tools

  10. Chemical and kinetic equilibrations via radiative parton transport

    International Nuclear Information System (INIS)

    Zhang Bin; Wortman, Warner A

    2011-01-01

    A hot and dense partonic system can be produced in the early stage of a relativistic heavy ion collision. How it equilibrates is important for the extraction of Quark-Gluon Plasma properties. We study the chemical and kinetic equilibrations of the Quark-Gluon Plasma using a radiative transport model. Thermal and Color-Glass-Condensate motivated initial conditions are used. We observe that screened parton interactions always lead to partial pressure isotropization. Different initial pressure anisotropies result in the same asymptotic evolution. Comparison of evolutions with and without radiative processes shows that chemical equilibration interacts with kinetic equilibration and radiative processes can contribute significantly to pressure isotropization.

  11. SCORCH - a zero dimensional plasma evolution and transport code for use in small and large tokamak systems

    International Nuclear Information System (INIS)

    Clancy, B.E.; Cook, J.L.

    1984-12-01

    The zero-dimensional code SCORCH determines number density and temperature evolution in plasmas using concepts derived from the Hinton and Hazeltine transport theory. The code uses the previously reported ADL-1 data library

  12. General relativistic radiative transfer code in rotating black hole space-time: ARTIST

    Science.gov (United States)

    Takahashi, Rohta; Umemura, Masayuki

    2017-02-01

    We present a general relativistic radiative transfer code, ARTIST (Authentic Radiative Transfer In Space-Time), that is a perfectly causal scheme to pursue the propagation of radiation with absorption and scattering around a Kerr black hole. The code explicitly solves the invariant radiation intensity along null geodesics in the Kerr-Schild coordinates, and therefore properly includes light bending, Doppler boosting, frame dragging, and gravitational redshifts. The notable aspect of ARTIST is that it conserves the radiative energy with high accuracy, and is not subject to the numerical diffusion, since the transfer is solved on long characteristics along null geodesics. We first solve the wavefront propagation around a Kerr black hole that was originally explored by Hanni. This demonstrates repeated wavefront collisions, light bending, and causal propagation of radiation with the speed of light. We show that the decay rate of the total energy of wavefronts near a black hole is determined solely by the black hole spin in late phases, in agreement with analytic expectations. As a result, the ARTIST turns out to correctly solve the general relativistic radiation fields until late phases as t ˜ 90 M. We also explore the effects of absorption and scattering, and apply this code for a photon wall problem and an orbiting hotspot problem. All the simulations in this study are performed in the equatorial plane around a Kerr black hole. The ARTIST is the first step to realize the general relativistic radiation hydrodynamics.

  13. Signal Processing Model for Radiation Transport

    Energy Technology Data Exchange (ETDEWEB)

    Chambers, D H

    2008-07-28

    This note describes the design of a simplified gamma ray transport model for use in designing a sequential Bayesian signal processor for low-count detection and classification. It uses a simple one-dimensional geometry to describe the emitting source, shield effects, and detector (see Fig. 1). At present, only Compton scattering and photoelectric absorption are implemented for the shield and the detector. Other effects may be incorporated in the future by revising the expressions for the probabilities of escape and absorption. Pair production would require a redesign of the simulator to incorporate photon correlation effects. The initial design incorporates the physical effects that were present in the previous event mode sequence simulator created by Alan Meyer. The main difference is that this simulator transports the rate distributions instead of single photons. Event mode sequences and other time-dependent photon flux sequences are assumed to be marked Poisson processes that are entirely described by their rate distributions. Individual realizations can be constructed from the rate distribution using a random Poisson point sequence generator.

  14. MC21 v.6.0 - A continuous-energy Monte Carlo particle transport code with integrated reactor feedback capabilities

    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

  15. 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)

  16. Development of a relativistic Particle In Cell code PARTDYN for linear accelerator beam transport

    Energy Technology Data Exchange (ETDEWEB)

    Phadte, D., E-mail: deepraj@rrcat.gov.in [LPD, Raja Ramanna Centre for Advanced Technology, Indore 452013 (India); Patidar, C.B.; Pal, M.K. [MAASD, Raja Ramanna Centre for Advanced Technology, Indore (India)

    2017-04-11

    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.

  17. 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

  18. Modification of the finite element heat and mass transfer code (FEHMN) to model multicomponent reactive transport

    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

  19. Multi-Dimensional Radiation Transport in Dense Z-pinch Wire Array Plasmas

    Science.gov (United States)

    Jennings, C. A.; Chittenden, J. P.; Ciardi, A.; Sherlock, M.; Lebedev, S. V.

    2004-11-01

    Z-pinch wire arrays have proven to be an extremely efficient high yield, short pulse x-ray source with potential application to ICF. The characteristics of the x-ray pulse produced have been shown to be largely determined by non-uniform break up of the wires leading to a highly irregular distribution of mass which implodes towards the axis. Modelling the inherent 3D nature of these plasmas is already computationally very expensive, and so energy exchange through radiation is frequently neglected, assuming instead an optically thin radiation loss model. With a significant fraction of the total energy at late stages being radiated through a dense, optically thick plasma this approach is potentially inadequate in fully describing the implosion. We analyse the effects of radiative cooling and radiation transport on stagnation and precursor development in wire array z-pinch implosions. A three temperature multidimensional MHD code using a single group radiation diffusion model is used to study radiation trapping in the precursor, and the effects of preheating on the implosion dynamics. Energy exchange in the final stagnated plasma and its effects on the x-ray pulse shape is also discussed. This work was partially supported by the SSAA program of the NNSA through DoE cooperative agreement DE-F03-02NA00057.

  20. Radiation Belt Transport Driven by Solar Wind Dynamic Pressure Fluctuations

    Science.gov (United States)

    Kress, B. T.; Hudson, M. K.; Ukhorskiy, A. Y.; Mueller, H.

    2012-12-01

    The creation of the Earth's outer zone radiation belts is attributed to earthward transport and adiabatic acceleration of electrons by drift-resonant interactions with electromagnetic fluctuations in the magnetosphere. Three types of radial transport driven by solar wind dynamic pressure fluctuations that have been identified are: (1) radial diffusion [Falthammer, 1965], (2) significant changes in the phase space density radial profile due to a single or few ULF drift-resonant interactions [Ukhorskiy et al., 2006; Degeling et al., 2008], and (3) shock associated injections of radiation belt electrons occurring in less than a drift period [Li et al., 1993]. A progress report will be given on work to fully characterize different forms of radial transport and their effect on the Earth's radiation belts. The work is being carried out by computing test-particle trajectories in electric and magnetic fields from a simple analytic ULF field model and from global MHD simulations of the magnetosphere. Degeling, A. W., L. G. Ozeke, R. Rankin, I. R. Mann, and K. Kabin (2008), Drift resonant generation of peaked relativistic electron distributions by Pc 5 ULF waves, textit{J. Geophys. Res., 113}, A02208, doi:10.1029/2007JA012411. Fälthammar, C.-G. (1965), Effects of Time-Dependent Electric Fields on Geomagnetically Trapped Radiation, J. Geophys. Res., 70(11), 2503-2516, doi:10.1029/JZ070i011p02503. Li, X., I. Roth, M. Temerin, J. R. Wygant, M. K. Hudson, and J. B. Blake (1993), Simulation of the prompt energization and transport of radiation belt particles during the March 24, 1991 SSC, textit{Geophys. Res. Lett., 20}(22), 2423-2426, doi:10.1029/93GL02701. Ukhorskiy, A. Y., B. J. Anderson, K. Takahashi, and N. A. Tsyganenko (2006), Impact of ULF oscillations in solar wind dynamic pressure on the outer radiation belt electrons, textit{Geophys. Res. Lett., 33}(6), L06111, doi:10.1029/2005GL024380.

  1. Computer code selection criteria for flow and transport code(s) to be used in undisturbed vadose zone calculations for TWRS environmental analyses

    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

  2. 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.

  3. Memory bottlenecks and memory contention in multi-core Monte Carlo transport codes

    International Nuclear Information System (INIS)

    Tramm, J.R.; Siegel, A.R.

    2013-01-01

    The simulation of whole nuclear cores through the use of Monte Carlo codes requires an impracticably long time-to-solution. We have extracted a kernel that executes only the most computationally expensive steps of the Monte Carlo particle transport algorithm - the calculation of macroscopic cross sections - in an effort to expose bottlenecks within multi-core, shared memory architectures. (authors)

  4. 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

  5. Use of the APOLLO2 transport code for PWR assembly studies

    International Nuclear Information System (INIS)

    Belhaffaf, D.; Coste, M.; Lenain, R.; Mathonniere, G.; Sanchez, R.; Stankovski, Z.; Zmijarevic, I.

    1992-01-01

    This paper presents some validation and application aspects of the APOLLO2, user oriented, modular code for multigroup transport assembly calculation which is developed at the French Commissariat a l'Energies Atomique. The main points approached in this paper are: the two dimensional collision probability convergence, critical leakage calculation schemes, self-shielding spatial discretization, and the equivalence procedure

  6. 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

  7. 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.

  8. The neutron transport code DTF-TRACA. User's manual and input data

    International Nuclear Information System (INIS)

    Anhert, C.

    1979-01-01

    A user's 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 description is given. The new options developped at JEN are included too. (author)

  9. User's manual for the Oak Ridge Tokamak Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. Overview and applications of the Monte Carlo radiation transport kit at LLNL

    International Nuclear Information System (INIS)

    Sale, K. E.

    1999-01-01

    Modern Monte Carlo radiation transport codes can be applied to model most applications of radiation, from optical to TeV photons, from thermal neutrons to heavy ions. Simulations can include any desired level of detail in three-dimensional geometries using the right level of detail in the reaction physics. The technology areas to which we have applied these codes include medical applications, defense, safety and security programs, nuclear safeguards and industrial and research system design and control. The main reason such applications are interesting is that by using these tools substantial savings of time and effort (i.e. money) can be realized. In addition it is possible to separate out and investigate computationally effects which can not be isolated and studied in experiments. In model calculations, just as in real life, one must take care in order to get the correct answer to the right question. Advancing computing technology allows extensions of Monte Carlo applications in two directions. First, as computers become more powerful more problems can be accurately modeled. Second, as computing power becomes cheaper Monte Carlo methods become accessible more widely. An overview of the set of Monte Carlo radiation transport tools in use a LLNL will be presented along with a few examples of applications and future directions

  11. Present status of nucleon-meson transport code NMTC/JAERI

    International Nuclear Information System (INIS)

    Takada, H.; Meigo, S.

