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Sample records for radiation transport calculation

  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. 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.

  3. 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

  4. 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...

  5. 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)

  6. Volume-based geometric modeling for radiation transport calculations

    International Nuclear Information System (INIS)

    Li, Z.; Williamson, J.F.

    1992-01-01

    Accurate theoretical characterization of radiation fields is a valuable tool in the design of complex systems, such as linac heads and intracavitary applicators, and for generation of basic dose calculation data that is inaccessible to experimental measurement. Both Monte Carlo and deterministic solutions to such problems require a system for accurately modeling complex 3-D geometries that supports ray tracing, point and segment classification, and 2-D graphical representation. Previous combinatorial approaches to solid modeling, which involve describing complex structures as set-theoretic combinations of simple objects, are limited in their ease of use and place unrealistic constraints on the geometric relations between objects such as excluding common boundaries. A new approach to volume-based solid modeling has been developed which is based upon topologically consistent definitions of boundary, interior, and exterior of a region. From these definitions, FORTRAN union, intersection, and difference routines have been developed that allow involuted and deeply nested structures to be described as set-theoretic combinations of ellipsoids, elliptic cylinders, prisms, cones, and planes that accommodate shared boundaries. Line segments between adjacent intersections on a trajectory are assigned to the appropriate region by a novel sorting algorithm that generalizes upon Siddon's approach. Two 2-D graphic display tools are developed to help the debugging of a given geometric model. In this paper, the mathematical basis of our system is described, it is contrasted to other approaches, and examples are discussed

  7. Goal based mesh adaptivity for fixed source radiation transport calculations

    International Nuclear Information System (INIS)

    Baker, C.M.J.; Buchan, A.G.; Pain, C.C.; Tollit, B.S.; Goffin, M.A.; Merton, S.R.; Warner, P.

    2013-01-01

    Highlights: ► Derives an anisotropic goal based error measure for shielding problems. ► Reduces the error in the detector response by optimizing the finite element mesh. ► Anisotropic adaptivity captures material interfaces using fewer elements than AMR. ► A new residual based on the numerical scheme chosen forms the error measure. ► The error measure also combines the forward and adjoint metrics in a novel way. - Abstract: In this paper, the application of goal based error measures for anisotropic adaptivity applied to shielding problems in which a detector is present is explored. Goal based adaptivity is important when the response of a detector is required to ensure that dose limits are adhered to. To achieve this, a dual (adjoint) problem is solved which solves the neutron transport equation in terms of the response variables, in this case the detector response. The methods presented can be applied to general finite element solvers, however, the derivation of the residuals are dependent on the underlying finite element scheme which is also discussed in this paper. Once error metrics for the forward and adjoint solutions have been formed they are combined using a novel approach. The two metrics are combined by forming the minimum ellipsoid that covers both the error metrics rather than taking the maximum ellipsoid that is contained within the metrics. Another novel approach used within this paper is the construction of the residual. The residual, used to form the goal based error metrics, is calculated from the subgrid scale correction which is inherent in the underlying spatial discretisation employed

  8. 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)

  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. Comparison of the results of radiation transport calculation obtained by means of different programs

    International Nuclear Information System (INIS)

    Gorbatkov, D.V.; Kruchkov, V.P.

    1995-01-01

    Verification of calculational results of radiation transport, obtained by the known, programs and constant libraries (MCNP+ENDF/B, ANISN+HILO, FLUKA92) by means of their comparison with the precision results calculations through ROZ-6N+Sadko program constant complex and with experimental data, is carried out. Satisfactory agreement is shown with the MCNP+ENDF/B package data for the energy range of E<14 MeV. Analysis of the results derivations, obtained trough the ANISN-HILO package for E<400 MeV and the FLUKA92 programs of E<200 GeV is carried out. 25 refs., 12 figs., 3 tabs

  11. 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

  12. 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.)

  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. Toolkit for high performance Monte Carlo radiation transport and activation calculations for shielding applications in ITER

    International Nuclear Information System (INIS)

    Serikov, A.; Fischer, U.; Grosse, D.; Leichtle, D.; Majerle, M.

    2011-01-01

    The Monte Carlo (MC) method is the most suitable computational technique of radiation transport for shielding applications in fusion neutronics. This paper is intended for sharing the results of long term experience of the fusion neutronics group at Karlsruhe Institute of Technology (KIT) in radiation shielding calculations with the MCNP5 code for the ITER fusion reactor with emphasizing on the use of several ITER project-driven computer programs developed at KIT. Two of them, McCad and R2S, seem to be the most useful in radiation shielding analyses. The McCad computer graphical tool allows to perform automatic conversion of the MCNP models from the underlying CAD (CATIA) data files, while the R2S activation interface couples the MCNP radiation transport with the FISPACT activation allowing to estimate nuclear responses such as dose rate and nuclear heating after the ITER reactor shutdown. The cell-based R2S scheme was applied in shutdown photon dose analysis for the designing of the In-Vessel Viewing System (IVVS) and the Glow Discharge Cleaning (GDC) unit in ITER. Newly developed at KIT mesh-based R2S feature was successfully tested on the shutdown dose rate calculations for the upper port in the Neutral Beam (NB) cell of ITER. The merits of McCad graphical program were broadly acknowledged by the neutronic analysts and its continuous improvement at KIT has introduced its stable and more convenient run with its Graphical User Interface. Detailed 3D ITER neutronic modeling with the MCNP Monte Carlo method requires a lot of computation resources, inevitably leading to parallel calculations on clusters. Performance assessments of the MCNP5 parallel runs on the JUROPA/HPC-FF supercomputer cluster permitted to find the optimal number of processors for ITER-type runs. (author)

  15. Data structures and language elements for automated transport calculations for neutron and gamma radiation

    International Nuclear Information System (INIS)

    Rexer, G.

    1978-12-01

    Computer-aided design of nuclear shielding and irradiation facilities is characterized by studies of different design variants in order to determine which facilities are safe and still economicol. The design engineer has a very complex task including the formulation of calculation models, data linking of programs and data, and the management of large data stores. Integrated modular program systems with centralized module and data management make it possible to treat these problems in a more simplified and automatic manner. The paper describes a system of this type for the field of radiation transport and radiation shielding. The basis is the modular system RSYST II which has a dynamic hierarchical scheme for the structuring of problem data in a central data base. (orig./RW) [de

  16. Radiation transport calculations for the ANS [Advanced Neutron Source] beam tubes

    International Nuclear Information System (INIS)

    Engle, W.W. Jr.; Lillie, R.A.; Slater, C.O.

    1988-01-01

    The Advanced Neutron Source facility (ANS) will incorporate a large number of both radial and no-line-of-sight (NLS) beam tubes to provide very large thermal neutron fluxes to experimental facilities. The purpose of this work was to obtain comparisons for the ANS single- and split-core designs of the thermal and damage neutron and gamma-ray scalar fluxes in these beams tubes. For experimental locations far from the reactor cores, angular flux data are required; however, for close-in experimental locations, the scalar fluxes within each beam tube provide a credible estimate of the various signal to noise ratios. In this paper, the coupled two- and three-dimensional radiation transport calculations employed to estimate the scalar neutron and gamma-ray fluxes will be described and the results from these calculations will be discussed. 6 refs., 2 figs

  17. Description of a neutron field perturbed by a probe using coupled Monte Carlo and discrete ordinates radiation transport calculations

    International Nuclear Information System (INIS)

    Zazula, J.M.

    1984-01-01

    This work concerns calculation of a neutron response, caused by a neutron field perturbed by materials surrounding the source or the detector. Solution of a problem is obtained using coupling of the Monte Carlo radiation transport computation for the perturbed region and the discrete ordinates transport computation for the unperturbed system. (author). 62 refs

  18. Determination of peripheral underdosage at the lung-tumor interface using Monte Carlo radiation transport calculations

    International Nuclear Information System (INIS)

    Taylor, Michael; Dunn, Leon; Kron, Tomas; Height, Felicity; Franich, Rick

    2012-01-01

    Prediction of dose distributions in close proximity to interfaces is difficult. In the context of radiotherapy of lung tumors, this may affect the minimum dose received by lesions and is particularly important when prescribing dose to covering isodoses. The objective of this work is to quantify underdosage in key regions around a hypothetical target using Monte Carlo dose calculation methods, and to develop a factor for clinical estimation of such underdosage. A systematic set of calculations are undertaken using 2 Monte Carlo radiation transport codes (EGSnrc and GEANT4). Discrepancies in dose are determined for a number of parameters, including beam energy, tumor size, field size, and distance from chest wall. Calculations were performed for 1-mm 3 regions at proximal, distal, and lateral aspects of a spherical tumor, determined for a 6-MV and a 15-MV photon beam. The simulations indicate regions of tumor underdose at the tumor-lung interface. Results are presented as ratios of the dose at key peripheral regions to the dose at the center of the tumor, a point at which the treatment planning system (TPS) predicts the dose more reliably. Comparison with TPS data (pencil-beam convolution) indicates such underdosage would not have been predicted accurately in the clinic. We define a dose reduction factor (DRF) as the average of the dose in the periphery in the 6 cardinal directions divided by the central dose in the target, the mean of which is 0.97 and 0.95 for a 6-MV and 15-MV beam, respectively. The DRF can assist clinicians in the estimation of the magnitude of potential discrepancies between prescribed and delivered dose distributions as a function of tumor size and location. Calculation for a systematic set of “generic” tumors allows application to many classes of patient case, and is particularly useful for interpreting clinical trial data.

  19. Determination of peripheral underdosage at the lung-tumor interface using Monte Carlo radiation transport calculations

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, Michael, E-mail: michael.taylor@rmit.edu.au [School of Applied Sciences, College of Science, Engineering and Health, RMIT University, Melbourne, Victoria (Australia); Physical Sciences, Peter MacCallum Cancer Centre, East Melbourne, Victoria (Australia); Dunn, Leon; Kron, Tomas; Height, Felicity; Franich, Rick [School of Applied Sciences, College of Science, Engineering and Health, RMIT University, Melbourne, Victoria (Australia); Physical Sciences, Peter MacCallum Cancer Centre, East Melbourne, Victoria (Australia)

    2012-04-01

    Prediction of dose distributions in close proximity to interfaces is difficult. In the context of radiotherapy of lung tumors, this may affect the minimum dose received by lesions and is particularly important when prescribing dose to covering isodoses. The objective of this work is to quantify underdosage in key regions around a hypothetical target using Monte Carlo dose calculation methods, and to develop a factor for clinical estimation of such underdosage. A systematic set of calculations are undertaken using 2 Monte Carlo radiation transport codes (EGSnrc and GEANT4). Discrepancies in dose are determined for a number of parameters, including beam energy, tumor size, field size, and distance from chest wall. Calculations were performed for 1-mm{sup 3} regions at proximal, distal, and lateral aspects of a spherical tumor, determined for a 6-MV and a 15-MV photon beam. The simulations indicate regions of tumor underdose at the tumor-lung interface. Results are presented as ratios of the dose at key peripheral regions to the dose at the center of the tumor, a point at which the treatment planning system (TPS) predicts the dose more reliably. Comparison with TPS data (pencil-beam convolution) indicates such underdosage would not have been predicted accurately in the clinic. We define a dose reduction factor (DRF) as the average of the dose in the periphery in the 6 cardinal directions divided by the central dose in the target, the mean of which is 0.97 and 0.95 for a 6-MV and 15-MV beam, respectively. The DRF can assist clinicians in the estimation of the magnitude of potential discrepancies between prescribed and delivered dose distributions as a function of tumor size and location. Calculation for a systematic set of 'generic' tumors allows application to many classes of patient case, and is particularly useful for interpreting clinical trial data.

  20. The use of symbolic computation in radiative, energy, and neutron transport calculations

    Science.gov (United States)

    Frankel, J. I.

    This investigation uses symbolic computation in developing analytical methods and general computational strategies for solving both linear and nonlinear, regular and singular, integral and integro-differential equations which appear in radiative and combined mode energy transport. This technical report summarizes the research conducted during the first nine months of the present investigation. The use of Chebyshev polynomials augmented with symbolic computation has clearly been demonstrated in problems involving radiative (or neutron) transport, and mixed-mode energy transport. Theoretical issues related to convergence, errors, and accuracy have also been pursued. Three manuscripts have resulted from the funded research. These manuscripts have been submitted to archival journals. At the present time, an investigation involving a conductive and radiative medium is underway. The mathematical formulation leads to a system of nonlinear, weakly-singular integral equations involving the unknown temperature and various Legendre moments of the radiative intensity in a participating medium. Some preliminary results are presented illustrating the direction of the proposed research.