    2001-01-01

    The nucleon-meson transport code NMTC/JAM has been developed for the neutronics design study of the joint project for high-intensity proton accelerators with a power of mega-watts. The applicable energy range is extended by the inclusion of the jet AA microscopic transport model (JAM). The nucleon-nucleus cross sections are also updated for accurate transport calculation. The applicability of NMTC/JAM is studied through the analyses of thick target experiments such as neutron transmission through shield and activation reaction rate measurements. (orig.)

  12. Modeling radiation belt dynamics using a 3-D layer method code

    Science.gov (United States)

    Wang, C.; Ma, Q.; Tao, X.; Zhang, Y.; Teng, S.; Albert, J. M.; Chan, A. A.; Li, W.; Ni, B.; Lu, Q.; Wang, S.

    2017-08-01

    A new 3-D diffusion code using a recently published layer method has been developed to analyze radiation belt electron dynamics. The code guarantees the positivity of the solution even when mixed diffusion terms are included. Unlike most of the previous codes, our 3-D code is developed directly in equatorial pitch angle (α0), momentum (p), and L shell coordinates; this eliminates the need to transform back and forth between (α0,p) coordinates and adiabatic invariant coordinates. Using (α0,p,L) is also convenient for direct comparison with satellite data. The new code has been validated by various numerical tests, and we apply the 3-D code to model the rapid electron flux enhancement following the geomagnetic storm on 17 March 2013, which is one of the Geospace Environment Modeling Focus Group challenge events. An event-specific global chorus wave model, an AL-dependent statistical plasmaspheric hiss wave model, and a recently published radial diffusion coefficient formula from Time History of Events and Macroscale Interactions during Substorms (THEMIS) statistics are used. The simulation results show good agreement with satellite observations, in general, supporting the scenario that the rapid enhancement of radiation belt electron flux for this event results from an increased level of the seed population by radial diffusion, with subsequent acceleration by chorus waves. Our results prove that the layer method can be readily used to model global radiation belt dynamics in three dimensions.

  13. 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)

  14. Recent Progress of the Synchrotron Radiation Calculation Code SPECTRA

    International Nuclear Information System (INIS)

    Tanaka, T.; Kitamura, H.

    2007-01-01

    SPECTRA is a computer software to calculate optical properties of synchrotron radiation (SR) emitted by electrons passing through magnetic devices such as bending magnets, wigglers and undulators. It has been used to design various devices in the SR beamline, such as high heat-load components in the front-end section and optical elements in the optics hutch. In addition, the electron beam quality can be estimated by comparison between the measured and calculated properties of SR. Since the first announcement, numerous improvements have been made to SPECTRA to achieve less computation time with higher numerical accuracy. In addition, a number of functions have been added to follow the user's demand. In this paper, recent progress of SPECTRA is presented and details of the new functions are explained together with several examples

  15. Improvements in the CHART D radiation-hydrodynamic code III: revised analytic equations of state

    International Nuclear Information System (INIS)

    Thompson, S.L.; Lauson, H.S.

    1974-03-01

    A revised set of in-line equation-of-state subroutines for the CHART D hydrodynamic code is described. The information generated is thermodynamically complete and self-consistent. The temperature and density range of validity is large. Solids, liquids, vapors, plasmas, and all types of phase mixtures are treated. Energy transport properties are calculated. The set of subroutines form a package which can easily be included in other hydrodynamic codes. (20 figures) (U.S.)

  16. 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)

  17. CIPHER: coded imager and polarimeter for high-energy radiation

    CERN Document Server

    Caroli, E; Dusi, W; Bertuccio, G; Sampietro, M

    2000-01-01

    The CIPHER instrument is a hard X- and soft gamma-ray spectroscopic and polarimetric coded mask imager based on an array of cadmium telluride micro-spectrometers. The position-sensitive detector (PSD) will be arranged in 4 modules of 32x32 crystals, each of 2x2 mm sup 2 cross section and 10 mm thickness giving a total active area of about 160 cm sup 2. The micro-spectrometer characteristics allow a wide operating range from approx 10 keV to 1 MeV, while the PSD is actively shielded by CsI crystals on the bottom in order to reduce background. The mask, based on a modified uniformly redundant array (MURA) pattern, is four times the area of the PSD and is situated at about 100 cm from the CdTe array top surface. The CIPHER instrument is proposed for a balloon experiment, both in order to assess the performance of such an instrumental concept for a small/medium-size satellite survey mission and to perform an innovative measurement of the Crab polarisation level. The CIPHER's field of view allows the instrument to...

  18. Improved core-edge tokamak transport simulations with the CORSICA 2 code

    International Nuclear Information System (INIS)

    Tarditi, A.; Cohen, R.H.; Crotinger, J.A.

    1996-01-01

    The CORSICA 2 code models the nonlinear transport between the core and the edge of a tokamak plasma. The code couples a 2D axisymmetric edge/SOL model (UEDGE) to a 1D model for the radial core transport in toroidal flux coordinates (the transport module from the CORSICA 1 code). The core density and temperature profiles are joined to the flux-surface average profiles from the 2D code sufficiently inside the magnetic separatrix, at a flux surface on which the edge profiles are approximately constant. In the present version of the code, the deuterium density and electron and ion temperatures are coupled. The electron density is determined by imposing quasi-neutrality, both in the core and in the edge. The model allows the core-edge coupling of multiple ion densities while retaining a single temperature (corresponding to the equilibration value) for the all ion species. Applications of CORSICA 2 to modeling the DIII-D tokamak are discussed. This work will focus on the simulation of the L-H transition, coupling a single ion species (deuterium) and the two (electron and ion) temperatures. These simulations will employ a new self-consistent model for the L-H transition that is being implemented in the UEDGE code. Applications to the modeling of ITER ignition scenarios are also discussed. This will involve coupling a second density species (the thermal alphas), bringing the total number of coupled variables up to four. Finally, the progress in evolving the magnetic geometry is discussed. Currently, this geometry is calculated by CORSICA's MHD equilibrium module (TEQ) at the beginning of the run and fixed thereafter. However, CORSICA 1 can evolve this geometry quasistatically, and this quasistatic treatment is being extended to include the edge/SOL geometry. Recent improvements for code speed-up are also presented

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

    Science.gov (United States)

    Tritsis, A.; Yorke, H.; Tassis, K.

    2018-05-01

    We describe PyRaTE, a new, non-local thermodynamic equilibrium (non-LTE) line radiative transfer code developed specifically for post-processing astrochemical simulations. Population densities are estimated using the escape probability method. When computing the escape probability, the optical depth is calculated towards all directions with density, molecular abundance, temperature and velocity variations all taken into account. A very easy-to-use interface, capable of importing data from simulations outputs performed with all major astrophysical codes, is also developed. The code is written in PYTHON using an "embarrassingly parallel" strategy and can handle all geometries and projection angles. We benchmark the code by comparing our results with those from RADEX (van der Tak et al. 2007) and against analytical solutions and present case studies using hydrochemical simulations. The code will be released for public use.

  20. THE UNPREDICTABILITY CLAUSE IN TRANSPORT CONTRACTS, ACCORDING TO THE NEW CIVIL CODE

    Directory of Open Access Journals (Sweden)

    Adriana Elena BELU

    2014-05-01

    Full Text Available Until the enforcement of the highly controversial transport law, transport companies must already observe the provisions of the new Civil Code1 in their transport business. One of the novelties in the new Civil Code, that came into force on October 1, 2011, refers to the unpredictability clause: recurring to this clause, in certain situations to be precisely analysed by courts, parties may even be exempted from certain contractual obligations, when the court decides to rescind the contract based on objective criteria, not imputable to the party that no longer can properly fulfil the obligations that had been undertaken when the contract had been made. However, this solution only is provided after all means of negotiation and mediation between parties are exhausted. The clause meets current market requirements, under which many companies have to deal with bad paying partners.

  1. Savannah River Laboratory DOSTOMAN code: a compartmental pathways computer model of contaminant transport

    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

  2. Parallel processing of Monte Carlo code MCNP for particle transport problem

    Energy Technology Data Exchange (ETDEWEB)

    Higuchi, Kenji; Kawasaki, Takuji

    1996-06-01

    It is possible to vectorize or parallelize Monte Carlo codes (MC code) for photon and neutron transport problem, making use of independency of the calculation for each particle. Applicability of existing MC code to parallel processing is mentioned. As for parallel computer, we have used both vector-parallel processor and scalar-parallel processor in performance evaluation. We have made (i) vector-parallel processing of MCNP code on Monte Carlo machine Monte-4 with four vector processors, (ii) parallel processing on Paragon XP/S with 256 processors. In this report we describe the methodology and results for parallel processing on two types of parallel or distributed memory computers. In addition, we mention the evaluation of parallel programming environments for parallel computers used in the present work as a part of the work developing STA (Seamless Thinking Aid) Basic Software. (author)

  3. 76 FR 2744 - Disclosure of Code-Share Service by Air Carriers and Sellers of Air Transportation

    Science.gov (United States)

    2011-01-14

    ... DEPARTMENT OF TRANSPORTATION Office of the Secretary Disclosure of Code-Share Service by Air Carriers and Sellers of Air Transportation AGENCY: Office of the Secretary, Department of Transportation..., their agents, and third party sellers of air transportation in view of recent amendments to 49 U.S.C...

  4. Transport of radioactive materials: the need for radiation protection programmes

    International Nuclear Information System (INIS)

    Masinza, S.A.

    2004-01-01

    The increase in the use of radioactive materials worldwide requires that these materials be moved from production sites to the end user or in the case of radioactive waste, from the waste generator to the repository. Tens of millions of packages containing radioactive material are consigned for transport each year throughout the world. The amount of radioactive material in these packages varies from negligible quantities in shipments of consumer products to very large quantities of shipments of irradiated nuclear fuel. Transport is the main way in which the radioactive materials being moved get into the public domain. The public is generally unaware of the lurking danger when transporting these hazardous goods. Thus radiation protection programmes are important to assure the public of the certainty of their safety during conveyance of these materials. Radioactive material is transported by land (road and rail), inland waterways, sea/ocean and air. These modes of transport are regulated by international 'modal' regulations. The international community has formulated controls to reduce the number of accidents and mitigate their consequences should they happen. When accidents involving the transport of radioactive material occur, it could result in injury, loss of life and pollution of the environment. In order to ensure the safety of people, property and the environment, national and international transport regulations have been developed. The appropriate authorities in each state utilise them to control the transport of radioactive material. Stringent measures are required in these regulations to ensure adequate containment, shielding and the prevention of criticality in all spheres of transport, i.e. routine, minor incidents and accident conditions. Despite the extensive application of these stringent safety controls, transport accidents involving packages containing radioactive material have occurred and will continue to occur. When a transport accident occurs, it

  5. Application of radiation protection programmes to transport of radioactive material

    International Nuclear Information System (INIS)

    Lopez Vietri, Jorge; Capadona, Nancy; Barenghi, Leonardo

    2008-01-01

    Full text: The principles for implementing radiation protection programmes (RPP) are detailed in the draft IAEA safety guide TS-G-1.5 'Radiation protection programmes for transport of radioactive material'. The document is described in this paper and analysis is made for typical applications to current operations carried out by consignors, carriers and consignees. Systematic establishment and application of RPPs is a way to control radiological protection during different steps of transport activity. The most widely transported packages in the world are radiopharmaceuticals by road. It is described an application of RPP for an organization involved in road transport of Type A packages containing radiopharmaceuticals. Considerations based on the radionuclides, quantities and activities transported are the basis to design and establish the scope of the RPP for the organizations involved in transport. Next stage is the determination of roles and responsibilities for each activity related to transport of radioactive materials. An approach to the dose received by workers is evaluated considering the type, category and quantity of packages, the radionuclides, the frequency of consignments and how long are the storages. The average of transports made in the last years must be taken into account and special measures intended to optimize the protection are evaluated. Tasks like monitoring, control of surface contamination and segregation measures, are designed based on the dose evaluation and optimization. The RPP also indicates main measures to follow in case of emergency during transport taking account of radionuclides, activities and category of packages for different accident scenarios. Basis for training personnel involved in handling of radioactive materials to insure they have appropriate knowledge about preparing packages, measuring dose rates, calculating transport index, labelling, marking and placarding, transport documents, etc, are considered. The RPP is a part

  6. Transport-constrained extensions of collision and track length estimators for solutions of radiative transport problems

    International Nuclear Information System (INIS)

    Kong, Rong; Spanier, Jerome

    2013-01-01

    In this paper we develop novel extensions of collision and track length estimators for the complete space-angle solutions of radiative transport problems. We derive the relevant equations, prove that our new estimators are unbiased, and compare their performance with that of more conventional estimators. Such comparisons based on numerical solutions of simple one dimensional slab problems indicate the the potential superiority of the new estimators for a wide variety of more general transport problems

  7. Obtaining a radiation beam poly energy using the code Penelope 2006

    International Nuclear Information System (INIS)

    Andrade, Lucio das Chagas; Peixoto, Jose Guilherme Pereira

    2013-01-01

    Obtaining a spectrum X-ray is not a very easy task, one of the techniques used is the simulation by Monte Carlo method. The Penelope code is a code based on this method that simulates the transport of particles such as electrons, positrons and photons in different media and materials. The versions of this program in 2003 and 2006 show significant differences for facilitating the use of the code. The program allows the construction of the desired geometry and definitions of simulation parameters. (author)

  8. FORLI radiative transfer and retrieval code for IASI

    International Nuclear Information System (INIS)

    Hurtmans, D.; Coheur, P.-F.; Wespes, C.; Clarisse, L.; Scharf, O.; Clerbaux, C.; Hadji-Lazaro, J.; George, M.; Turquety, S.

    2012-01-01

    This paper lays down the theoretical bases and the methods used in the Fast Optimal Retrievals on Layers for IASI (FORLI) software, which is developed and maintained at the “Université Libre de Bruxelles” (ULB) with the support of the “Laboratoire Atmosphères, Milieux, Observations Spatiales” (LATMOS) to process radiance spectra from the Infrared Atmospheric Sounding Interferometer (IASI) in the perspective of local to global chemistry applications. The forward radiative transfer model (RTM) and the retrieval approaches are formulated and numerical approximations are described. The aim of FORLI is near-real-time provision of global scale concentrations of trace gases from IASI, either integrated over the altitude range of the atmosphere (total columns) or vertically resolved. To this end, FORLI uses precalculated table of absorbances. At the time of writing three gas-specific versions of this algorithm have been set up: FORLI-CO, FORLI-O 3 and FORLI-HNO 3 . The performances of each are reviewed and illustrations of results and early validations are provided, making the link to recent scientific publications. In this paper we stress the challenges raised by near-real-time processing of IASI, shortly describe the processing chain set up at ULB and draw perspectives for future developments and applications.

  9. Open-Source Development of the Petascale Reactive Flow and Transport Code PFLOTRAN

    Science.gov (United States)

    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

  10. Radiation transport modelling for the interpretation of oblique ECE measurements

    Directory of Open Access Journals (Sweden)

    Denk Severin S.

    2017-01-01

    Since radiation transport modelling is required for the interpretation of oblique ECE diagnostics we present in this paper an extended forward model that supports oblique lines of sight. To account for the refraction of the line of sight, ray tracing in the cold plasma approximation was added to the model. Furthermore, an absorption coefficient valid for arbitrary propagation was implemented. Using the revised model it is shown that for the oblique ECE Imaging diagnostic at ASDEX Upgrade there can be a significant difference between the cold resonance position and the point from which most of the observed radiation originates.

  11. MILSTAMP TACs: Military Standard Transportation and Movement Procedures Transportation Account Codes. Volume 2

    Science.gov (United States)

    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

  12. EPRI-LATTICE: a multigroup neutron transport code for light water reactor lattice physics calculations

    International Nuclear Information System (INIS)

    Jones, D.B.

    1986-01-01

    EPRI-LATTICE is a multigroup neutron transport computer code for the analysis of light water reactor fuel assemblies. It can solve the two-dimensional neutron transport problem by two distinct methods: (a) the method of collision probabilities and (b) the method of discrete ordinates. The code was developed by S. Levy Inc. as an account of work sponsored by the Electric Power Research Institute (EPRI). The collision probabilities calculation in EPRI-LATTICE (L-CP) is based on the same methodology that exists in the lattice codes CPM-2 and EPRI-CPM. Certain extensions have been made to the data representations of the CPM programs to improve the overall accuracy of the calculation. The important extensions include unique representations of scattering matrices and fission fractions (chi) for each composition in the problem. A new capability specifically developed for the EPRI-LATTICE code is a discrete ordinates methodology. The discrete ordinates calculation in EPRI-LATTICE (L-SN) is based on the discrete S/sub n/ methodology that exists in the TWODANT program. In contrast to TWODANT, which utilizes synthetic diffusion acceleration and supports multiple geometries, only the transport equations are solved by L-SN and only the data representations for the two-dimensional geometry are treated

  13. A 3D transport-based core analysis code for research reactors with unstructured geometry

    International Nuclear Information System (INIS)

    Zhang, Tengfei; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi; Li, Yunzhao

    2013-01-01

    Highlights: • A core analysis code package based on 3D neutron transport calculation in complex geometry is developed. • The fine considerations on flux mapping, control rod effects and isotope depletion are modeled. • The code is proved to be with high accuracy and capable of handling flexible operational cases for research reactors. - Abstract: As an effort to enhance the accuracy in simulating the operations of research reactors, a 3D transport core analysis code system named REFT was developed. HELIOS is employed due to the flexibility of describing complex geometry. A 3D triangular nodal S N method transport solver, DNTR, endows the package the capability of modeling cores with unstructured geometry assemblies. A series of dedicated methods were introduced to meet the requirements of research reactor simulations. Afterwards, to make it more user friendly, a graphical user interface was also developed for REFT. In order to validate the developed code system, the calculated results were compared with the experimental results. Both the numerical and experimental results are in close agreement with each other, with the relative errors of k eff being less than 0.5%. Results for depletion calculations were also verified by comparing them with the experimental data and acceptable consistency was observed in results

  14. Novel Parallel Numerical Methods for Radiation and Neutron Transport

    International Nuclear Information System (INIS)

    Brown, P N

    2001-01-01

    In many of the multiphysics simulations performed at LLNL, transport calculations can take up 30 to 50% of the total run time. If Monte Carlo methods are used, the percentage can be as high as 80%. Thus, a significant core competence in the formulation, software implementation, and solution of the numerical problems arising in transport modeling is essential to Laboratory and DOE research. In this project, we worked on developing scalable solution methods for the equations that model the transport of photons and neutrons through materials. Our goal was to reduce the transport solve time in these simulations by means of more advanced numerical methods and their parallel implementations. These methods must be scalable, that is, the time to solution must remain constant as the problem size grows and additional computer resources are used. For iterative methods, scalability requires that (1) the number of iterations to reach convergence is independent of problem size, and (2) that the computational cost grows linearly with problem size. We focused on deterministic approaches to transport, building on our earlier work in which we performed a new, detailed analysis of some existing transport methods and developed new approaches. The Boltzmann equation (the underlying equation to be solved) and various solution methods have been developed over many years. Consequently, many laboratory codes are based on these methods, which are in some cases decades old. For the transport of x-rays through partially ionized plasmas in local thermodynamic equilibrium, the transport equation is coupled to nonlinear diffusion equations for the electron and ion temperatures via the highly nonlinear Planck function. We investigated the suitability of traditional-solution approaches to transport on terascale architectures and also designed new scalable algorithms; in some cases, we investigated hybrid approaches that combined both