  1. Calculation of radiation effects in solids by direct numerical solution of the adjoint transport equation

    International Nuclear Information System (INIS)

    Matthes, W.K.

    1998-01-01

    The 'adjoint transport equation in its integro-differential form' is derived for the radiation damage produced by atoms injected into solids. We reduce it to the one-dimensional form and prepare it for a numerical solution by: --discretizing the continuous variables energy, space and direction, --replacing the partial differential quotients by finite differences and --evaluating the collision integral by a double sum. By a proper manipulation of this double sum the adjoint transport equation turns into a (very large) set of linear equations with tridiagonal matrix which can be solved by a special (simple and fast) algorithm. The solution of this set of linear equations contains complete information on a specified damage type (e.g. the energy deposited in a volume V) in terms of the function D(i,E,c,x) which gives the damage produced by all particles generated in a cascade initiated by a particle of type i starting at x with energy E in direction c. It is essential to remark that one calculation gives the damage function D for the complete ranges of the variables {i,E,c and x} (for numerical reasons of course on grid-points in the {E,c,x}-space). This is most useful to applications where a general source-distribution S(i,E,c,x) of particles is given by the experimental setup (e.g. beam-window and and target in proton accelerator work. The beam-protons along their path through the window--or target material generate recoil atoms by elastic collisions or nuclear reactions. These recoil atoms form the particle source S). The total damage produced then is eventually given by: D = (Σ)i ∫ ∫ ∫ S(i, E, c, x)*D(i, E, c, x)*dE*dc*dx A Fortran-77 program running on a PC-486 was written for the overall procedure and applied to some problems

  2. Dosimetric impact of Acuros XB deterministic radiation transport algorithm for heterogeneous dose calculation in lung cancer

    International Nuclear Information System (INIS)

    Han Tao; Followill, David; Repchak, Roman; Molineu, Andrea; Howell, Rebecca; Salehpour, Mohammad; Mikell, Justin; Mourtada, Firas

    2013-01-01

    Purpose: The novel deterministic radiation transport algorithm, Acuros XB (AXB), has shown great potential for accurate heterogeneous dose calculation. However, the clinical impact between AXB and other currently used algorithms still needs to be elucidated for translation between these algorithms. The purpose of this study was to investigate the impact of AXB for heterogeneous dose calculation in lung cancer for intensity-modulated radiation therapy (IMRT) and volumetric-modulated arc therapy (VMAT). Methods: The thorax phantom from the Radiological Physics Center (RPC) was used for this study. IMRT and VMAT plans were created for the phantom in the Eclipse 11.0 treatment planning system. Each plan was delivered to the phantom three times using a Varian Clinac iX linear accelerator to ensure reproducibility. Thermoluminescent dosimeters (TLDs) and Gafchromic EBT2 film were placed inside the phantom to measure delivered doses. The measurements were compared with dose calculations from AXB 11.0.21 and the anisotropic analytical algorithm (AAA) 11.0.21. Two dose reporting modes of AXB, dose-to-medium in medium (D m,m ) and dose-to-water in medium (D w,m ), were studied. Point doses, dose profiles, and gamma analysis were used to quantify the agreement between measurements and calculations from both AXB and AAA. The computation times for AAA and AXB were also evaluated. Results: For the RPC lung phantom, AAA and AXB dose predictions were found in good agreement to TLD and film measurements for both IMRT and VMAT plans. TLD dose predictions were within 0.4%–4.4% to AXB doses (both D m,m and D w,m ); and within 2.5%–6.4% to AAA doses, respectively. For the film comparisons, the gamma indexes (±3%/3 mm criteria) were 94%, 97%, and 98% for AAA, AXB Dm,m , and AXB Dw,m , respectively. The differences between AXB and AAA in dose–volume histogram mean doses were within 2% in the planning target volume, lung, heart, and within 5% in the spinal cord. However

  3. Nuclear data for fusion: Validation of typical pre-processing methods for radiation transport calculations

    International Nuclear Information System (INIS)

    Hutton, T.; Sublet, J.C.; Morgan, L.; Leadbeater, T.W.

    2015-01-01

    Highlights: • We quantify the effect of processing nuclear data from ENDF to ACE format. • We consider the differences between fission and fusion angular distributions. • C-nat(n,el) at 2.0 MeV has a 0.6% deviation between original and processed data. • Fe-56(n,el) at 14.1 MeV has a 11.0% deviation between original and processed data. • Processed data do not accurately depict ENDF distributions for fusion energies. - Abstract: Nuclear data form the basis of the radiation transport codes used to design and simulate the behaviour of nuclear facilities, such as the ITER and DEMO fusion reactors. Typically these data and codes are biased towards fission and high-energy physics applications yet are still applied to fusion problems. With increasing interest in fusion applications, the lack of fusion specific codes and relevant data libraries is becoming increasingly apparent. Industry standard radiation transport codes require pre-processing of the evaluated data libraries prior to use in simulation. Historically these methods focus on speed of simulation at the cost of accurate data representation. For legacy applications this has not been a major concern, but current fusion needs differ significantly. Pre-processing reconstructs the differential and double differential interaction cross sections with a coarse binned structure, or more recently as a tabulated cumulative distribution function. This work looks at the validity of applying these processing methods to data used in fusion specific calculations in comparison to fission. The relative effects of applying this pre-processing mechanism, to both fission and fusion relevant reaction channels are demonstrated, and as such the poor representation of these distributions for the fusion energy regime. For the nat C(n,el) reaction at 2.0 MeV, the binned differential cross section deviates from the original data by 0.6% on average. For the 56 Fe(n,el) reaction at 14.1 MeV, the deviation increases to 11.0%. We

  4. The use of symbolic computation in radiative, energy, and neutron transport calculations. Final report

    International Nuclear Information System (INIS)

    Frankel, J.I.

    1997-01-01

    This investigation used sysmbolic manipulation in developing analytical methods and general computational strategies for solving both linear and nonlinear, regular and singular integral and integro-differential equations which appear in radiative and mixed-mode energy transport. Contained in this report are seven papers which present the technical results as individual modules

  5. Evaluation of radiation shielding performance in sea transport of radioactive material by using simple calculation method

    International Nuclear Information System (INIS)

    Odano, N.; Ohnishi, S.; Sawamura, H.; Tanaka, Y.; Nishimura, K.

    2004-01-01

    A modified code system based on the point kernel method was developed to use in evaluation of shielding performance for maritime transport of radioactive material. For evaluation of shielding performance accurately in the case of accident, it is required to preciously model the structure of transport casks and shipping vessel, and source term. To achieve accurate modelling of the geometry and source term condition, we aimed to develop the code system by using equivalent information regarding structure and source term used in the Monte Carlo calculation code, MCNP. Therefore, adding an option to use point kernel method to the existing Monte Carlo code, MCNP4C, the code system was developed. To verify the developed code system, dose rate distribution in an exclusive shipping vessel to transport the low level radioactive wastes were calculated by the developed code and the calculated results were compared with measurements and Monte Carlo calculations. It was confirmed that the developed simple calculation method can obtain calculation results very quickly with enough accuracy comparing with the Monte Carlo calculation code MCNP4C

  6. Modeling of radionuclide transport through rock formations and the resulting radiation exposure of reference persons. Calculations using Asse II parameters

    International Nuclear Information System (INIS)

    Kueppers, Christian; Ustohalova, Veronika; Steinhoff, Mathias

    2012-01-01

    The long-term release of radioactivity into the ground water path cannot be excluded for the radioactive waste repository Asse II. The possible radiological consequences were analyzed using a radio-ecological scenario developed by GRS. A second scenario was developed considering the solubility of radionuclides in salt saturated solutions and retarding/retention effects during the radionuclide transport through the cap rock layers. The modeling of possible radiation exposure was based on the lifestyle habits of reference persons. In Germany the calculation procedure for the prediction of radionuclide release from final repositories is not defined by national standards, the used procedures are based on analogue methods from other radiation protection calculations.

  7. 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.

  8. Dose reconstruction in radioactively contaminated areas based on radiation transport calculations and measurements

    International Nuclear Information System (INIS)

    Hiller, Mauritius Michael

    2015-01-01

    The external radiation exposure at the former village of Metlino, Russia, was reconstructed. The Techa river in Metlino was contaminated by water from the Majak plant. The village was evacuated in 1956 and a reservoir lake created. Absorbed doses in bricks were measured and a model of the present-day and the historic Metlino was created for Monte Carlo calculations. By combining both, the air kerma at shoreline could be reconstructed to evaluate the Techa River Dosimetry System.

  9. Photon and electron data bases and their use in radiation transport calculations

    International Nuclear Information System (INIS)

    Cullen, D.E.; Perkins, S.T.; Seltzer, S.M.

    1992-02-01

    The ENDF/B-VI photon interaction library includes data to describe the interaction of photons with the elements Z=1 to 100 over the energy range 10 eV to 100 MeV. This library has been designed to meet the traditional needs of users to model the interaction and transport of primary photons. However, this library contains additional information which used in a combination with our other data libraries can be used to perform much more detailed calculations, e.g., emission of secondary fluorescence photons. This paper describes both traditional and more detailed uses of this library

  10. Benchmark experiment to verify radiation transport calculations for dosimetry in radiation therapy; Benchmark-Experiment zur Verifikation von Strahlungstransportrechnungen fuer die Dosimetrie in der Strahlentherapie

    Energy Technology Data Exchange (ETDEWEB)

    Renner, Franziska [Physikalisch-Technische Bundesanstalt (PTB), Braunschweig (Germany)

    2016-11-01

    Monte Carlo simulations are regarded as the most accurate method of solving complex problems in the field of dosimetry and radiation transport. In (external) radiation therapy they are increasingly used for the calculation of dose distributions during treatment planning. In comparison to other algorithms for the calculation of dose distributions, Monte Carlo methods have the capability of improving the accuracy of dose calculations - especially under complex circumstances (e.g. consideration of inhomogeneities). However, there is a lack of knowledge of how accurate the results of Monte Carlo calculations are on an absolute basis. A practical verification of the calculations can be performed by direct comparison with the results of a benchmark experiment. This work presents such a benchmark experiment and compares its results (with detailed consideration of measurement uncertainty) with the results of Monte Carlo calculations using the well-established Monte Carlo code EGSnrc. The experiment was designed to have parallels to external beam radiation therapy with respect to the type and energy of the radiation, the materials used and the kind of dose measurement. Because the properties of the beam have to be well known in order to compare the results of the experiment and the simulation on an absolute basis, the benchmark experiment was performed using the research electron accelerator of the Physikalisch-Technische Bundesanstalt (PTB), whose beam was accurately characterized in advance. The benchmark experiment and the corresponding Monte Carlo simulations were carried out for two different types of ionization chambers and the results were compared. Considering the uncertainty, which is about 0.7 % for the experimental values and about 1.0 % for the Monte Carlo simulation, the results of the simulation and the experiment coincide.

  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. Polarization effects in coherent and incoherent photon scattering: survey of measurements and theory relevant to radiation transport calculations

    International Nuclear Information System (INIS)

    Hubbell, J.H.

    1993-01-01

    This report reviews available information on polarization effects arising when photons in the X-ray and gamma-ray energy regime undergo coherent (Rayleigh) scattering and incoherent (Compton) scattering by atomic electrons. In addition to descriptions and discussions of these effects, including estimates of their magnitudes as they apply to radiation transport calculations, an annotated bibliography of 102 selected works covering the period 1905-1991 is provided, with particularly relevant works for the purpose of this report flagged with asterisks (*). A major resource for this report is a 1948 unpublished informal report by L.V. Spencer which has been quoted here almost in its entirety, since, of all the works cited in the annotated bibliography, it appears to be the only one which explicitly and directly addresses the purpose of this report. Hence this valuable material should be re-introduced into the available and current literature. (author). 119 refs., 7 figs

  13. ALBEMO, a program for the calculation of the radiation transport in void volumes with reflecting walls

    International Nuclear Information System (INIS)

    Mueller, K.; Vossebrecker, H.

    The Monte Carlo Program ALBEMO calculates the distribution of neutrons and gamma rays in void volumes which are bounded by reflecting walls with x, y, z coordinates. The program is based on the albedo method. The effect of significant simplifying assumptions is investigated. Comparisons with experiments show satisfying agreement

  14. Optimization of the Spent Fuel Attribute Tester using radiation transport calculations

    International Nuclear Information System (INIS)

    Laub, T.W.; Dupree, S.A.; Arlt, R.