  15. Radiation shielding and criticality safety assessment for KN-12 spent nuclear fuel transport cask

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Kyung; Shin, Chang Ho; Kim, Gi Hwan [Hanyang Univ., Seoul (Korea, Republic of)

    2001-08-15

    Because SNFs involve TRU (Transuranium), fission products, and fissile materials, they are highly radioactive and also have a possibility to be critical. Therefore, radiation shielding and criticality safety for transport casks containing the SNFs should be guaranteed through reliable valuation procedure. IAEA safety standard series No ST-1 recommends regulation for safe transportation of the SNFs by transport casks, and United States is carrying out it according to the regulation guide, 10 CFR parts 71 and 72. Present research objective is to evaluate the KN-12 spent nuclear fuel transport cask that is designed for transportation of up to 12 assemblies and is standby status for being licensed in accordance with Korea Atomic Energy Act. Both radiation shielding and criticality analysis using the accurate Monte Carlo transport code, MCNP-4B are carried out for the KN-12 SNF cask as a benchmark calculation. Source terms for radiation shielding calculation are obtained using ORIGEN-S computer code. In this work, for normal transport conditions, the results from MCNP-4B shows the maximum dose rate of 0.557 mSv/hr at the side surface. And the maximum dose rate of 0.0871 mSv/hr was resulted at the 2 m distance from the cask. The level of calculated dose rate is 27.9% of the limit at the cask surface, 87.1% at 2 m from the cask surface for normal transport condition. For hypothetical accident conditions, the maximum rate of 2.5144 mSv/hr was resulted at the 1 m distance from the cask and this level is 25.1% of the limit for hypothetical accident conditions. In criticality calculations using MCNP-4B, the k{sub eff} values yielded for 5.0 w/o U-235 enriched fresh fuel are 0.92098 {+-} 0.00065. This result confirms subcritical condition of the KN-12 SNF cask and gives 96.95% of recommendations for criticality safety evaluation by US NRC these results will be useful as a basis for approval for the KN-12 SNF cask.

  16. Process of cross section generation for radiation shielding calculations, using the NJOY code

    International Nuclear Information System (INIS)

    Ono, S.; Corcuera, R.P.

    1986-10-01

    The process of multigroup cross sections generation for radiation shielding calculations, using the NJOY code, is explained. Photon production cross sections, processed by the GROUPR module, and photon interaction cross sections processed by the GAMINR are given. These data are compared with the data produced by the AMPX system and published data. (author) [pt

  17. 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)

  18. 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.

  19. Parallelization of MCNP Monte Carlo neutron and photon transport code in parallel virtual machine and message passing interface

    International Nuclear Information System (INIS)

    Deng Li; Xie Zhongsheng

    1999-01-01

    The coupled neutron and photon transport Monte Carlo code MCNP (version 3B) has been parallelized in parallel virtual machine (PVM) and message passing interface (MPI) by modifying a previous serial code. The new code has been verified by solving sample problems. The speedup increases linearly with the number of processors and the average efficiency is up to 99% for 12-processor. (author)

  20. Development of a coupling code for PWR reactor cavity radiation streaming calculation

    International Nuclear Information System (INIS)

    Zheng, Z.; Wu, H.; Cao, L.; Zheng, Y.; Zhang, H.; Wang, M.

    2012-01-01

    PWR reactor cavity radiation streaming is important for the safe of the personnel and equipment, thus calculation has to be performed to evaluate the neutron flux distribution around the reactor. For this calculation, the deterministic codes have difficulties in fine geometrical modeling and need huge computer resource; and the Monte Carlo codes require very long sampling time to obtain results with acceptable precision. Therefore, a coupling method has been developed to eliminate the two problems mentioned above in each code. In this study, we develop a coupling code named DORT2MCNP to link the Sn code DORT and Monte Carlo code MCNP. DORT2MCNP is used to produce a combined surface source containing top, bottom and side surface simultaneously. Because SDEF card is unsuitable for the combined surface source, we modify the SOURCE subroutine of MCNP and compile MCNP for this application. Numerical results demonstrate the correctness of the coupling code DORT2MCNP and show reasonable agreement between the coupling method and the other two codes (DORT and MCNP). (authors)

  1. User's manual for the Heat Pipe Space Radiator design and analysis Code (HEPSPARC)

    Science.gov (United States)

    Hainley, Donald C.

    1991-01-01

    A heat pipe space radiatior code (HEPSPARC), was written for the NASA Lewis Research Center and is used for the design and analysis of a radiator that is constructed from a pumped fluid loop that transfers heat to the evaporative section of heat pipes. This manual is designed to familiarize the user with this new code and to serve as a reference for its use. This manual documents the completed work and is intended to be the first step towards verification of the HEPSPARC code. Details are furnished to provide a description of all the requirements and variables used in the design and analysis of a combined pumped loop/heat pipe radiator system. A description of the subroutines used in the program is furnished for those interested in understanding its detailed workings.

  2. New scope covered by PHITS. Particle and heavy ion transport code system

    International Nuclear Information System (INIS)

    Nakamura, Takashi; Niita, Koji; Iwase, Hiroshi; Sato, Tatsuhiko

    2006-01-01

    PHITS is a general high energy transport calculation code from hadron to heavy ions, which embedded in NMTC-JAM with JQMD code. Outline of PHITS and many application examples are stated. PHITS has been used by the shielding calculations of J-PARC, GSI, RIA and Big-RIPS and the good results were reported. The evaluation of exposure dose of astronauts, airmen, proton and heavy ion therapy, and estimation of error frequency of semiconductor software are explained as the application examples. Relation between the event generator and Monte Carlo method and the future are described. (S.Y.)

  3. Accuracy estimation for intermediate and low energy neutron transport calculation with Monte Carlo code MCNP

    International Nuclear Information System (INIS)

    Kotegawa, Hiroshi; Sasamoto, Nobuo; Tanaka, Shun-ichi

    1987-02-01

    Both ''measured radioactive inventory due to neutron activation in the shield concrete of JPDR'' and ''measured intermediate and low energy neutron spectra penetrating through a graphite sphere'' are analyzed using a continuous energy model Monte Carlo code MCNP so as to estimate calculational accuracy of the code for neutron transport in thermal and epithermal energy regions. Analyses reveal that MCNP calculates thermal neutron spectra fairly accurately, while it apparently over-estimates epithermal neutron spectra (of approximate 1/E distribution) as compared with the measurements. (author)

  4. Linking the plasma code EDGE2D to the neutral code NIMBUS for a self consistent transport model of the boundary

    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

  5. Modification of the finite element heat and mass transfer code (FEHM) to model multicomponent reactive transport

    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

  6. Hybrid formulation of radiation transport in optically thick divertor plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Rosato, J.; Marandet, Y.; Bufferand, H.; Stamm, R. [PIIM, UMR 7345 Aix-Marseille Universite / CNRS, Centre de St-Jerome, Marseille (France); Reiter, D. [IEK-4 Plasmaphysik, Forschungszentrum Juelich GmbH, Juelich (Germany)

    2016-08-15

    Kinetic Monte Carlo simulations of coupled atom-radiation transport in optically thick divertor plasmas can be computationally very demanding, in particular in ITER relevant conditions or even larger devices, e.g. for power plant divertor studies. At high (∝ 10{sup 15} cm{sup -3}) atomic densities, it can be shown that sufficiently large divertors behave in certain areas like a black body near the first resonance line of hydrogen (Lyman α). This suggests that, at least in part, the use of continuum model (radiation hydrodynamics) can be sufficiently accurate, while being less time consuming. In this work, we report on the development of a hybrid model devoted to switch automatically between a kinetic and a continuum description according to the plasma conditions. Calculations of the photo-excitation rate in a homogeneous slab are performed as an illustration. The outlined hybrid concept might be also applicable to neutral atom transport, due to mathematical analogy of transport equations for neutrals and radiation. (copyright 2016 The Authors. Contributions to Plasma Physics published by Wiley-VCH Verlag GmbH and Co. KGaA Weinheim. This)

  7. 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

  8. The VENUS-7 benchmarks. Results from state-of-the-art transport codes and nuclear data

    International Nuclear Information System (INIS)

    Zwermann, Winfried; Pautz, Andreas; Timm, Wolf

    2010-01-01

    For the validation of both nuclear data and computational methods, comparisons with experimental data are necessary. Most advantageous are assemblies where not only the multiplication factors or critical parameters were measured, but also additional quantities like reactivity differences or pin-wise fission rate distributions have been assessed. Currently there is a comprehensive activity to evaluate such measure-ments and incorporate them in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. A large number of such experiments was performed at the VENUS zero power reactor at SCK/CEN in Belgium in the sixties and seventies. The VENUS-7 series was specified as an international benchmark within the OECD/NEA Working Party on Scientific Issues of Reactor Systems (WPRS), and results obtained with various codes and nuclear data evaluations were summarized. In the present paper, results of high-accuracy transport codes with full spatial resolution with up-to-date nuclear data libraries from the JEFF and ENDF/B evaluations are presented. The comparisons of the results, both code-to-code and with the measured data are augmented by uncertainty and sensitivity analyses with respect to nuclear data uncertainties. For the multiplication factors, these are performed with the TSUNAMI-3D code from the SCALE system. In addition, uncertainties in the reactivity differences are analyzed with the TSAR code which is available from the current SCALE-6 version. (orig.)

  9. Dispersive effects in radiation transport and radiation hydrodynamics in matter at high density

    International Nuclear Information System (INIS)

    Crowley, B.J.B.

    1983-01-01

    In a recent research program (reported in AWRE 0 20/82) I have investigated the generalisation of the equations of radiation hydrodynamics when electromagnetic radiation is assumed to obey a linear-response dispersion relation of the form nω=kc where the refractive index n depends on the frequency ω and/or wave number k. From the application of the Boltzmann-Liouville transport theory to photons in the short-wavelength (geometrical optics) limit, I derive the energy and momentum equations which, when combined with a classical (Euler-Lagrange-Navier-Stokes) treatment of a fluid material medium in LTE, yield a complete dynamical theory of linear interactions (+ stimulated processes) between incoherent (thermal) radiation and dense, locally isotropic matter. The theory includes an account of pondero-motive forces and electro (magneto) striction. Moreover, it is apparently capable of being generalised to non-linear interactions in which the refractive index depends on the local specific intensity of the radiation field, and, to some extent, to the treatment of high-frequency coherent radiation. The generalisation of various approximated forms of radiation-transport theory (esp. diffusion) has been considered in detail. Some problems remain however. One such is the treatment of anomalous dispersion. Current research work is concentrating on the interesting atomic physics aspects of electromagnetic (esp. radiative) properties of a dispersive material medium

  10. One-, two- and three-dimensional transport codes using multi-group double-differential form cross sections

    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)

  11. Development of the computer code to monitor gamma radiation in the nuclear facility environment

    International Nuclear Information System (INIS)

    Akhmad, Y. R.; Pudjiyanto, M.S.

    1998-01-01

    Computer codes for gamma radiation monitoring in the vicinity of nuclear facility which have been developed could be introduced to the commercial potable gamma analyzer. The crucial stage of the first year activity was succeeded ; that is the codes have been tested to transfer data file (pulse high distribution) from Micro NOMAD gamma spectrometer (ORTEC product) and the convert them into dosimetry and physics quantities. Those computer codes are called as GABATAN (Gamma Analyzer of Batan) and NAGABAT (Natural Gamma Analyzer of Batan). GABATAN code can isable to used at various nuclear facilities for analyzing gamma field up to 9 MeV, while NAGABAT could be used for analyzing the contribution of natural gamma rays to the exposure rate in the certain location

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

    International Nuclear Information System (INIS)

    Taji, Yukichi

    1984-06-01

    Development has been made of molecular dynamical codes DGR and GGR to simulate radiation damages yielded in the diamond and graphite structure crystals, respectively. Though the usual molecular dynamical codes deal only with the central forces as the mutual interactions between atoms, the present codes can take account of noncentral forces to represent the effect of the covalent bonds characteristic of diamond or graphite crystals. It is shown that lattice defects yielded in these crystals are stable by themselves in the present method without any supports of virtual surface forces set on the crystallite surfaces. By this effect the behavior of lattice defects has become possible to be simulated in a more realistic manner. Some examples of the simulation with these codes are shown. (author)

  13. Radiation doses arising from the air transport of radioactive materials

    International Nuclear Information System (INIS)

    Gelder, R.; Shaw, K.B.; Wilson, C.K.

    1989-01-01

    There is a compelling need for the transport of radioactive material by air because of the requirement by hospitals throughout the world for urgent delivery for medical purposes. Many countries have no radionuclide-producing capabilities and depend on imports: a range of such products is supplied from the United Kingdom. Many of these are short lived, which explains the need for urgent delivery. The only satisfactory method of delivery on a particular day to a particular destination is often by the use of scheduled passenger air service. The International Civil Aviation Organization's Technical Instructions for the Safe Transport of Dangerous Goods by Air (ICAO 1987-1988), prescribe the detailed requirements applicable to the international transport of dangerous goods by air. Radioactive materials are required to be separated from persons and from undeveloped photographic films or plates: minimum distances as a function of the total sum of transport indexes are given in the Instructions. A study, which included the measurement and assessment of the radiation doses resulting from the transport of radioactive materials by air from the UK, has been performed by the National Radiological Protection Board (NRPB) on behalf of the Civil Aviation Authority (CAA) and the Department of Transport (DTp)

  14. Design of tallying function for general purpose Monte Carlo particle transport code JMCT

    International Nuclear Information System (INIS)

    Shangguan Danhua; Li Gang; Deng Li; Zhang Baoyin

    2013-01-01

    A new postponed accumulation algorithm was proposed. Based on JCOGIN (J combinatorial geometry Monte Carlo transport infrastructure) framework and the postponed accumulation algorithm, the tallying function of the general purpose Monte Carlo neutron-photon transport code JMCT was improved markedly. JMCT gets a higher tallying efficiency than MCNP 4C by 28% for simple geometry model, and JMCT is faster than MCNP 4C by two orders of magnitude for complicated repeated structure model. The available ability of tallying function for JMCT makes firm foundation for reactor analysis and multi-step burnup calculation. (authors)

  15. 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.

  16. Draft ASME code case on ductile cast iron for transport packaging

    International Nuclear Information System (INIS)

    Saegusa, T.; Arai, T.; Hirose, M.; Kobayashi, T.; Tezuka, Y.; Urabe, N.; Hueggenberg, R.

    2004-01-01

    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

  17. Computer codes for three dimensional mass transport with non-linear sorption

    International Nuclear Information System (INIS)

    Noy, D.J.

    1985-03-01

    The report describes the mathematical background and data input to finite element programs for three dimensional mass transport in a porous medium. The transport equations are developed and sorption processes are included in a general way so that non-linear equilibrium relations can be introduced. The programs are described and a guide given to the construction of the required input data sets. Concluding remarks indicate that the calculations require substantial computer resources and suggest that comprehensive preliminary analysis with lower dimensional codes would be important in the assessment of field data. (author)

  18. Application of the ASME code in designing containment vessels for packages used to transport radioactive materials

    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

  19. Analytical Radiation Transport Benchmarks for The Next Century

    International Nuclear Information System (INIS)

    Ganapol, B.D.

    2005-01-01

    Verification of large-scale computational algorithms used in nuclear engineering and radiological applications is an essential element of reliable code performance. For this reason, the development of a suite of multidimensional semi-analytical benchmarks has been undertaken to provide independent verification of proper operation of codes dealing with the transport of neutral particles. The benchmarks considered cover several one-dimensional, multidimensional, monoenergetic and multigroup, fixed source and critical transport scenarios. The first approach, called the Green's Function. In slab geometry, the Green's function is incorporated into a set of integral equations for the boundary fluxes. Through a numerical Fourier transform inversion and subsequent matrix inversion for the boundary fluxes, a semi-analytical benchmark emerges. Multidimensional solutions in a variety of infinite media are also based on the slab Green's function. In a second approach, a new converged SN method is developed. In this method, the SN solution is ''minded'' to bring out hidden high quality solutions. For this case multigroup fixed source and criticality transport problems are considered. Remarkably accurate solutions can be obtained with this new method called the Multigroup Converged SN (MGCSN) method as will be demonstrated

  20. Calibration of neutron yield activation measurements at JET using MCNP and furnace neutron transport codes

    International Nuclear Information System (INIS)

    Pillon, M.; Martone, M.; Verschuur, K.A.; Jarvis, O.N.; Kaellne, J.

    1989-01-01

    Neutron transport calculations have been performed using fluence ray tracing (FURNACE code) and Monte Carlo particle trajectory sampling methods (MCNP code) in order to determine the neutron fluence and energy distributions at different locations in the JET tokamak. These calculations were used to calibrate the activation measurements used in the determination of the absolute fusion neutron yields from the JET plasma. We present here the neutron activation response coefficients calculated for three different materials. Comparison of the MCNP and FURNACE results helps identify the sources of error in these neutron transport calculations. The accuracy of these calculations was tested by comparing the total 2.5 MeV neutron yields derived from the activation measurements with those obtained with calibrated fission chambers; agreement at the ±15% level was demonstrate. (orig.)

  1. Sensitivity analysis of a low-level waste environmental transport code

    International Nuclear Information System (INIS)

    Hiromoto, G.

    1989-01-01

    Results are presented from a sensivity analysis of a computer code designed to simulate the environmental transport of radionuclides buried at shallow land waste repositories. A sensitivity analysis methodology, based on the surface response replacement and statistic sensitivity estimators, was developed to address the relative importance of the input parameters on the model output. Response surface replacement for the model was constructed by stepwise regression, after sampling input vectors from range and distribution of the input variables, and running the code to generate the associated output data. Sensitivity estimators were compute using the partial rank correlation coefficients and the standardized rank regression coefficients. The results showed that the tecniques employed in this work provides a feasible means to perform a sensitivity analysis of a general not-linear environmental radionuclides transport models. (author) [pt

  2. Atmospheric transport, clouds and the Arctic longwave radiation paradox

    Science.gov (United States)

    Sedlar, Joseph

    2016-04-01

    Clouds interact with radiation, causing variations in the amount of electromagnetic energy reaching the Earth's surface, or escaping the climate system to space. While globally clouds lead to an overall cooling radiative effect at the surface, over the Arctic, where annual cloud fractions are high, the surface cloud radiative effect generally results in a warming. The additional energy input from absorption and re-emission of longwave radiation by the clouds to the surface can have a profound effect on the sea ice state. Anomalous atmospheric transport of heat and moisture into the Arctic, promoting cloud formation and enhancing surface longwave radiation anomalies, has been identified as an important mechanism in preconditioning Arctic sea ice for melt. Longwave radiation is emitted equally in all directions, and changes in the atmospheric infrared emission temperature and emissivity associated with advection of heat and moisture over the Arctic should correspondingly lead to an anomalous signal in longwave radiation at the top of the atmosphere (TOA). To examine the role of atmospheric heat and moisture transport into the Arctic on TOA longwave radiation, infrared satellite sounder observations from AIRS during 2003-2014 are analyzed for summer (JJAS). Thermodynamic metrics are developed to identify months characterized by a high frequency of warm and moist advection into the Arctic, and segregate the 2003-14 time period into climatological and anomalously warm, moist summer months. We find that anomalously warm, moist months result in a significant TOA longwave radiative cooling, which is opposite the forcing signal that the surface experiences during these months. At the timescale of the advective events, 3-10 days, the TOA cooling can be as large as the net surface energy budget during summer. When averaged on the monthly time scale, and over the full Arctic basin (poleward of 75°N), summer months experiencing frequent warm, moist advection events are

  3. Monte Carlo analysis of radiative transport in oceanographic lidar measurements

    Energy Technology Data Exchange (ETDEWEB)

    Cupini, E.; Ferro, G. [ENEA, Divisione Fisica Applicata, Centro Ricerche Ezio Clementel, Bologna (Italy); Ferrari, N. [Bologna Univ., Bologna (Italy). Dipt. Ingegneria Energetica, Nucleare e del Controllo Ambientale