    1993-01-01

    The International Atomic Energy Agency uses the Spent Fuel Attribute Tester (SFAT) to measure gamma signatures from fuel assemblies stored in spent fuel pools. It consists of a shielded, collimated NaI(Tl) detector attached to an air-filled pipe. The purpose of the present study was to define design changes, within operational constraints, that would improve the target assembly 137 Cs signal relative to the background signals from adjacent assemblies. This improvement is essential to reducing to an acceptable level the measurement time during an inspection. Monte Carlo calculations of the entire geometry were impractical, therefore, a hybrid method was developed that combined one-dimensional discrete ordinates models of the spent fuel pool, three-dimensional Monte Carlo calculations of the SFAT, and detector response calculations. The method compared well with measurements taken with the existing baseline SFAT. Calculations predicted significant improvements in signal-to-noise ratio. Recommended changes included shortening the pipe and increasing its wall thickness, placing low-Z filters in the crystal line of sight, reducing the thickness of shielding around the collimator aperture and adding shielding around the crystal, and reducing the diameter of the crystal. An instrument incorporating these design changes is being fabricated in Finland and will be tested this year

  15. Radiation Transport Calculation of the UGXR Collimators for the Jules Horowitz Reactor (JHR)

    Science.gov (United States)

    Chento, Yelko; Hueso, César; Zamora, Imanol; Fabbri, Marco; Fuente, Cristina De La; Larringan, Asier

    2017-09-01

    Jules Horowitz Reactor (JHR), a major infrastructure of European interest in the fission domain, will be built and operated in the framework of an international cooperation, including the development and qualification of materials and nuclear fuel used in nuclear industry. For this purpose UGXR Collimators, two multi slit gamma and X-ray collimation mechatronic systems, will be installed at the JHR pool and at the Irradiated Components Storage pool. Expected amounts of radiation produced by the spent fuel and X-ray accelerator implies diverse aspects need to be verified to ensure adequate radiological zoning and personnel radiation protection. A computational methodology was devised to validate the Collimators design by means of coupling different engineering codes. In summary, several assessments were performed by means of MCNP5v1.60 to fulfil all the radiological requirements in Nominal scenario (TEDE < 25µSv/h) and in Maintenance scenario (TEDE < 2mSv/h) among others, detailing the methodology, hypotheses and assumptions employed.

  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

    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

  17. 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

  18. Range calculations using multigroup transport methods

    International Nuclear Information System (INIS)

    Hoffman, T.J.; Robinson, M.T.; Dodds, H.L. Jr.

    1979-01-01

    Several aspects of radiation damage effects in fusion reactor neutron and ion irradiation environments are amenable to treatment by transport theory methods. In this paper, multigroup transport techniques are developed for the calculation of particle range distributions. These techniques are illustrated by analysis of Au-196 atoms recoiling from (n,2n) reactions with gold. The results of these calculations agree very well with range calculations performed with the atomistic code MARLOWE. Although some detail of the atomistic model is lost in the multigroup transport calculations, the improved computational speed should prove useful in the solution of fusion material design problems

  19. 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)

  20. 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

  1. 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

  2. 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)

  3. 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

  4. Electron stopping powers for transport calculations

    International Nuclear Information System (INIS)

    Berger, M.J.

    1988-01-01

    The reliability of radiation transport calculations depends on the accuracy of the input cross sections. Therefore, it is essential to review and update the cross sections from time to time. Even though the main interest of the author's group at NBS is in transport calculations and their applications, the group spends almost as much time on the analysis and preparation of cross sections as on the development of transport codes. Stopping powers, photon attenuation coefficients, bremsstrahlung cross sections, and elastic-scattering cross sections in recent years have claimed attention. This chapter deals with electron stopping powers (with emphasis on collision stopping powers), and reviews the state of the art as reflected by Report 37 of the International Commission on Radiation Units and Measurements

  5. Calculating radiation exposure and dose

    International Nuclear Information System (INIS)

    Hondros, J.

    1987-01-01

    This paper discusses the methods and procedures used to calculate the radiation exposures and radiation doses to designated employees of the Olympic Dam Project. Each of the three major exposure pathways are examined. These are: gamma irradiation, radon daughter inhalation and radioactive dust inhalation. A further section presents ICRP methodology for combining individual pathway exposures to give a total dose figure. Computer programs used for calculations and data storage are also presented briefly

  6. Modeling of radionuclide transport through rock formations and the resulting radiation exposure of reference persons. Calculations using Asse II parameters; Modellierung des Transports von Radionukliden durch Gesteinsschichten und der resultierenden Strahlenexposition von Referenzpersonen. Berechnungen mit Parametern der Asse II

    Energy Technology Data Exchange (ETDEWEB)

    Kueppers, Christian; Ustohalova, Veronika; Steinhoff, Mathias

    2012-05-21

    The long-term release of radioactivity into the ground water path cannot be excluded for the radioactive waste repository Asse II. The possible radiological consequences were analyzed using a radio-ecological scenario developed by GRS. A second scenario was developed considering the solubility of radionuclides in salt saturated solutions and retarding/retention effects during the radionuclide transport through the cap rock layers. The modeling of possible radiation exposure was based on the lifestyle habits of reference persons. In Germany the calculation procedure for the prediction of radionuclide release from final repositories is not defined by national standards, the used procedures are based on analogue methods from other radiation protection calculations.

  7. Prenatal radiation exposure. Dose calculation

    International Nuclear Information System (INIS)

    Scharwaechter, C.; Schwartz, C.A.; Haage, P.; Roeser, A.

    2015-01-01

    The unborn child requires special protection. In this context, the indication for an X-ray examination is to be checked critically. If thereupon radiation of the lower abdomen including the uterus cannot be avoided, the examination should be postponed until the end of pregnancy or alternative examination techniques should be considered. Under certain circumstances, either accidental or in unavoidable cases after a thorough risk assessment, radiation exposure of the unborn may take place. In some of these cases an expert radiation hygiene consultation may be required. This consultation should comprise the expected risks for the unborn while not perturbing the mother or the involved medical staff. For the risk assessment in case of an in-utero X-ray exposition deterministic damages with a defined threshold dose are distinguished from stochastic damages without a definable threshold dose. The occurrence of deterministic damages depends on the dose and the developmental stage of the unborn at the time of radiation. To calculate the risks of an in-utero radiation exposure a three-stage concept is commonly applied. Depending on the amount of radiation, the radiation dose is either estimated, roughly calculated using standard tables or, in critical cases, accurately calculated based on the individual event. The complexity of the calculation thereby increases from stage to stage. An estimation based on stage one is easily feasible whereas calculations based on stages two and especially three are more complex and often necessitate execution by specialists. This article demonstrates in detail the risks for the unborn child pertaining to its developmental phase and explains the three-stage concept as an evaluation scheme. It should be noted, that all risk estimations are subject to considerable uncertainties.

  8. 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.)

  9. Radiation-damage calculations with NJOY

    International Nuclear Information System (INIS)

    MacFarlane, R.E.; Muir, D.W.; Mann, F.W.

    1983-01-01

    Atomic displacement, gas production, transmutation, and nuclear heating can all be calculated with the NJOY nuclear data processing system using evaluated data in ENDF/B format. Using NJOY helps assure consistency between damage cross sections and those used for transport, and NJOY provides convenient interface formats for linking data to application codes. Unique features of the damage calculation include a simple momentum balance treatment for radiative capture and a new model for (n, particle) reactions based on statistical model calculations. Sample results for iron and nickel are given and compared with the results of other methods

  10. Spontaneous Radiation Background Calculation for LCLS

    CERN Document Server

    Reiche, Sven

    2004-01-01

    The intensity of undulator radiation, not amplified by the FEL interaction, can be larger than the maximum FEL signal in the case of an X-ray FEL. In the commissioning of a SASE FEL it is essential to extract an amplified signal early to diagnose eventual misalignment of undulator modules or errors in the undulator field strength. We developed a numerical code to calculate the radiation pattern at any position behind a multi-segmented undulator with arbitrary spacing and field profiles. The output can be run through numerical spatial and frequency filters to model the radiation beam transport and diagnostic. In this presentation we estimate the expected background signal for the FEL diagnostic and at what point along the undulator the FEL signal can be separated from the background. We also discusses how much information on the undulator field and alignment can be obtained from the incoherent radiation signal itself.

  11. Whole body MR-PET: a new internal dosimetry method for radiation transport calculation from biokinetic model data

    International Nuclear Information System (INIS)

    Nunes, Ana; Alves, Francisco; Patrício, Miguel

    2014-01-01

    In order to ensure the safe usage of new radiopharmaceuticals in Positron Emission Tomography (PET), it is necessary to quantify the doses delivered to the organs and tissues within the patients’ bodies. A framework that allows estimating the dose delivered by PET has been established by the MIRD Committee [1, 2] and ICRP []. Although this covers the most important terms and concepts in Internal Radiation Dosimetry (IRD), it does not provide a detailed guide to assist in the development of a full dosimetric study. We discuss the development, implementation, assessment and validation of an accurate method for IRD studies of PET radiotracers.

  12. New calculation of derived limits for the 1960 radiation protection guides reflecting updated models for dosimetry and biological transport

    International Nuclear Information System (INIS)

    Eckerman, K.F.; Watson, S.B.; Nelson, C.B.; Nelson, D.R.; Richardson, A.C.B.; Sullivan, R.E.

    1984-12-01

    This report presents revised values for the radioactivity concentration guides (RCGs), based on the 1960 primary radiation protection guides (RPGs) for occupational exposure (FRC 1960) and for underground uranium miners (EPA 1971a) using the updated dosimetric models developed to prepare ICRP Publication 30. Unlike the derived quantities presented in Publication 30, which are based on limitation of the weighted sum of doses to all irradiated tissues, these RCGs are based on the ''critical organ'' approach of the 1960 guidance, which was a single limit for the most critically irradiated organ or tissue. This report provides revised guides for the 1960 Federal guidance which are consistent with current dosimetric relationships. 2 figs., 4 tabs

  13. 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...

  14. Biosphere transport and radiation dose calculations resulting from radioactive waste stored in deep salt formation (PACOMA-project)

    International Nuclear Information System (INIS)

    Jong, E.J. de; Koester, H.W.; Vries, W.J. de; Lembrechts, J.F.

    1990-03-01

    Parts are presented of the results of a safety-assessment study of disposal of medium and low level radioactive waste in salt formations in the Netherlands. The study concerns several disposal concepts for 2 kinds of salt formation, a deep dome and a shallow dome. 7 cases were studied with the same Dutch inventory and 1 with a reference inventory R, in order to compare results with those of other PACOMA participants. The total activity of the reference inventory R is 30 percent lower than the Dutch inventory, but some long living nuclides such as I-129, Np-237 and U-238 have a considerably higher activity. This reference inventor R has been combined with the disposal concept of mined cavities in a shallow salt dome. In each case. the released fraction of stored radio-nuclides moves gradually with water through the geosphere to the bio-sphere where it enters a river. River water is used for sprinkler irrigation and for drinking by man and livestock. The dispersal of the radionuclides into the biosphere is calculated with the BIOS program of the NRPB. Subroutines linked to the program add doses via different pathways to obtain a maximum individual dose, a collective dose and an integrated collective dose. This study presents results of these calculations. (author). 11 refs.; 39 figs.; 111 tabs

  15. Engineering calculations in radiative heat transfer

    CERN Document Server

    Gray, W A; Hopkins, D W

    1974-01-01

    Engineering Calculations in Radiative Heat Transfer is a six-chapter book that first explains the basic principles of thermal radiation and direct radiative transfer. Total exchange of radiation within an enclosure containing an absorbing or non-absorbing medium is then described. Subsequent chapters detail the radiative heat transfer applications and measurement of radiation and temperature.

  16. The use of symbolic computation in radiative, energy, and neutron transport calculations. Technical report, 15 August 1992--14 August 1994

    International Nuclear Information System (INIS)

    Frankel, J.I.