    2001-07-01

    The analysis of oceanographic lidar systems measurements is often carried out with semi-empirical methods, since there is only a rough understanding of the effects of many environmental variables. The development of techniques for interpreting the accuracy of lidar measurements is needed to evaluate the effects of various environmental situations, as well as of different experimental geometric configurations and boundary conditions. A Monte Carlo simulation model represents a tool that is particularly well suited for answering these important questions. The PREMAR-2F Monte Carlo code has been developed taking into account the main molecular and non-molecular components of the marine environment. The laser radiation interaction processes of diffusion, re-emission, refraction and absorption are treated. In particular are considered: the Rayleigh elastic scattering, produced by atoms and molecules with small dimensions with respect to the laser emission wavelength (i.e. water molecules), the Mie elastic scattering, arising from atoms or molecules with dimensions comparable to the laser wavelength (hydrosols), the Raman inelastic scattering, typical of water, the absorption of water, inorganic (sediments) and organic (phytoplankton and CDOM) hydrosols, the fluorescence re-emission of chlorophyll and yellow substances. PREMAR-2F is an extension of a code for the simulation of the radiative transport in atmospheric environments (PREMAR-2). The approach followed in PREMAR-2 was to combine conventional Monte Carlo techniques with analytical estimates of the probability of the receiver to have a contribution from photons coming back after an interaction in the field of view of the lidar fluorosensor collecting apparatus. This offers an effective mean for modelling a lidar system with realistic geometric constraints. The retrieved semianalytic Monte Carlo radiative transfer model has been developed in the frame of the Italian Research Program for Antarctica (PNRA) and it is

  4. Development of computer code on sodium-water reaction products transport

    International Nuclear Information System (INIS)

    Arikawa, H.; Yoshioka, N.; Suemori, M.; Nishida, K.

    1988-01-01

    The LMFBR concept eliminating the secondary sodium system has been considered to be one of the most promissing concepts for offering cost reductions. In this reactor concept, the evaluation of effects on reactor core by the sodium-water reaction products (SWRPs) during sodium-water reaction at primary steam generator becomes one of the major safety issues. In this study, the calculation code was developed as the first step of the processes of establishing the evaluation method for SWRP effects. The calculation code, called SPROUT, simulates the SWRPs transport and distribution in primary sodium system using the system geometry, thermal hydraulic data and sodium-water reacting conditions as input. This code principally models SWRPs behavior. The paper contain the modelings for SWRPs behaviors, with solution, precipation, deposition and so on, and the results and discussions of the demonstration calculation for a typical FBR plant eliminating the secondary sodium system

  5. MAX: an expert system for running the modular transport code APOLLO II

    International Nuclear Information System (INIS)

    Loussouarn, O.; Ferraris, C.; Boivineau, A.

    1990-01-01

    MAX is an expert system built to help users of the APOLLO II code to prepare the input data deck to run a job. APOLLO II is a modular transport-theory code for calculating the neutron flux in various geometries. The associated GIBIANE command language allows the user to specify the physical structure and the computational method to be used in the calculation. The purpose of MAX is to bring into play expertise in both neutronic and computing aspects of the code, as well as various computational schemes, in order to generate automatically a batch data set corresponding to the APOLLO II calculation desired by the user. MAX is implemented on the SUN 3/60 workstation with the S1 tool and graphic interface external functions

  6. Collection of regulatory texts relative to radiation protection. Part 2: orders and decisions taken in application of the Public Health Code and Labour Code concerning the protection of populations, patients and workers against the risks of ionizing radiations

    International Nuclear Information System (INIS)

    2007-05-01

    This collection of texts includes the general measures of population protection, exposure to natural radiations, general system of authorizations and statements, protection of persons exposed to ionizing radiations for medical purpose, situations of radiological emergency and long exposure to ionizing radiations, penal dispositions, application of the Public Health code and application of the Labour code. Chronological contents by date of publication is given. (N.C.)

  7. 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

  8. RIVER-RAD: A computer code for simulating the transport of radionuclides in rivers

    International Nuclear Information System (INIS)

    Hetrick, D.M.; McDowell-Boyer, L.M.; Sjoreen, A.L.; Thorne, D.J.; Patterson, M.R.

    1992-11-01

    A screening-level model, RIVER-RAD, has been developed to assess the potential fate of radionuclides released to rivers. The model is simplified in nature and is intended to provide guidance in determining the potential importance of the surface water pathway, relevant transport mechanisms, and key radionuclides in estimating radiological dose to man. The purpose of this report is to provide a description of the model and a user's manual for the FORTRAN computer code

  9. Intercomparison of three microwave/infrared high resolution line-by-line radiative transfer codes

    Science.gov (United States)

    Schreier, Franz; Milz, Mathias; Buehler, Stefan A.; von Clarmann, Thomas

    2018-05-01

    An intercomparison of three line-by-line (lbl) codes developed independently for atmospheric radiative transfer and remote sensing - ARTS, GARLIC, and KOPRA - has been performed for a thermal infrared nadir sounding application assuming a HIRS-like (High resolution Infrared Radiation Sounder) setup. Radiances for the 19 HIRS infrared channels and a set of 42 atmospheric profiles from the "Garand dataset" have been computed. The mutual differences of the equivalent brightness temperatures are presented and possible causes of disagreement are discussed. In particular, the impact of path integration schemes and atmospheric layer discretization is assessed. When the continuum absorption contribution is ignored because of the different implementations, residuals are generally in the sub-Kelvin range and smaller than 0.1 K for some window channels (and all atmospheric models and lbl codes). None of the three codes turned out to be perfect for all channels and atmospheres. Remaining discrepancies are attributed to different lbl optimization techniques. Lbl codes seem to have reached a maturity in the implementation of radiative transfer that the choice of the underlying physical models (line shape models, continua etc) becomes increasingly relevant.

  10. Load balancing in highly parallel processing of Monte Carlo code for particle transport

    International Nuclear Information System (INIS)

    Higuchi, Kenji; Takemiya, Hiroshi; Kawasaki, Takuji

    1998-01-01

    In parallel processing of Monte Carlo (MC) codes for neutron, photon and electron transport problems, particle histories are assigned to processors making use of independency of the calculation for each particle. Although we can easily parallelize main part of a MC code by this method, it is necessary and practically difficult to optimize the code concerning load balancing in order to attain high speedup ratio in highly parallel processing. In fact, the speedup ratio in the case of 128 processors remains in nearly one hundred times when using the test bed for the performance evaluation. Through the parallel processing of the MCNP code, which is widely used in the nuclear field, it is shown that it is difficult to attain high performance by static load balancing in especially neutron transport problems, and a load balancing method, which dynamically changes the number of assigned particles minimizing the sum of the computational and communication costs, overcomes the difficulty, resulting in nearly fifteen percentage of reduction for execution time. (author)

  11. Review and assessment of thermodynamic and transport properties for the CONTAIN Code

    International Nuclear Information System (INIS)

    Valdez, G.D.

    1988-12-01

    A study was carried out to review available data and correlations on the thermodynamic and transport properties of materials applicable to the CONTAIN computer code. CONTAIN is the NRC's best-estimate, mechanistic computer code for modeling containment response to a severe accident. Where appropriate, recommendations have been made for suitable approximations for material properties of interests. Based on a modified Benedict-Webb-Rubin (BWR) equation of state, a procedure is introduced for calculating thermodynamic properties for common gases in the CONTAIN code. These gases are nitrogen, oxygen, hydrogen, carbon dioxide, carbon monoxide, steam, helium, and argon. The thermodynamic equations for density, currently represented in CONTAIN by relatively simple fits, were independently checked and are recommended to be replaced by the Lee-Kesler equation of state which substantially improves accuracy without too much sacrifice in computational efficiency. The accuracy of the calculated values have been found to be generally acceptable. Various correlations and models for single component gas transport properties, viscosity and thermal conductivity, were also assessed with available experimental data. When a suitable correlation or model was not available, transport properties were obtained by performing least-squares fit on experimental data. 50 refs., 126 figs., 3 tabs

  12. Integral transport computation of gamma detector response with the CPM2 code

    International Nuclear Information System (INIS)

    Jones, D.B.

    1989-12-01

    CPM-2 Version 3 is an enhanced version of the CPM-2 lattice physics computer code which supports the capabilities to (1) perform a two-dimensional gamma flux calculation and (2) perform Restart/Data file maintenance operations. The Gamma Calculation Module implemented in CPM-2 was first developed for EPRI in the CASMO-1 computer code by Studsvik Energiteknik under EPRI Agreement RP2352-01. The gamma transport calculation uses the CPM-HET code module to calculate the transport of gamma rays in two dimensions in a mixed cylindrical-rectangular geometry, where the basic fuel assembly and component regions are maintained in a rectangular geometry, but the fuel pins are represented as cylinders within a square pin cell mesh. Such a capability is needed to represent gamma transport in an essentially transparent medium containing spatially distributed ''black'' cylindrical pins. Under a subcontract to RP2352-01, RPI developed the gamma production and gamma interaction library used for gamma calculation. The CPM-2 gamma calculation was verified against reference results generated by Studsvik using the CASMO-1 program. The CPM-2 Restart/Data file maintenance capabilities provide the user with options to copy files between Restart/Data tapes and to purge files from the Restart/Data tapes

  13. Radioactive materials' transportation main routes in Brazil. Radiation protection aspects about radioactive materials transportation

    International Nuclear Information System (INIS)

    Vaz, Solange dos Reis e; Andrade, Fernando de Menezes; Aleixo, Luiz Claudio Martins

    2007-01-01

    The heavy transportation in Brazil is generally done by highways. The radioactive material transportation follow this same rule. Whenever a radioactive material is carried by the road, by the sea or by the air, in some cases, a kind of combination of those transportation ways, the transport manager has to create a Transportation Plan and submit it to CNEN. Only after CNEN's approval, the transportation can be done. The plan must have the main action on Radiation Protection, giving responsibilities and showing all the directing that will be take. Although, the Brazilian's highways are not in good conditions, one could say that some of them are not good enough for any kind of transportation. But we are facing radioactive material use increase but the hospitals and industries, that the reason it's much more common that kind of transportation nowadays. So, because of that, a special attention by the governments must be provide to those activities. This paper goal is to show the real conditions of some important highways in Brazil in a radioactive protection's perspective and give some suggestions to adjust some of those roads to this new reality. (author)

  14. A Radiation Chemistry Code Based on the Green's Function of the Diffusion Equation

    Science.gov (United States)

    Plante, Ianik; Wu, Honglu

    2014-01-01

    Stochastic radiation track structure codes are of great interest for space radiation studies and hadron therapy in medicine. These codes are used for a many purposes, notably for microdosimetry and DNA damage studies. In the last two decades, they were also used with the Independent Reaction Times (IRT) method in the simulation of chemical reactions, to calculate the yield of various radiolytic species produced during the radiolysis of water and in chemical dosimeters. Recently, we have developed a Green's function based code to simulate reversible chemical reactions with an intermediate state, which yielded results in excellent agreement with those obtained by using the IRT method. This code was also used to simulate and the interaction of particles with membrane receptors. We are in the process of including this program for use with the Monte-Carlo track structure code Relativistic Ion Tracks (RITRACKS). This recent addition should greatly expand the capabilities of RITRACKS, notably to simulate DNA damage by both the direct and indirect effect.