    1995-01-01

    This investigation uses symbolic computation in developing analytical methods and general computational strategies for solving both linear and nonlinear, regular and singular, integral and integro-differential equations which appear in radiative and combined mode energy transport. This technical report summarizes the research conducted during the first nine months of the present investigation. The use of Chebyshev polynomials augmented with symbolic computation has clearly been demonstrated in problems involving radiative (or neutron) transport, and mixed-mode energy transport. Theoretical issues related to convergence, errors, and accuracy have also been pursued. Three manuscripts have resulted from the funded research. These manuscripts have been submitted to archival journals. At the present time, an investigation involving a conductive and radiative medium is underway. The mathematical formulation leads to a system of nonlinear, weakly-singular integral equations involving the unknown temperature and various Legendre moments of the radiative intensity in a participating medium. Some preliminary results are presented illustrating the direction of the proposed research

  17. 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

  18. Calculation of neutron and gamma transport at the FOA:type of problems and calculation methods

    International Nuclear Information System (INIS)

    Lefvert, T.

    1975-11-01

    Protection against the effects of nuclear warfare involves the analysis of the forms of results of a nuclear charge explosion producing neutron and gamma radiation. It brings out problems leading to the calculation of criticality, leakage, and deep transmission. Methods have been developed for various kinds of particle transport problems. Applications to radiation therapy, storage of fissile materials, and fast reactors are discussed. A list (with brief description) of all neutron and gamma transport programmes of the FOA is given. (J.S.)

  19. Optimal calculational schemes for solving multigroup photon transport problem

    International Nuclear Information System (INIS)

    Dubinin, A.A.; Kurachenko, Yu.A.

    1987-01-01

    A scheme of complex algorithm for solving multigroup equation of radiation transport is suggested. The algorithm is based on using the method of successive collisions, the method of forward scattering and the spherical harmonics method, and is realized in the FORAP program (FORTRAN, BESM-6 computer). As an example the results of calculating reactor photon transport in water are presented. The considered algorithm being modified may be used for solving neutron transport problems

  20. 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

  1. Utilization of Monte Carlo Calculations in Radiation Transport Analyses to Support the Design of the U.S. Spallation Neutron Source (SNS)

    International Nuclear Information System (INIS)

    Johnson, J.O.

    2000-01-01

    The Department of Energy (DOE) has given the Spallation Neutron Source (SNS) project approval to begin Title I design of the proposed facility to be built at Oak Ridge National Laboratory (ORNL) and construction is scheduled to commence in FY01 . The SNS initially will consist of an accelerator system capable of delivering an ∼0.5 microsecond pulse of 1 GeV protons, at a 60 Hz frequency, with 1 MW of beam power, into a single target station. The SNS will eventually be upgraded to a 2 MW facility with two target stations (a 60 Hz station and a 10 Hz station). The radiation transport analysis, which includes the neutronic, shielding, activation, and safety analyses, is critical to the design of an intense high-energy accelerator facility like the proposed SNS, and the Monte Carlo method is the cornerstone of the radiation transport analyses

  2. Stochastic calculations for radiation risk assessment: a Monte-Carlo approach to the simulation of radiocesium transport in the pasture-cow-milk food chain

    Energy Technology Data Exchange (ETDEWEB)

    Mathies, M; Eisfeld, K; Paretzke, H; Wirth, E [Gesellschaft fuer Strahlen- und Umweltforschung m.b.H. Muenchen, Neuherberg (Germany, F.R.). Inst. fuer Strahlenschutz

    1981-05-01

    The effects of introducing probability distributions of the parameters in radionuclide transport models are investigated. Results from a Monte-Carlo simulation were presented for the transport of /sup 137/Cs via the pasture-cow-milk pathway, taking into the account the uncertainties and naturally occurring fluctuations in the rate constants. The results of the stochastic model calculations characterize the activity concentrations at a given time t and provide a great deal more information for analysis of the environmental transport of radionuclides than deterministic calculations in which the variation of parameters is not taken into consideration. Moreover the stochastic model permits an estimate of the variation of the physico-chemical behaviour of radionuclides in the environment in a more realistic way than by using only the highest transfer coefficients in deterministic approaches, which can lead to non-realistic overestimates of the probability with which high activity levels will be encountered.

  3. Calculating Risk: Radiation and Chernobyl.

    Science.gov (United States)

    Gale, Robert Peter

    1987-01-01

    Considers who is at risk in a disaster such as Chernobyl. Assesses the difficulty in translating information regarding radiation to the public and in determining the acceptability of technological risks. (NKA)

  4. SR 97 - Radionuclide transport calculations

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Maria [Kemakta Konsult AB, Stockholm (Sweden); Lindstroem, Fredrik [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden)

    1999-12-01

    An essential component of a safety assessment is to calculate radionuclide release and dose consequences for different scenarios and cases. The SKB tools for such a quantitative assessment are used to calculate the maximum releases and doses for the hypothetical repository sites Aberg, Beberg and Ceberg for the initial canister defect scenario and also for the glacial melting case for Aberg. The reasonable cases, i.e. all parameters take reasonable values, results in maximum biosphere doses of 5x10{sup -8} Sv/yr for Aberg, 3x10{sup -8} Sv/yr for Beberg and 1x10{sup -8} Sv/yr for Ceberg for peat area. These doses lie significantly below 0.15 mSv/yr. (A dose of 0.15 mSv/yr for unit probability corresponds to the risk limit of 10{sup -5} per year for the most exposed individuals recommended in regulations.) The conclusion that the maximum risk would lie well below 10{sup -5} per year is also demonstrated by results from the probabilistic calculations, which directly assess the resulting risk by combining dose and probability estimates. The analyses indicate that the risk is 2x10{sup -5} Sv/yr for Aberg, 8x10{sup -7} Sv/yr for Beberg and 3x10{sup -8} Sv/yr for Ceberg. The analysis shows that the most important parameters in the near field are the number of defective canisters and the instant release fraction. The influence from varying one parameter never changes the doses as much as an order of magnitude. In the far field the most important uncertainties affecting release and retention are associated with permeability and connectivity of the fractures in the rock. These properties affect several parameters. Highly permeable and well connected fractures imply high groundwater fluxes and short groundwater travel times. Sparsely connected or highly variable fracture properties implies low flow wetted surface along migration paths. It should, however, be remembered that the far-field parameters have little importance if the near-field parameters take their reasonable

  5. SR 97 - Radionuclide transport calculations

    International Nuclear Information System (INIS)

    Lindgren, Maria; Lindstroem, Fredrik

    1999-12-01

    An essential component of a safety assessment is to calculate radionuclide release and dose consequences for different scenarios and cases. The SKB tools for such a quantitative assessment are used to calculate the maximum releases and doses for the hypothetical repository sites Aberg, Beberg and Ceberg for the initial canister defect scenario and also for the glacial melting case for Aberg. The reasonable cases, i.e. all parameters take reasonable values, results in maximum biosphere doses of 5x10 -8 Sv/yr for Aberg, 3x10 -8 Sv/yr for Beberg and 1x10 -8 Sv/yr for Ceberg for peat area. These doses lie significantly below 0.15 mSv/yr. (A dose of 0.15 mSv/yr for unit probability corresponds to the risk limit of 10 -5 per year for the most exposed individuals recommended in regulations.) The conclusion that the maximum risk would lie well below 10 -5 per year is also demonstrated by results from the probabilistic calculations, which directly assess the resulting risk by combining dose and probability estimates. The analyses indicate that the risk is 2x10 -5 Sv/yr for Aberg, 8x10 -7 Sv/yr for Beberg and 3x10 -8 Sv/yr for Ceberg. The analysis shows that the most important parameters in the near field are the number of defective canisters and the instant release fraction. The influence from varying one parameter never changes the doses as much as an order of magnitude. In the far field the most important uncertainties affecting release and retention are associated with permeability and connectivity of the fractures in the rock. These properties affect several parameters. Highly permeable and well connected fractures imply high groundwater fluxes and short groundwater travel times. Sparsely connected or highly variable fracture properties implies low flow wetted surface along migration paths. It should, however, be remembered that the far-field parameters have little importance if the near-field parameters take their reasonable values. In that case almost all

  6. Radiation protection calculations for diagnostic medical equipment

    International Nuclear Information System (INIS)

    Klueter, R.

    1992-01-01

    The standards DIN 6812 and DIN 6844 define the radiation protection requirements to be met by biomedical radiography equipment or systems for nuclear medicine. The paper explains the use of a specific computer program for radiation protection calculations. The program offers menu-controlled calculation, with free choice of the relevant nuclides. (DG) [de

  7. Sn transport calculations on vector and parallel processors

    International Nuclear Information System (INIS)

    Rhoades, W.A.; Childs, R.L.

    1987-01-01

    The transport of radiation from the source to the location of people or equipment gives rise to some of the most challenging of calculations. A problem may involve as many as a billion unknowns, each evaluated several times to resolve interdependence. Such calculations run many hours on a Cray computer, and a typical study involves many such calculations. This paper will discuss the steps taken to vectorize the DOT code, which solves transport problems in two space dimensions (2-D); the extension of this code to 3-D; and the plans for extension to parallel processors

  8. 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

  9. 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.

  10. Lagrangian Transport Calculations Using UARS Data. Part I: Passive Tracers

    Science.gov (United States)

    Manney, G. L.; Lahoz, W. A.; Harwood, R. S.; Zurek, R. W.; Kumer, J. B.; Mergenthaler, J. L.; Roche, A. E.; O'Neill, A; Swinbank, R.; Waters, J. W.

    1994-01-01

    The transport of passive tracers observed by UARS has been simulated using computed trajectories of thousands of air parcels initialized on a three-dimensional stratospheric grid. These trajectories are calculated in isentropic coordinates using horizontal winds provided by the United Kingdom Meteorological Office data assimilation system and vertical (cross-isentropic) velocities computed using a fast radiation code.

  11. Discussion of electron cross sections for transport calculations

    International Nuclear Information System (INIS)

    Berger, M.J.

    1983-01-01

    This paper deals with selected aspects of the cross sections needed as input for transport calculations and for the modeling of radiation effects in biological materials. Attention is centered mainly on the cross sections for inelastic interactions between electrons and water molecules and the use of these cross sections for the calculation of energy degradation spectra and of ionization and excitation yields. 40 references, 3 figures, 1 table

  12. 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

  13. 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

  14. Radiation shielding calculation using MCNP

    International Nuclear Information System (INIS)

    Masukawa, Fumihiro

    2001-01-01

    To verify the Monte Carlo code MCNP4A as a tool to generate the reference data in the shielding designs and the safety evaluations, various shielding benchmark experiments were analyzed using this code. These experiments were categorized in three types of the shielding subjects; bulk shielding, streaming, and skyshine. For the variance reduction technique, which is indispensable to get meaningful results with the Monte Carlo shielding calculation, we mainly used the weight window, the energy dependent Russian roulette and spitting. As a whole, our analyses performed enough small statistical errors and showed good agreements with these experiments. (author)

  15. Calculating the Responses of Self-Powered Radiation Detectors.

    Science.gov (United States)

    Thornton, D. A.

    Available from UMI in association with The British Library. The aim of this research is to review and develop the theoretical understanding of the responses of Self -Powered Radiation Detectors (SPDs) in Pressurized Water Reactors (PWRs). Two very different models are considered. A simple analytic model of the responses of SPDs to neutrons and gamma radiation is presented. It is a development of the work of several previous authors and has been incorporated into a computer program (called GENSPD), the predictions of which have been compared with experimental and theoretical results reported in the literature. Generally, the comparisons show reasonable consistency; where there is poor agreement explanations have been sought and presented. Two major limitations of analytic models have been identified; neglect of current generation in insulators and over-simplified electron transport treatments. Both of these are developed in the current work. A second model based on the Explicit Representation of Radiation Sources and Transport (ERRST) is presented and evaluated for several SPDs in a PWR at beginning of life. The model incorporates simulation of the production and subsequent transport of neutrons, gamma rays and electrons, both internal and external to the detector. Neutron fluxes and fuel power ratings have been evaluated with core physics calculations. Neutron interaction rates in assembly and detector materials have been evaluated in lattice calculations employing deterministic transport and diffusion methods. The transport of the reactor gamma radiation has been calculated with Monte Carlo, adjusted diffusion and point-kernel methods. The electron flux associated with the reactor gamma field as well as the internal charge deposition effects of the transport of photons and electrons have been calculated with coupled Monte Carlo calculations of photon and electron transport. The predicted response of a SPD is evaluated as the sum of contributions from individual

  16. Three dimensions transport calculations for PWR core

    International Nuclear Information System (INIS)

    Richebois, E.