  15. Self-shielding phenomenon modelling in multigroup transport code Apollo-2; Modelisation du phenomene d'autoprotection dans le code de transport multigroupe Apollo 2

    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)

  16. Self-shielding phenomenon modelling in multigroup transport code Apollo-2; Modelisation du phenomene d'autoprotection dans le code de transport multigroupe Apollo 2

    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)

  17. Analytical three-dimensional neutron transport benchmarks for verification of nuclear engineering codes. Final report

    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

  18. Radiation protection of workers in the mining and milling of radioactive ores. Code of practice and technical addendum. 1983 ed

    International Nuclear Information System (INIS)

    1983-01-01

    This Code of Practice sets forth the means of ensuring protection against ionizing radiation for workers engaged in mining and milling of radioactive ores: general provisions outlining the responsibilities of the employer and the worker, limits of radiation exposure, administrative organization of radiation protection, radiation surveillance, engineering and administrative protective measures and medical surveillance. It is designed to facilitate the preparation and adoption of national and local regulations and factory rules for radiation protection in mining and milling of radioactive ores.

  19. Fast algorithm for two-dimensional data table use in hydrodynamic and radiative-transfer codes

    International Nuclear Information System (INIS)

    Slattery, W.L.; Spangenberg, W.H.

    1982-01-01

    A fast algorithm for finding interpolated atomic data in irregular two-dimensional tables with differing materials is described. The algorithm is tested in a hydrodynamic/radiative transfer code and shown to be of comparable speed to interpolation in regularly spaced tables, which require no table search. The concepts presented are expected to have application in any situation with irregular vector lengths. Also, the procedures that were rejected either because they were too slow or because they involved too much assembly coding are described

  20. User manual for version 4.3 of the Tripoli-4 Monte-Carlo method particle transport computer code; Notice d'utilisation du code Tripoli-4, version 4.3: code de transport de particules par la methode de Monte Carlo

    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)

  1. Comparative study of Monte Carlo particle transport code PHITS and nuclear data processing code NJOY for PKA energy spectra and heating number under neutron irradiation

    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

  2. Comparative study of Monte Carlo particle transport code PHITS and nuclear data processing code NJOY for recoil cross section spectra under neutron irradiation

    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.

  3. 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)

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

    International Nuclear Information System (INIS)

    Baes, C.F. III; Sharp, R.D.; Sjoreen, A.L.; Hermann, O.W.

    1984-11-01

    TERRA is a computer code which calculates concentrations of radionuclides and ingrowing daughters in surface and root-zone soil, produce and feed, beef, and milk from a given deposition rate at any location in the conterminous United States. The code is fully integrated with seven other computer codes which together comprise a Computerized Radiological Risk Investigation System, CRRIS. Output from either the long range (> 100 km) atmospheric dispersion code RETADD-II or the short range (<80 km) atmospheric dispersion code ANEMOS, in the form of radionuclide air concentrations and ground deposition rates by downwind location, serves as input to TERRA. User-defined deposition rates and air concentrations may also be provided as input to TERRA through use of the PRIMUS computer code. The environmental concentrations of radionuclides predicted by TERRA serve as input to the ANDROS computer code which calculates population and individual intakes, exposures, doses, and risks. TERRA incorporates models to calculate uptake from soil and atmospheric deposition on four groups of produce for human consumption and four groups of livestock feeds. During the environmental transport simulation, intermediate calculations of interception fraction for leafy vegetables, produce directly exposed to atmospherically depositing material, pasture, hay, and silage are made based on location-specific estimates of standing crop biomass. Pasture productivity is estimated by a model which considers the number and types of cattle and sheep, pasture area, and annual production of other forages (hay and silage) at a given location. Calculations are made of the fraction of grain imported from outside the assessment area. TERRA output includes the above calculations and estimated radionuclide concentrations in plant produce, milk, and a beef composite by location

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

    Energy Technology Data Exchange (ETDEWEB)

    Baes, C.F. III; Sharp, R.D.; Sjoreen, A.L.; Hermann, O.W.

    1984-11-01

    TERRA is a computer code which calculates concentrations of radionuclides and ingrowing daughters in surface and root-zone soil, produce and feed, beef, and milk from a given deposition rate at any location in the conterminous United States. The code is fully integrated with seven other computer codes which together comprise a Computerized Radiological Risk Investigation System, CRRIS. Output from either the long range (> 100 km) atmospheric dispersion code RETADD-II or the short range (<80 km) atmospheric dispersion code ANEMOS, in the form of radionuclide air concentrations and ground deposition rates by downwind location, serves as input to TERRA. User-defined deposition rates and air concentrations may also be provided as input to TERRA through use of the PRIMUS computer code. The environmental concentrations of radionuclides predicted by TERRA serve as input to the ANDROS computer code which calculates population and individual intakes, exposures, doses, and risks. TERRA incorporates models to calculate uptake from soil and atmospheric deposition on four groups of produce for human consumption and four groups of livestock feeds. During the environmental transport simulation, intermediate calculations of interception fraction for leafy vegetables, produce directly exposed to atmospherically depositing material, pasture, hay, and silage are made based on location-specific estimates of standing crop biomass. Pasture productivity is estimated by a model which considers the number and types of cattle and sheep, pasture area, and annual production of other forages (hay and silage) at a given location. Calculations are made of the fraction of grain imported from outside the assessment area. TERRA output includes the above calculations and estimated radionuclide concentrations in plant produce, milk, and a beef composite by location.

  6. ZZ SAIL, Albedo Scattering Data Library for 3-D Monte-Carlo Radiation Transport in LWR Pressure Vessel

    International Nuclear Information System (INIS)

    1982-01-01

    1 - Description of problem or function: Format: SAIL format; Number of groups: 23 neutron / 17 gamma-ray; Nuclides: Type 04 Concrete and Low Carbon Steel (A533B). Origin: Science Applications, Inc (SAI); Weighting spectrum: yes. SAIL is a library of albedo scattering data to be used in three-dimensional Monte Carlo codes to solve radiation transport problems specific to the reactor pressure vessel cavity region of a LWR. The library contains data for Type 04 Concrete and Low Carbon Steel (A533B). 2 - Method of solution: The calculation of the albedo data was perform- ed with a version of the discrete ordinates transport code DOT which treats the transport of neutrons, secondary gamma-rays and gamma- rays in one dimension, while maintaining the complete two-dimension- al treatment of the angular dependence

  7. Fission product transport in the primary system, important phenomena, and code status

    International Nuclear Information System (INIS)

    Gieseke, J.A.; Jordan, H.; Kuhlman, M.R.

    1990-01-01

    The purpose of this paper is to identify important issues concerning the transport and deposition of radionuclides in the reactor coolant system (RCS) under accident conditions and to examine how such issues are being treated or should be treated by the various available computer codes. In general, the RCS is a very important section of the transport pathway along which radionuclides move and by which they are attenuated as they travel after being released from the fuel. The RCS can serve as a sink for radionuclides that may deposit from the gas and react with surfaces, or can serve as a repository for materials deposited from the gas which are then available for later release into the transporting gas stream. The RCS may also have thermal hydraulic conditions that foster aerosol growth by condensation or agglomeration, and may provide an environment in which gas phase or heterogeneous chemical reactions may occur

  8. Radiation safety of crew and passengers of air transportation in civil aviation. Provisional standards

    Science.gov (United States)

    Aksenov, A. F.; Burnazyan, A. I.

    1985-01-01

    The purpose and application of the provisional standards for radiation safety of crew and passengers in civil aviation are given. The radiation effect of cosmic radiation in flight on civil aviation air transport is described. Standard levels of radiation and conditions of radiation safety are discussed.

  9. Estimated routine radiation doses to transportation workers in alternative spent-fuel transportation systems

    International Nuclear Information System (INIS)

    Schneider, K.J.; Smith, R.I.; Daling, P.M.; Ross, W.A.; McNair, G.W.

    1988-01-01

    The federal system for the management of spent fuel and high-level radioactive waste includes the acceptance by the US Department of Energy (DOE) of the spent fuel or waste loaded in casks at the reactor or other waste generators, its transportation to a repository, and its handling and final emplacement in the repository. The DOE plans to implement a transportation system that is safe, secure, efficient, and cost-effective and will meet applicable regulatory safety and security requirements. The DOE commissioned the Pacific Northwest Laboratory (PNL) to develop estimates of the routine radiation doses that would result from the operation of a system postulated using current designs and practices. From that evaluation, PNL identified activities/operations that result in the higher fraction of doses, proposed conceptual alternatives that would effectively reduce such exposures, and evaluated the cost-effectiveness of such alternatives. The study is one of a series used in making overall system design and operational decisions in the development of the DOE's spent-fuel/high-level waste transportation system. This paper contains the highlights from the PNL study of the estimated radiation doses to the transportation workers in a postulated reference transportation system and potential alternatives to that system

  10. Modeling radiative transport in ICF plasmas on an IBM SP2 supercomputer

    International Nuclear Information System (INIS)

    Johansen, J.A.; MacFarlane, J.J.; Moses, G.A.