    2000-01-01

    The objective of this work is to define improved 3-D core calculation methods based on the transport theory. These methods can be particularly useful and lead to more precise computations in areas of the core where anisotropy and steep flux gradients occur, especially near interface and boundary conditions and in regions of high heterogeneity (bundle with absorbent rods). In order to apply the transport theory a new method for calculating reflector constants has been developed, since traditional methods were only suited for 2-group diffusion core calculations and could not be extrapolated to transport calculations. In this thesis work, the new method for obtaining reflector constants is derived regardless of the number of energy groups and of the operator used. The core calculations results using the reflector constants thereof obtained have been validated on the EDF's power reactor Saint Laurent B1 with MOX loading. The advantages of a 3-D core transport calculation scheme have been highlighted as opposed to diffusion methods; there are a considerable number of significant effects and potential advantages to be gained in rod worth calculations for instance. These preliminary results obtained with on particular cycle will have to be confirmed by more systematic analysis. Accidents like MSLB (main steam line break) and LOCA (loss of coolant accident) should also be investigated and constitute challenging situations where anisotropy is high and/or flux gradients are steep. This method is now being validated for others EDF's PWRs' reactors, as well as for experimental reactors and other types of commercial reactors. (author)

  17. Molecular transport calculations with Wannier Functions

    DEFF Research Database (Denmark)

    Thygesen, Kristian Sommer; Jacobsen, Karsten Wedel

    2005-01-01

    We present a scheme for calculating coherent electron transport in atomic-scale contacts. The method combines a formally exact Green's function formalism with a mean-field description of the electronic structure based on the Kohn-Sham scheme of density functional theory. We use an accurate plane...

  18. Kinetic neoclassical calculations of impurity radiation profiles

    Directory of Open Access Journals (Sweden)

    D.P. Stotler

    2017-08-01

    Full Text Available Modifications of the drift-kinetic transport code XGC0 to include the transport, ionization, and recombination of individual charge states, as well as the associated radiation, are described. The code is first applied to a simulation of an NSTX H-mode discharge with carbon impurity to demonstrate the approach to coronal equilibrium. The effects of neoclassical phenomena on the radiated power profile are examined sequentially through the activation of individual physics modules in the code. Orbit squeezing and the neoclassical inward pinch result in increased radiation for temperatures above a few hundred eV and changes to the ratios of charge state emissions at a given electron temperature. Analogous simulations with a neon impurity yield qualitatively similar results.

  19. Feasibility study on embedded transport core calculations

    International Nuclear Information System (INIS)

    Ivanov, B.; Zikatanov, L.; Ivanov, K.

    2007-01-01

    The main objective of this study is to develop an advanced core calculation methodology based on embedded diffusion and transport calculations. The scheme proposed in this work is based on embedded diffusion or SP 3 pin-by-pin local fuel assembly calculation within the framework of the Nodal Expansion Method (NEM) diffusion core calculation. The SP 3 method has gained popularity in the last 10 years as an advanced method for neutronics calculation. NEM is a multi-group nodal diffusion code developed, maintained and continuously improved at the Pennsylvania State University. The developed calculation scheme is a non-linear iteration process, which involves cross-section homogenization, on-line discontinuity factors generation, and boundary conditions evaluation by the global solution passed to the local calculation. In order to accomplish the local calculation, a new code has been developed based on the Finite Elements Method (FEM), which is capable of performing both diffusion and SP 3 calculations. The new code will be used in the framework of the NEM code in order to perform embedded pin-by-pin diffusion and SP 3 calculations on fuel assembly basis. The development of the diffusion and SP 3 FEM code is presented first following by its application to several problems. Description of the proposed embedded scheme is provided next as well as the obtained preliminary results of the C3 MOX benchmark. The results from the embedded calculations are compared with direct pin-by-pin whole core calculations in terms of accuracy and efficiency followed by conclusions made about the feasibility of the proposed embedded approach. (authors)

  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. Charged-particle calculations using Boltzmann transport methods

    International Nuclear Information System (INIS)

    Hoffman, T.J.; Dodds, H.L. Jr.; Robinson, M.T.; Holmes, D.K.

    1981-01-01

    Several aspects of radiation damage effects in fusion reactor neutron and ion irradiation environments are amenable to treatment by transport theory methods. In this paper, multigroup transport techniques are developed for the calculation of charged particle range distributions, reflection coefficients, and sputtering yields. The Boltzmann transport approach can be implemented, with minor changes, in standard neutral particle computer codes. With the multigroup discrete ordinates code, ANISN, determination of ion and target atom distributions as functions of position, energy, and direction can be obtained without the stochastic error associated with atomistic computer codes such as MARLOWE and TRIM. With the multigroup Monte Carlo code, MORSE, charged particle effects can be obtained for problems associated with very complex geometries. Results are presented for several charged particle problems. Good agreement is obtained between quantities calculated with the multigroup approach and those obtained experimentally or by atomistic computer codes

  2. 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)

  3. A Methodology for Calculating Radiation Signatures

    Energy Technology Data Exchange (ETDEWEB)

    Klasky, Marc Louis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wilcox, Trevor [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bathke, Charles G. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); James, Michael R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-05-01

    A rigorous formalism is presented for calculating radiation signatures from both Special Nuclear Material (SNM) as well as radiological sources. The use of MCNP6 in conjunction with CINDER/ORIGEN is described to allow for the determination of both neutron and photon leakages from objects of interest. In addition, a description of the use of MCNP6 to properly model the background neutron and photon sources is also presented. Examinations of the physics issues encountered in the modeling are investigated so as to allow for guidance in the user discerning the relevant physics to incorporate into general radiation signature calculations. Furthermore, examples are provided to assist in delineating the pertinent physics that must be accounted for. Finally, examples of detector modeling utilizing MCNP are provided along with a discussion on the generation of Receiver Operating Curves, which are the suggested means by which to determine detectability radiation signatures emanating from objects.

  4. Calculations on safe storage and transportation of radioactive materials

    Energy Technology Data Exchange (ETDEWEB)

    Hathout, A M; El-Messiry, A M; Amin, E [National Center for Nuclear Safety and Radiation Control and AEA, Cairo (Egypt)

    1997-12-31

    In this work the safe storage and transportation of fresh fuel as a radioactive material studied. Egypt planned ET RR 2 reactor which is of relatively high power and would require adequate handling and transportation. Therefore, the present work is initiated to develop a procedure for safe handling and transportation of radioactive materials. The possibility of reducing the magnitude of radiation transmitted on the exterior of the packages is investigated. Neutron absorbers are used to decrease the neutron flux. Criticality calculations are carried out to ensure the achievement of subcriticality so that the inherent safety can be verified. The discrete ordinate transport code ANISN was used. The results show good agreement with other techniques. 2 figs., 2 tabs.

  5. A meshless approach to radionuclide transport calculations

    International Nuclear Information System (INIS)

    Perko, J.; Sarler, B.

    2005-01-01

    Over the past thirty years numerical modelling has emerged as an interdisciplinary scientific discipline which has a significant impact in engineering and design. In the field of numerical modelling of transport phenomena in porous media, many commercial codes exist, based on different numerical methods. Some of them are widely used for performance assessment and safety analysis of radioactive waste repositories and groundwater modelling. Although they proved to be an accurate and reliable tool, they have certain limitations and drawbacks. Realistic problems often involve complex geometry which is difficult and time consuming to discretize. In recent years, meshless methods have attracted much attention due to their flexibility in solving engineering and scientific problems. In meshless methods the cumbersome polygonization of calculation domain is not necessary. By this the discretization time is reduced. In addition, the simulation is not as discretization density dependent as in traditional methods because of the lack of polygon interfaces. In this work fully meshless Diffuse Approximate Method (DAM) is used for calculation of radionuclide transport. Two cases are considered; First 1D comparison of 226 Ra transport and decay solved by the commercial Finite Volume Method (FVM) and Finite Element Method (FEM) based packages and DAM. This case shows the level of discretization density dependence. And second realistic 2D case of near-field modelling of radionuclide transport from the radioactive waste repository. Comparison is made again between FVM based code and DAM simulation for two radionuclides: Long-lived 14 C and short-lived 3 H. Comparisons indicate great capability of meshless methods to simulate complex transport problems and show that they should be seriously considered in future commercial simulation tools. (author)

  6. Calculation of transportation energy for biomass collection

    Energy Technology Data Exchange (ETDEWEB)

    Kanai, G.; Takekura, K.; Kato, H.; Kobayashi, Y.; Yakushido, K. [National Agricultural Research Center, Tsukuba, Ibaraki (Japan)

    2010-07-01

    This paper reported on a study at a rice straw facility in Japan that produces bioethanol. Simulation modeling and calculations methods were used to examine the characteristics of field-to-facility transportation. Fuel consumption was found to be influenced by the conversion rate from straw to ethanol, the quantity of straw collected, and the ratio of the field area to that around the facility. Standard conditions were assumed based on reported data and actual observations for 15 ML/yr ethanol production, 0.3 kL output of ethanol from 1 t dry straw, 53.6 day/yr working days, 2.7 t truck load capacity, and 0.128 as the ratio of field to the area around the facility. According to calculations, a quantity of 50 kt dry straw requires 2.78 L of fuel to transport 1 t of dry straw, 109.5 trucks, and a 19.1 km collection area radius. The fuel consumption for transportation was found to be proportional to the quantity of straw to the 0.5 power, but inversely proportional to the ratio of field to the 0.5 power. The rate of increase in the number of trucks needed to collect straw increases with the decrease in the ratio of the field to area surface around the facility.

  7. Comparison of 2 accelerators of Monte Carlo radiation transport calculations, NVIDIA tesla M2090 GPU and Intel Xeon Phi 5110p coprocessor: a case study for X-ray CT Imaging Dose calculation

    International Nuclear Information System (INIS)

    Liu, T.; Xu, X.G.; Carothers, C.D.

    2013-01-01

    Hardware accelerators are currently becoming increasingly important in boosting high performance computing systems. In this study, we tested the performance of two accelerator models, NVIDIA Tesla M2090 GPU and Intel Xeon Phi 5110p coprocessor, using a new Monte Carlo photon transport package called ARCHER-CT we have developed for fast CT imaging dose calculation. The package contains three code variants, ARCHER-CT(CPU), ARCHER-CT(GPU) and ARCHER-CT(COP) to run in parallel on the multi-core CPU, GPU and coprocessor architectures respectively. A detailed GE LightSpeed Multi-Detector Computed Tomography (MDCT) scanner model and a family of voxel patient phantoms were included in the code to calculate absorbed dose to radiosensitive organs under specified scan protocols. The results from ARCHER agreed well with those from the production code Monte Carlo N-Particle eXtended (MCNPX). It was found that all the code variants were significantly faster than the parallel MCNPX running on 12 MPI processes, and that the GPU and coprocessor performed equally well, being 2.89-4.49 and 3.01-3.23 times faster than the parallel ARCHER-CT(CPU) running with 12 hyper-threads. (authors)

  8. Comparison of Two Accelerators for Monte Carlo Radiation Transport Calculations, NVIDIA Tesla M2090 GPU and Intel Xeon Phi 5110p Coprocessor: A Case Study for X-ray CT Imaging Dose Calculation

    Science.gov (United States)

    Liu, Tianyu; Xu, X. George; Carothers, Christopher D.

    2014-06-01

    Hardware accelerators are currently becoming increasingly important in boosting high performance computing sys- tems. In this study, we tested the performance of two accelerator models, NVIDIA Tesla M2090 GPU and Intel Xeon Phi 5110p coprocessor, using a new Monte Carlo photon transport package called ARCHER-CT we have developed for fast CT imaging dose calculation. The package contains three code variants, ARCHER - CTCPU, ARCHER - CTGPU and ARCHER - CTCOP to run in parallel on the multi-core CPU, GPU and coprocessor architectures respectively. A detailed GE LightSpeed Multi-Detector Computed Tomography (MDCT) scanner model and a family of voxel patient phantoms were included in the code to calculate absorbed dose to radiosensitive organs under specified scan protocols. The results from ARCHER agreed well with those from the production code Monte Carlo N-Particle eXtended (MCNPX). It was found that all the code variants were significantly faster than the parallel MCNPX running on 12 MPI processes, and that the GPU and coprocessor performed equally well, being 2.89~4.49 and 3.01~3.23 times faster than the parallel ARCHER - CTCPU running with 12 hyperthreads.

  9. 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)

  10. Comparison of two accelerators for Monte Carlo radiation transport calculations, Nvidia Tesla M2090 GPU and Intel Xeon Phi 5110p coprocessor: A case study for X-ray CT imaging dose calculation

    International Nuclear Information System (INIS)

    Liu, T.; Xu, X.G.; Carothers, C.D.

    2015-01-01

    Highlights: • A new Monte Carlo photon transport code ARCHER-CT for CT dose calculations is developed to execute on the GPU and coprocessor. • ARCHER-CT is verified against MCNP. • The GPU code on an Nvidia M2090 GPU is 5.15–5.81 times faster than the parallel CPU code on an Intel X5650 6-core CPU. • The coprocessor code on an Intel Xeon Phi 5110p coprocessor is 3.30–3.38 times faster than the CPU code. - Abstract: Hardware accelerators are currently becoming increasingly important in boosting high performance computing systems. In this study, we tested the performance of two accelerator models, Nvidia Tesla M2090 GPU and Intel Xeon Phi 5110p coprocessor, using a new Monte Carlo photon transport package called ARCHER-CT we have developed for fast CT imaging dose calculation. The package contains three components, ARCHER-CT CPU , ARCHER-CT GPU and ARCHER-CT COP designed to be run on the multi-core CPU, GPU and coprocessor architectures respectively. A detailed GE LightSpeed Multi-Detector Computed Tomography (MDCT) scanner model and a family of voxel patient phantoms are included in the code to calculate absorbed dose to radiosensitive organs under user-specified scan protocols. The results from ARCHER agree well with those from the production code Monte Carlo N-Particle eXtended (MCNPX). It is found that all the code components are significantly faster than the parallel MCNPX run on 12 MPI processes, and that the GPU and coprocessor codes are 5.15–5.81 and 3.30–3.38 times faster than the parallel ARCHER-CT CPU , respectively. The M2090 GPU performs better than the 5110p coprocessor in our specific test. Besides, the heterogeneous computation mode in which the CPU and the hardware accelerator work concurrently can increase the overall performance by 13–18%

  11. Problems in radiation shielding calculations with Monte Carlo methods

    International Nuclear Information System (INIS)

    Ueki, Kohtaro

    1985-01-01

    The Monte Carlo method is a very useful tool for solving a large class of radiation transport problem. In contrast with deterministic method, geometric complexity is a much less significant problem for Monte Carlo calculations. However, the accuracy of Monte Carlo calculations is of course, limited by statistical error of the quantities to be estimated. In this report, we point out some typical problems to solve a large shielding system including radiation streaming. The Monte Carlo coupling technique was developed to settle such a shielding problem accurately. However, the variance of the Monte Carlo results using the coupling technique of which detectors were located outside the radiation streaming, was still not enough. So as to bring on more accurate results for the detectors located outside the streaming and also for a multi-legged-duct streaming problem, a practicable way of ''Prism Scattering technique'' is proposed in the study. (author)

  12. 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

  13. An improved algorithm for calculating cloud radiation

    International Nuclear Information System (INIS)

    Yuan Guibin; Sun Xiaogang; Dai Jingmin

    2005-01-01

    Clouds radiation characteristic is very important in cloud scene simulation, weather forecasting, pattern recognition, and other fields. In order to detect missiles against cloud backgrounds, to enhance the fidelity of simulation, it is critical to understand a cloud's thermal radiation model. Firstly, the definition of cloud layer infrared emittance is given. Secondly, the discrimination conditions of judging a pixel of focal plane on a satellite in daytime or night time are shown and equations are given. Radiance such as reflected solar radiance, solar scattering, diffuse solar radiance, solar and thermal sky shine, solar and thermal path radiance, cloud blackbody and background radiance are taken into account. Thirdly, the computing methods of background radiance for daytime and night time are given. Through simulations and comparison, this algorithm is proved to be an effective calculating algorithm for cloud radiation

  14. Uncertainty analysis of neutron transport calculation

    International Nuclear Information System (INIS)

    Oka, Y.; Furuta, K.; Kondo, S.

    1987-01-01

    A cross section sensitivity-uncertainty analysis code, SUSD was developed. The code calculates sensitivity coefficients for one and two-dimensional transport problems based on the first order perturbation theory. Variance and standard deviation of detector responses or design parameters can be obtained using cross section covariance matrix. The code is able to perform sensitivity-uncertainty analysis for secondary neutron angular distribution(SAD) and secondary neutron energy distribution(SED). Covariances of 6 Li and 7 Li neutron cross sections in JENDL-3PR1 were evaluated including SAD and SED. Covariances of Fe and Be were also evaluated. The uncertainty of tritium breeding ratio, fast neutron leakage flux and neutron heating was analysed on four types of blanket concepts for a commercial tokamak fusion reactor. The uncertainty of tritium breeding ratio was less than 6 percent. Contribution from SAD/SED uncertainties are significant for some parameters. Formulas to estimate the errors of numerical solution of the transport equation were derived based on the perturbation theory. This method enables us to deterministically estimate the numerical errors due to iterative solution, spacial discretization and Legendre polynomial expansion of transfer cross-sections. The calculational errors of the tritium breeding ratio and the fast neutron leakage flux of the fusion blankets were analysed. (author)

  15. Transportation channels calculation method in MATLAB

    International Nuclear Information System (INIS)

    Averyanov, G.P.; Budkin, V.A.; Dmitrieva, V.V.; Osadchuk, I.O.; Bashmakov, Yu.A.

    2014-01-01

    Output devices and charged particles transport channels are necessary components of any modern particle accelerator. They differ both in sizes and in terms of focusing elements depending on particle accelerator type and its destination. A package of transport line designing codes for magnet optical channels in MATLAB environment is presented in this report. Charged particles dynamics in a focusing channel can be studied easily by means of the matrix technique. MATLAB usage is convenient because its information objects are matrixes. MATLAB allows the use the modular principle to build the software package. Program blocks are small in size and easy to use. They can be executed separately or commonly. A set of codes has a user-friendly interface. Transport channel construction consists of focusing lenses (doublets and triplets). The main of the magneto-optical channel parameters are total length and lens position and parameters of the output beam in the phase space (channel acceptance, beam emittance - beam transverse dimensions, particles divergence and image stigmaticity). Choice of the channel operation parameters is based on the conditions for satisfying mutually competing demands. And therefore the channel parameters calculation is carried out by using the search engine optimization techniques.

  16. Effects of scattering anisotropy approximation in multigroup radiation shielding calculations

    International Nuclear Information System (INIS)

    Altiparmakov, D.

    1983-01-01

    Expansion of the scattering cross sections into Legendre series is the usual way of solving neutron transport problems. Because of the large space gradients of the neutron flux, the effects of that approximation become especially remarkable in the radiation shielding calculations. In this paper, a method taking into account the scattering anisotropy is presented. From the point od view of the accuracy and computing rate, the optimal approximation of the scattering anisotropy is established for the basic protective materials on the basis of simple problem calculations. (author)

  17. 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)

  18. Study of the accuracy of radiation field calculations in media

    International Nuclear Information System (INIS)

    Bolyatko, V.V.; Vyrskij, M.Yu.; Ilyushkin, A.I.; Mashkovich, V.P.; Sakharov, V.K.; Stroganov, A.A.

    1981-01-01

    The sensitivity p of the radiation transport calculations to variations of input parameters Xsub(i) is theoretically analyzed, and the calculational errors induced by uncertainties of initial data are evaluated. Two calculational methods are considered: the direct substitution method using the ROZ-5 code and method using the linear perturbation theory. In order to calculate p(Xsub(i)) and bilinear convolutions of the conjugated transport equations the ZAKAT code has been developed. The calculations use the ZAKAT, ROZ-11 and APAMAKO-2F codes. As an example of practical use of the method proposed a shielding composition characteristic for fast reactors was analyzed. A plane monodirectional neutron beam of the BR-10 reactor falls onto a 5-layer stainless steel (1Kh18N10T)-carbon barrier. The sensitivily of the neutron dose absorbed in tissue to the cross sections of all the shielding constituents and to the source and detector representation functions has been calculated. A comparison of the calculations with experimental data proves the validity of the calculational method [ru

  19. Radiative forcing calculations for CH3Cl

    International Nuclear Information System (INIS)

    Grossman, A.S.; Grant, K.E.; Wuebbles, D.J.

    1994-06-01

    Methyl chloride, CH 3 Cl, is the major natural source of chlorine to the stratosphere. The production of CH 3 Cl is dominated by biological sources from the oceans and biomass burning. Production has a seasonal cycle which couples with the short lifetime of tropospheric CH 3 Cl to produce nonuniform global mixing. As an absorber of infrared radiation, CH 3 Cl is of interest for its potential affect on the tropospheric energy balance as well as for its chemical interactions. In this study, we estimate the radiative forcing and global warming potential (GWP) of CH 3 Cl. Our calculations use an infrared radiative transfer model based on the correlated k-distribution algorithm for band absorption. Global and annual average vertical profiles of temperature and trace gas concentration were assumed. The effects of clouds are modeled using three layers of global and annual average cloud optical properties. A radiative forcing value of 0.0053 W/m 2 ppbv was obtained for CH 3 Cl and is approximately linear in the background abundance. This value is about 2 percent of the forcing of CFC-11 and about 300 times the forcing of CO 2 , on a per molecule basis. The radiative forcing calculation for CH 3 Cl is used to estimate the global warming potential (GWP) of CH 3 Cl. The results give GWPs for CH 3 Cl of the order of 25 at a time of 20 years(CO 2 = 1). This result indicates that CH 3 Cl has the potential to be a major greenhouse gas if significant human related emissions were introduced into the atmosphere

  20. Agriculture-related radiation dose calculations

    International Nuclear Information System (INIS)

    Furr, J.M.; Mayberry, J.J.; Waite, D.A.

    1987-10-01

    Estimates of radiation dose to the public must be made at each stage in the identification and qualification process leading to siting a high-level nuclear waste repository. Specifically considering the ingestion pathway, this paper examines questions of reliability and adequacy of dose calculations in relation to five stages of data availability (geologic province, region, area, location, and mass balance) and three methods of calculation (population, population/food production, and food production driven). Calculations were done using the model PABLM with data for the Permian and Palo Duro Basins and the Deaf Smith County area. Extra effort expended in gathering agricultural data at succeeding environmental characterization levels does not appear justified, since dose estimates do not differ greatly; that effort would be better spent determining usage of food types that contribute most to the total dose; and that consumption rate and the air dispersion factor are critical to assessment of radiation dose via the ingestion pathway. 17 refs., 9 figs., 32 tabs

  1. Monte Carlo calculations of channeling radiation

    International Nuclear Information System (INIS)

    Bloom, S.D.; Berman, B.L.; Hamilton, D.C.; Alguard, M.J.; Barrett, J.H.; Datz, S.; Pantell, R.H.; Swent, R.H.

    1981-01-01

    Results of classical Monte Carlo calculations are presented for the radiation produced by ultra-relativistic positrons incident in a direction parallel to the (110) plane of Si in the energy range 30 to 100 MeV. The results all show the characteristic CR(channeling radiation) peak in the energy range 20 keV to 100 keV. Plots of the centroid energies, widths, and total yields of the CR peaks as a function of energy show the power law dependences of γ 1 5 , γ 1 7 , and γ 2 5 respectively. Except for the centroid energies and power-law dependence is only approximate. Agreement with experimental data is good for the centroid energies and only rough for the widths. Adequate experimental data for verifying the yield dependence on γ does not yet exist

  2. New nonlinear methods for linear transport calculations

    International Nuclear Information System (INIS)

    Adams, M.L.

    1993-01-01

    We present a new family of methods for the numerical solution of the linear transport equation. With these methods an iteration consists of an 'S N sweep' followed by an 'S 2 -like' calculation. We show, by analysis as well as numerical results, that iterative convergence is always rapid. We show that this rapid convergence does not depend on a consistent discretization of the S 2 -like equations - they can be discretized independently from the S N equations. We show further that independent discretizations can offer significant advantages over consistent ones. In particular, we find that in a wide range of problems, an accurate discretization of the S 2 -like equation can be combined with a crude discretization of the S N equations to produce an accurate S N answer. We demonstrate this by analysis as well as numerical results. (orig.)

  3. Radiation damage calculations for the LANSCE degrader

    International Nuclear Information System (INIS)

    Ferguson, P.D.; Sommer, W.F.; Dudziak, D.J.; Wechsler, M.S.; Barnett, M.H.; Corzine, R.K.