    1995-01-01

    At the University of Wisconsin-Madison the authors have integrated a collisional-radiative-equilibrium model into their CONRAD radiation-hydrodynamics code. This integrated package allows them to accurately simulate the transport processes involved in ICF plasmas; including the important effects of self-absorption of line-radiation. However, as they increase the amount of atomic structure utilized in their transport models, the computational demands increase nonlinearly. In an attempt to meet this increased computational demand, they have recently embarked on a mission to parallelize the CONRAD program. The parallel CONRAD development is being performed on an IBM SP2 supercomputer. The parallelism is based on a message passing paradigm, and is being implemented using PVM. At the present time they have determined that approximately 70% of the sequential program can be executed in parallel. Accordingly, they expect that the parallel version will yield a speedup on the order of three times that of the sequential version. This translates into only 10 hours of execution time for the parallel version, whereas the sequential version required 30 hours

  11. Code-To-Code Benchmarking Of The Porflow And GoldSim Contaminant Transport Models Using A Simple 1-D Domain - 11191

    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

  12. Extensions of the 3-dimensional plasma transport code E3D

    International Nuclear Information System (INIS)

    Runov, A.; Schneider, R.; Kasilov, S.; Reiter, D.

    2004-01-01

    One important aspect of modern fusion research is plasma edge physics. Fluid transport codes extending beyond the standard 2-D code packages like B2-Eirene or UEDGE are under development. A 3-dimensional plasma fluid code, E3D, based upon the Multiple Coordinate System Approach and a Monte Carlo integration procedure has been developed for general magnetic configurations including ergodic regions. These local magnetic coordinates lead to a full metric tensor which accurately accounts for all transport terms in the equations. Here, we discuss new computational aspects of the realization of the algorithm. The main limitation to the Monte Carlo code efficiency comes from the restriction on the parallel jump of advancing test particles which must be small compared to the gradient length of the diffusion coefficient. In our problems, the parallel diffusion coefficient depends on both plasma and magnetic field parameters. Usually, the second dependence is much more critical. In order to allow long parallel jumps, this dependence can be eliminated in two steps: first, the longitudinal coordinate x 3 of local magnetic coordinates is modified in such a way that in the new coordinate system the metric determinant and contra-variant components of the magnetic field scale along the magnetic field with powers of the magnetic field module (like in Boozer flux coordinates). Second, specific weights of the test particles are introduced. As a result of increased parallel jump length, the efficiency of the code is about two orders of magnitude better. (copyright 2004 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  13. Towards A Genetic Business Code For Growth in the South African Transport Industry

    Directory of Open Access Journals (Sweden)

    J.H. Vermeulen

    2003-11-01

    Full Text Available As with each living organism, it is proposed that an organisation possesses a genetic code. In the fast-changing business environment it would be invaluable to know what constitutes organisational growth and success in terms of such a code. To identify this genetic code a quantitative methodological framework, supplemented by a qualitative approach, was used and the views of top management in the Transport Industry were solicited. The Repertory Grid was used as the primary data-collection method. Through a phased data-analysis process an integrated profile of first- and second-order constructs, and opposite poles, was compiled. By utilising deductive and inductive strategies three strands of a Genetic Business Growth Code were identified, namely a Leadership Strand, Organisational Architecture Strand and Internal Orientation Strand. The study confirmed the value of a Genetic Business Code for growth in the Transport Industry. Opsomming Daar word voorgestel dat ’n organisasie, soos elke lewende organisme, oor ’n genetiese kode beskik. In die snelveranderende sake-omgewing sal dit onskatbaar wees om te weet wat organisasiegroei en –sukses veroorsaak. ’n Kwantitatiewe metodologie-raamwerk, aangevul deur ’n kwalitatiewe benadering is gebruik om hierdie genetiese kode te identifiseer, en die menings van topbestuur in die Vervoerbedryf is ingewin met behulp van die “Repertory Grid" as die vernaamste metode van data-insameling. ’n Geïntegreerde profiel van eerste- en tweedeordekonstrukte, met hulle teenoorgestelde pole, is opgestel. Drie stringe van ’n Genetiese Sakegroeikode, nl. ’n Leierskapstring, die Organisasieargitektuur-string en die Innerlike-ingesteldheidstring is geïdentifiseer deur deduktiewe en induktiewe strategieë te gebruik. Die studie bevestig die waarde van ’n Genetiese Sakekode vir groei in die Vervoerbedryf.

  14. 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

  15. 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

  16. Simulation of radiation transport using MCNP for a teletherapy machine; Simulacion del transporte de radiacion usando MCNP para una maquina de teleterapia

    Energy Technology Data Exchange (ETDEWEB)

    Flores O, F.E.; Mireles G, F.; Davila R, J.I.; Pinedo V, J.L.; Risorios M, C.; Lopez del Rio, H. [UAZ, Unidad Academica de Estudios Nucleares, 98068 Zacatecas (Mexico)

    2008-07-01

    The MCNP code is used to simulate the radiation transport taking as tools the transport physics of each particle, either photon, neutron or electron, and the generation of random numbers. Developed in the Los Alamos National Laboratory, this code has been used thoroughly with great success, because the results of the simulations are broadly validated with representative experiments. In the one present work the room of radiotherapy of the Institute Zacatecano of the Tumor it is simulated, located in the city of Zacatecas where one is Theratron 780C machine manufactured by MSD Nordion, with the purpose of estimating the contribution to the dose that would be received in different points of the structure, included three directly under the source. Three results of analytical calculations for points located at different distances from the source are presented, and they are compared against those obtained by the simulation. Its are also presented results for the simulation of 10 points more distributed around the source. (Author)

  17. A FORMALISM FOR COVARIANT POLARIZED RADIATIVE TRANSPORT BY RAY TRACING

    International Nuclear Information System (INIS)

    Gammie, Charles F.; Leung, Po Kin

    2012-01-01

    We write down a covariant formalism for polarized radiative transfer appropriate for ray tracing through a turbulent plasma. The polarized radiation field is represented by the polarization tensor (coherency matrix) N αβ ≡ (a α k a* β k ), where a k is a Fourier coefficient for the vector potential. Using Maxwell's equations, the Liouville-Vlasov equation, and the WKB approximation, we show that the transport equation in vacuo is k μ ∇ μ N αβ = 0. We show that this is equivalent to Broderick and Blandford's formalism based on invariant Stokes parameters and a rotation coefficient, and suggest a modification that may reduce truncation error in some situations. Finally, we write down several alternative approaches to integrating the transfer equation.

  18. Radiation transport in high-level waste form

    International Nuclear Information System (INIS)

    Arakali, V.S.; Barnes, S.M.

    1992-01-01

    The waste form selected for vitrifying high-level nuclear waste stored in underground tanks at West Valley, NY is borosilicate glass. The maximum radiation level at the surface of a canister filled with the high-level waste form is prescribed by repository design criteria for handling and disposition of the vitrified waste. This paper presents an evaluation of the radiation transport characteristics for the vitreous waste form expected to be produced at West Valley and the resulting neutron and gamma dose rates. The maximum gamma and neutron dose rates are estimated to be less than 7500 R/h and 10 mRem/h respectively at the surface of a West Valley canister filled with borosilicate waste glass

  19. Compendium of Material Composition Data for Radiation Transport Modeling

    International Nuclear Information System (INIS)

    Williams, Ralph G.; Gesh, Christopher J.; Pagh, Richard T.

    2006-01-01

    Computational modeling of radiation transport problems including homeland security, radiation shielding and protection, and criticality safety all depend upon material definitions. This document has been created to serve two purposes: (1) to provide a quick reference of material compositions for analysts and (2) a standardized reference to reduce the differences between results from two independent analysts. Analysts are always encountering a variety of materials for which elemental definitions are not readily available or densities are not defined. This document provides a location where unique or hard to define materials will be located to reduce duplication in research for modeling purposes. Additionally, having a common set of material definitions helps to standardize modeling across PNNL and provide two separate researchers the ability to compare different modeling results from a common materials basis.

  20. The Intercomparison of 3D Radiation Codes (I3RC): Showcasing Mathematical and Computational Physics in a Critical Atmospheric Application

    Science.gov (United States)

    Davis, A. B.; Cahalan, R. F.

    2001-05-01

    The Intercomparison of 3D Radiation Codes (I3RC) is an on-going initiative involving an international group of over 30 researchers engaged in the numerical modeling of three-dimensional radiative transfer as applied to clouds. Because of their strong variability and extreme opacity, clouds are indeed a major source of uncertainty in the Earth's local radiation budget (at GCM grid scales). Also 3D effects (at satellite pixel scales) invalidate the standard plane-parallel assumption made in the routine of cloud-property remote sensing at NASA and NOAA. Accordingly, the test-cases used in I3RC are based on inputs and outputs which relate to cloud effects in atmospheric heating rates and in real-world remote sensing geometries. The main objectives of I3RC are to (1) enable participants to improve their models, (2) publish results as a community, (3) archive source code, and (4) educate. We will survey the status of I3RC and its plans for the near future with a special emphasis on the mathematical models and computational approaches. We will also describe some of the prime applications of I3RC's efforts in climate models, cloud-resolving models, and remote-sensing observations of clouds, or that of the surface in their presence. In all these application areas, computational efficiency is the main concern and not accuracy. One of I3RC's main goals is to document the performance of as wide a variety as possible of three-dimensional radiative transfer models for a small but representative number of ``cases.'' However, it is dominated by modelers working at the level of linear transport theory (i.e., they solve the radiative transfer equation) and an overwhelming majority of these participants use slow-but-robust Monte Carlo techniques. This means that only a small portion of the efficiency vs. accuracy vs. flexibility domain is currently populated by I3RC participants. To balance this natural clustering the present authors have organized a systematic outreach towards