    1998-01-01

    The A-6 water degrader at the Los Alamos Neutron Science Center (LANSCE) linear proton accelerator has an outer shell of Inconel 718. The degrader was irradiated by 800-MeV protons during 1988--1993 to an exposure of 5.3 ampere-hours (A h). As described in Ref. 1, material from the Inconel is currently being cut into specimens for microhardness, three-point bending, ball punch, microscopy, and corrosion tests. This paper is devoted to calculations of radiation damage, particularly displacement and He production, sustained by the degrader Inconel

  4. 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.

  5. A solution algorithm for calculating photon radiation fields with the aid of the Monte Carlo method

    International Nuclear Information System (INIS)

    Zappe, D.

    1978-04-01

    The MCTEST program and its subroutines for the solution of the Boltzmann transport equation is presented. The program renders possible to calculate photon radiation fields of point or plane gamma sources. After changing two subroutines the calculation can also be carried out for the case of directed incidence of radiation on plane shields of iron or concrete. (author)

  6. ASOP, Shield Calculation, 1-D, Discrete Ordinates Transport

    International Nuclear Information System (INIS)

    1993-01-01

    1 - Nature of physical problem solved: ASOP is a shield optimization calculational system based on the one-dimensional discrete ordinates transport program ANISN. It has been used to design optimum shields for space applications of SNAP zirconium-hydride-uranium- fueled reactors and uranium-oxide fueled thermionic reactors and to design beam stops for the ORELA facility. 2 - Method of solution: ASOP generates coefficients of linear equations describing the logarithm of the dose and dose-weight derivatives as functions of position from data obtained in an automated sequence of ANISN calculations. With the dose constrained to a design value and all dose-weight derivatives required to be equal, the linear equations may be solved for a new set of shield dimensions. Since changes in the shield dimensions may cause the linear functions to change, the entire procedure is repeated until convergence is obtained. The detailed calculations of the radiation transport through shield configurations for every step in the procedure distinguish ASOP from other shield optimization computer code systems which rely on multiple component sources and attenuation coefficients to describe the transport. 3 - Restrictions on the complexity of the problem: Problem size is limited only by machine size

  7. Review of the radiation protection calculations for the encapsulation plant

    International Nuclear Information System (INIS)

    Ranta-aho, A.

    2008-09-01

    The radiation protection calculations of the encapsulation plant have been carried out with the MCNP5 Monte Carlo code. The focus of the study has been in the parts of the encapsulation plant where the spent fuel is handled after discharge from the transportation casks i.e. the fuel handling cell, the fuel drying station, the canister transfer corridor, the welding chamber, the weld inspection room, the canister buffer storage and the canister lift. The protection against radiation hazard has been mainly designed with thick concrete walls. Additionally, the entrances to the rooms with shielding requirements have been equipped with mazes. The present design excludes doors with shielding properties. The aim of this work was to verify and evaluate the necessary wall thicknesses and the functioning of the mazes in the current design. The calculations verified that for the most parts of the facility, the currently designed walls thicknesses provide adequate protection against radiation from the different spent fuel assembly configurations. Some corrective actions however seem necessary in order to stay clearly below desired radiation limits. For the most parts the functioning of the mazes was inadequate. In some of the cases a different design of the maze will be sufficient action but in some cases the radiation protection can only be secured by heavy doors for practical reasons. (orig.)

  8. Neutron and gamma ray transport calculations in shielding system

    Energy Technology Data Exchange (ETDEWEB)

    Masukawa, Fumihiro; Sakamoto, Hiroki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    In the shields for radiation in nuclear facilities, the penetrating holes of various kinds and irregular shapes are made for the reasons of operation, control and others. These penetrating holes and gaps are filled with air or the substances with relatively small shielding performance, and radiation flows out through them, which is called streaming. As the calculation techniques for the shielding design or analysis related to the streaming problem, there are the calculations by simplified evaluation, transport calculation and Monte Carlo method. In this report, the example of calculation by Monte Carlo method which is represented by MCNP code is discussed. A number of variance reduction techniques which seem effective for the analysis of streaming problem were tried. As to the investigation of the applicability of MCNP code to streaming analysis, the object of analysis which are the concrete walls without hole and with horizontal hole, oblique hole and bent oblique hole, the analysis procedure, the composition of concrete, and the conversion coefficient of dose equivalent, and the results of analysis are reported. As for variance reduction technique, cell importance was adopted. (K.I.)

  9. Radiation shielding calculations for the vista spacecraft

    International Nuclear Information System (INIS)

    Sahin, Suemer; Sahin, Haci Mehmet; Acir, Adem

    2005-01-01

    The VISTA spacecraft design concept has been proposed for manned or heavy cargo deep space missions beyond earth orbit with inertial fusion energy propulsion. Rocket propulsion is provided by fusion power deposited in the inertial confined fuel pellet debris and with the help of a magnetic nozzle. The calculations for the radiation shielding have been revised under the fact that the highest jet efficiency of the vehicle could be attained only if the propelling plasma would have a narrow temperature distribution. The shield mass could be reduced from 600 tons in the original design to 62 tons. Natural and enriched lithium were the principle shielding materials. The allowable nuclear heating in the superconducting magnet coils (up to 5 mW/cm 3 ) is taken as the crucial criterion for dimensioning the radiation shielding structure of the spacecraft. The space craft mass is 6000 tons. Total peak nuclear power density in the coils is calculated as ∼5.0 mW/cm 3 for a fusion power output of 17 500 MW. The peak neutron heating density is ∼2.0 mW/cm 3 , and the peak γ-ray heating density is ∼3.0 mW/cm 3 (on different points) using natural lithium in the shielding. However, the volume averaged heat generation in the coils is much lower, namely 0.21, 0.71 and 0.92 mW/cm 3 for the neutron, γ-ray and total nuclear heating, respectively. The coil heating will be slightly lower if highly enriched 6 Li (90%) is used instead of natural lithium. Peak values are then calculated as 2.05, 2.15 and 4.2 mW/cm 3 for the neutron, γ-ray and total nuclear heating, respectively. The corresponding volume averaged heat generation in the coils became 0.19, 0.58 and 0.77 mW/cm 3

  10. 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.

  11. 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

  12. Radiative forcing calculations for CH3Br

    International Nuclear Information System (INIS)

    Grossman, A.S.; Blass, W.E.; Wuebbles, D.J.

    1995-06-01

    Methyl Bromide, CH 3 Br, is the major organobromine species in the lower atmosphere and is a primary source of bromine in the stratosphere. It has a lifetime of 1.3 years. The IR methyl bromide spectra in the atmospheric window region, 7--13μ, was determined using a well tested Coriolis resonance and ell-doubling (and ell-resonance) computational system. A radiative forcing value of 0.00493 W/m 2 /ppbv was obtained for CH 3 Br and is approximately linear in the background abundance. This value is about 2 percent of the forcing of CFC-11 and about 278 times the forcing of C0 2 , on a per molecule basis. The radiative forcing calculation is used to estimate the global warming potential (GWP) of CH 3 Br. The results give GWPs for CH 3 Br of the order of 13 for an integration period of 20 years and 4 for an integration period of 100 years (assuming C0 2 = 1, following IPCC [1994]). While CH 3 Br has a GWP which is approximately 25 percent of the GWP of CH 4 , the current emission rates are too low to cause serious atmospheric greenhouse heating effects at this time

  13. 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)

  14. A retrospective and prospective survey of three-dimensional transport calculations

    International Nuclear Information System (INIS)

    Nakahara, Yasuaki

    1985-01-01

    A retrospective survey is made on the three-dimensional radiation transport calculations. Introduction is given to computer codes based on the distinctive numerical methods such as the Monte Carlo, Direct Integration, Ssub(n) and Finite Element Methods to solve the three-dimensional transport equations. Prospective discussions are made on pros and cons of these methods. (author)

  15. Mathematical models for calculating radiation dose to the fetus

    International Nuclear Information System (INIS)

    Watson, E.E.

    1992-01-01

    Estimates of radiation dose from radionuclides inside the body are calculated on the basis of energy deposition in mathematical models representing the organs and tissues of the human body. Complex models may be used with radiation transport codes to calculate the fraction of emitted energy that is absorbed in a target tissue even at a distance from the source. Other models may be simple geometric shapes for which absorbed fractions of energy have already been calculated. Models of Reference Man, the 15-year-old (Reference Woman), the 10-year-old, the five-year-old, the one-year-old, and the newborn have been developed and used for calculating specific absorbed fractions (absorbed fractions of energy per unit mass) for several different photon energies and many different source-target combinations. The Reference woman model is adequate for calculating energy deposition in the uterus during the first few weeks of pregnancy. During the course of pregnancy, the embryo/fetus increases rapidly in size and thus requires several models for calculating absorbed fractions. In addition, the increases in size and changes in shape of the uterus and fetus result in the repositioning of the maternal organs and in different geometric relationships among the organs and the fetus. This is especially true of the excretory organs such as the urinary bladder and the various sections of the gastrointestinal tract. Several models have been developed for calculating absorbed fractions of energy in the fetus, including models of the uterus and fetus for each month of pregnancy and complete models of the pregnant woman at the end of each trimester. In this paper, the available models and the appropriate use of each will be discussed. (Author) 19 refs., 7 figs

  16. Calculation of transport coefficients in an axisymmetric plasma

    International Nuclear Information System (INIS)

    Shumaker, D.E.

    1977-01-01

    A method of calculating the transport coefficient in an axisymmetric toroidal plasma is presented. This method is useful in calculating the transport coefficients in a Tokamak plasma confinement device. The particle density and temperature are shown to be a constant on a magnetic flux surface. Transport equations are given for the total particle flux and total energy flux crossing a closed toroidal surface. Also transport equations are given for the toroidal magnetic flux. A computer code was written to calculate the transport coefficients for a three species plasma, electrons and two species of ions. This is useful for calculating the transport coefficients of a plasma which contains impurities. It was found that the particle and energy transport coefficients are increased by a large amount, and the transport coefficients for the toroidal magnetic field are reduced by a small amount

  17. Uncertainty calculation in transport models and forecasts

    DEFF Research Database (Denmark)

    Manzo, Stefano; Prato, Carlo Giacomo

    Transport projects and policy evaluations are often based on transport model output, i.e. traffic flows and derived effects. However, literature has shown that there is often a considerable difference between forecasted and observed traffic flows. This difference causes misallocation of (public...... implemented by using an approach based on stochastic techniques (Monte Carlo simulation and Bootstrap re-sampling) or scenario analysis combined with model sensitivity tests. Two transport models are used as case studies: the Næstved model and the Danish National Transport Model. 3 The first paper...... in a four-stage transport model related to different variable distributions (to be used in a Monte Carlo simulation procedure), assignment procedures and levels of congestion, at both the link and the network level. The analysis used as case study the Næstved model, referring to the Danish town of Næstved2...

  18. Methods for calculating radiation attenuation in shields

    Energy Technology Data Exchange (ETDEWEB)

    Butler, J; Bueneman, D; Etemad, A; Lafore, P; Moncassoli, A M; Penkuhn, H; Shindo, M; Stoces, B

    1964-10-01

    In recent years the development of high-speed digital computers of large capacity has revolutionized the field of reactor shield design. For compact special-purpose reactor shields, Monte-Carlo codes in two- and three dimensional geometries are now available for the proper treatment of both the neutron and gamma- ray problems. Furthermore, techniques are being developed for the theoretical optimization of minimum-weight shield configurations for this type of reactor system. In the design of land-based power reactors, on the other hand, there is a strong incentive to reduce the capital cost of the plant, and economic considerations are also relevant to reactors designed for merchant ship propulsion. In this context simple methods are needed which are economic in their data input and computing time requirements and which, at the same time, are sufficiently accurate for design work. In general the computing time required for Monte-Carlo calculations in complex geometry is excessive for routine design calculations and the capacity of the present codes is inadequate for the proper treatment of large reactor shield systems in three dimensions. In these circumstances a wide range of simpler techniques are currently being employed for design calculations. The methods of calculation for neutrons in reactor shields fall naturally into four categories: Multigroup diffusion theory; Multigroup diffusion with removal sources; Transport codes; and Monte Carlo methods. In spite of the numerous Monte- Carlo techniques which are available for penetration and back scattering, serious problems are still encountered in practice with the scattering of gamma rays from walls of buildings which contain critical facilities and also concrete-lined discharge shafts containing irradiated fuel elements. The considerable volume of data in the unclassified literature on the solution of problems of this type in civil defence work appears not to have been evaluated for reactor shield design. In

  19. The internal radiation dose calculations based on Chinese mathematical phantom

    International Nuclear Information System (INIS)

    Wang Haiyan; Li Junli; Cheng Jianping; Fan Jiajin

    2006-01-01

    The internal radiation dose calculations built on Chinese facts become more and more important according to the development of nuclear medicine. the MIRD method developed and consummated by the society of Nuclear Medicine (America) is based on the European and American mathematical phantom and can't fit Chinese well. The transport of γ-ray in the Chinese mathematical phantom was simulated with Monte Carlo method in programs as MCNP4C. the specific absorbed fraction (Φ) of Chinese were calculated and the Chinese Φ database was created. The results were compared with the recommended values by ORNL. the method was proved correct by the coherence when the target organ was the same with the source organ. Else, the difference was due to the different phantom and the choice of different physical model. (authors)

  20. Radiation doses from radiation sources of neutrons and photons by different computer calculation

    International Nuclear Information System (INIS)

    Siciliano, F.; Lippolis, G.; Bruno, S.G.

    1995-12-01

    In the present paper the calculation technique aspects of dose rate from neutron and photon radiation sources are covered with reference both to the basic theoretical modeling of the MERCURE-4, XSDRNPM-S and MCNP-3A codes and from practical point of view performing safety analyses of irradiation risk of two transportation casks. The input data set of these calculations -regarding the CEN 10/200 HLW container and dry PWR spent fuel assemblies shipping cask- is frequently commented as for as connecting points of input data and understanding theoric background are concerned

  1. Parallel MCNP Monte Carlo transport calculations with MPI

    International Nuclear Information System (INIS)

    Wagner, J.C.; Haghighat, A.

    1996-01-01

    The steady increase in computational performance has made Monte Carlo calculations for large/complex systems possible. However, in order to make these calculations practical, order of magnitude increases in performance are necessary. The Monte Carlo method is inherently parallel (particles are simulated independently) and thus has the potential for near-linear speedup with respect to the number of processors. Further, the ever-increasing accessibility of parallel computers, such as workstation clusters, facilitates the practical use of parallel Monte Carlo. Recognizing the nature of the Monte Carlo method and the trends in available computing, the code developers at Los Alamos National Laboratory implemented the message-passing general-purpose Monte Carlo radiation transport code MCNP (version 4A). The PVM package was chosen by the MCNP code developers because it supports a variety of communication networks, several UNIX platforms, and heterogeneous computer systems. This PVM version of MCNP has been shown to produce speedups that approach the number of processors and thus, is a very useful tool for transport analysis. Due to software incompatibilities on the local IBM SP2, PVM has not been available, and thus it is not possible to take advantage of this useful tool. Hence, it became necessary to implement an alternative message-passing library package into MCNP. Because the message-passing interface (MPI) is supported on the local system, takes advantage of the high-speed communication switches in the SP2, and is considered to be the emerging standard, it was selected

  2. 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.)

  3. 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

  4. 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)

  5. Guide to calculating transportation demand management benefits

    Energy Technology Data Exchange (ETDEWEB)

    Litman, T. [Victoria Transport Policy Institute, Victoria, BC (Canada)

    1997-02-14

    The full benefits of transportation demand management (TDM) programs were discussed. TDM includes several policies, programs and measures designed to change travel patterns. TDM programs include commute trip reductions, pricing policies, land use management strategies, and programs to support alternative modes of transportation such as public transit, carpooling, bicycling, walking and telecommuting. In addition to reduction in traffic congestion and reduction in air pollution, other impacts of TDM programs were also evaluated. The value of these impacts based on external cost savings was estimated. A list of documents, software and organizations which could be helpful for TDM planning and evaluation was provided. 34 refs., 14 tabs., 1 fig.

  6. Calculation of transport coefficients in an axisymmetric plasma

    International Nuclear Information System (INIS)

    Shumaker, D.E.

    1976-01-01

    A method of calculating the transport coefficient in an axisymmetric toroidal plasma is presented. This method is useful in calculating the transport coefficients in a Tokamak plasma confinement device. The particle density and temperature are shown to be a constant on a magnetic flux surface. Transport equations are given for the total particle flux and total energy flux crossing a closed toroidal surface. Also transport equations are given for the toroidal magnetic flux. A computer code was written to calculate the transport coefficients for a three species plasma, electrons and two species of ions. This is useful for calculating the transport coefficients of a plasma which contains impurities. It was found that the particle and energy transport coefficients are increased by a large amount, and the transport coefficients for the toroidal magnetic field are reduced by a small amount. For example, a deuterium plasma with 1.3 percent oxygen, one of the particle transport coefficients is increased by a factor of about four. The transport coefficients for the toroidal magnetic flux are reduced by about 20 percent. The increase in the particle transport coefficient is due to the collisional scattering of the deuterons by the heavy oxygen ions which is larger than the deuteron electron scattering, the normal process for particle transport in a two species plasma. The reduction in the toroidal magnetic flux transport coefficients are left unexplained

  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. Acceleration methods for assembly-level transport calculations

    International Nuclear Information System (INIS)

    Adams, Marvin L.; Ramone, Gilles

    1995-01-01

    A family acceleration methods for the iterations that arise in assembly-level transport calculations is presented. A single iteration in these schemes consists of a transport sweep followed by a low-order calculation which is itself a simplified transport problem. It is shown that a previously-proposed method fitting this description is unstable in two and three dimensions. It is presented a family of methods and shown that some members are unconditionally stable. (author). 8 refs, 4 figs, 4 tabs

  9. Considerations of beta and electron transport in internal dose calculations

    International Nuclear Information System (INIS)

    Bolch, W.E.; Poston, J.W. Sr.

    1990-12-01

    Ionizing radiation has broad uses in modern science and medicine. These uses often require the calculation of energy deposition in the irradiated media and, usually, the medium of interest is the human body. Energy deposition from radioactive sources within the human body and the effects of such deposition are considered in the field of internal dosimetry. In July of 1988, a three-year research project was initiated by the Nuclear Engineering Department at Texas A ampersand M University under the sponsorship of the US Department of Energy. The main thrust of the research was to consider, for the first time, the detailed spatial transport of electron and beta particles in the estimation of average organ doses under the Medical Internal Radiation Dose (MIRD) schema. At the present time (December of 1990), research activities are continuing within five areas. Several are new initiatives begun within the second or third year of the current contract period. They include: (1) development of small-scale dosimetry; (2) development of a differential volume phantom; (3) development of a dosimetric bone model; (4) assessment of the new ICRP lung model; and (5) studies into the mechanisms of DNA damage. A progress report is given for each of these tasks within the Comprehensive Report. In each use, preliminary results are very encouraging and plans for further research are detailed within this document. 22 refs., 13 figs., 1 tab

  10. Considerations of beta and electron transport in internal dose calculations

    Energy Technology Data Exchange (ETDEWEB)

    Bolch, W.E.; Poston, J.W. Sr.

    1990-12-01

    Ionizing radiation has broad uses in modern science and medicine. These uses often require the calculation of energy deposition in the irradiated media and, usually, the medium of interest is the human body. Energy deposition from radioactive sources within the human body and the effects of such deposition are considered in the field of internal dosimetry. In July of 1988, a three-year research project was initiated by the Nuclear Engineering Department at Texas A M University under the sponsorship of the US Department of Energy. The main thrust of the research was to consider, for the first time, the detailed spatial transport of electron and beta particles in the estimation of average organ doses under the Medical Internal Radiation Dose (MIRD) schema. At the present time (December of 1990), research activities are continuing within five areas. Several are new initiatives begun within the second or third year of the current contract period. They include: (1) development of small-scale dosimetry; (2) development of a differential volume phantom; (3) development of a dosimetric bone model; (4) assessment of the new ICRP lung model; and (5) studies into the mechanisms of DNA damage. A progress report is given for each of these tasks within the Comprehensive Report. In each case, preliminary results are very encouraging and plans for further research are detailed within this document.

  11. Considerations of beta and electron transport in internal dose calculations

    Energy Technology Data Exchange (ETDEWEB)

    Bolch, W.E.; Poston, J.W. Sr. (Texas A and M Univ., College Station, TX (USA). Dept. of Nuclear Engineering)

    1990-12-01

    Ionizing radiation has broad uses in modern science and medicine. These uses often require the calculation of energy deposition in the irradiated media and, usually, the medium of interest is the human body. Energy deposition from radioactive sources within the human body and the effects of such deposition are considered in the field of internal dosimetry. In July of 1988, a three-year research project was initiated by the Nuclear Engineering Department at Texas A M University under the sponsorship of the US Department of Energy. The main thrust of the research was to consider, for the first time, the detailed spatial transport of electron and beta particles in the estimation of average organ doses under the Medical Internal Radiation Dose (MIRD) schema. At the present time (December of 1990), research activities are continuing within five areas. Several are new initiatives begun within the second or third year of the current contract period. They include: (1) development of small-scale dosimetry; (2) development of a differential volume phantom; (3) development of a dosimetric bone model; (4) assessment of the new ICRP lung model; and (5) studies into the mechanisms of DNA damage. A progress report is given for each of these tasks within the Comprehensive Report. In each use, preliminary results are very encouraging and plans for further research are detailed within this document. 22 refs., 13 figs., 1 tab.

  12. Considerations of beta and electron transport in internal dose calculations

    International Nuclear Information System (INIS)

    Bolch, W.E.; Poston, J.W. Sr.

    1990-12-01

    Ionizing radiation has broad uses in modern science and medicine. These uses often require the calculation of energy deposition in the irradiated media and, usually, the medium of interest is the human body. Energy deposition from radioactive sources within the human body and the effects of such deposition are considered in the field of internal dosimetry. In July of 1988, a three-year research project was initiated by the Nuclear Engineering Department at Texas A ampersand M University under the sponsorship of the US Department of Energy. The main thrust of the research was to consider, for the first time, the detailed spatial transport of electron and beta particles in the estimation of average organ doses under the Medical Internal Radiation Dose (MIRD) schema. At the present time (December of 1990), research activities are continuing within five areas. Several are new initiatives begun within the second or third year of the current contract period. They include: (1) development of small-scale dosimetry; (2) development of a differential volume phantom; (3) development of a dosimetric bone model; (4) assessment of the new ICRP lung model; and (5) studies into the mechanisms of DNA damage. A progress report is given for each of these tasks within the Comprehensive Report. In each case, preliminary results are very encouraging and plans for further research are detailed within this document

  13. Radiation damage calculations for compound materials

    International Nuclear Information System (INIS)

    Greenwood, L.R.

    1990-01-01

    This paper reports on the SPECOMP computer code, developed to calculate neutron-induced displacement damage cross sections for compound materials such as alloys, insulators, and ceramic tritium breeders for fusion reactors. These new calculations rely on recoil atom energy distributions previously computed with the DISCS computer code, the results of which are stored in SPECTER computer code master libraries. All reaction channels were considered in the DISCS calculations and the neutron cross sections were taken from ENDF/B-V. Compound damage calculations with SPECOMP thus do not need to perform any recoil atom calculations and consequently need no access to ENDF or other neutron data bases. The calculations proceed by determining secondary displacements for each combination of recoil atom and matrix atom using the Lindhard partition of the recoil energy to establish the damage energy

  14. Nonlinear acceleration of SN transport calculations

    Energy Technology Data Exchange (ETDEWEB)

    Fichtl, Erin D [Los Alamos National Laboratory; Warsa, James S [Los Alamos National Laboratory; Calef, Matthew T [Los Alamos National Laboratory

    2010-12-20

    The use of nonlinear iterative methods, Jacobian-Free Newton-Krylov (JFNK) in particular, for solving eigenvalue problems in transport applications has recently become an active subject of research. While JFNK has been shown to be effective for k-eigenvalue problems, there are a number of input parameters that impact computational efficiency, making it difficult to implement efficiently in a production code using a single set of default parameters. We show that different selections for the forcing parameter in particular can lead to large variations in the amount of computational work for a given problem. In contrast, we present a nonlinear subspace method that sits outside and effectively accelerates nonlinear iterations of a given form and requires only a single input parameter, the subspace size. It is shown to consistently and significantly reduce the amount of computational work when applied to fixed-point iteration, and this combination of methods is shown to be more efficient than JFNK for our application.

  15. Mesh requirements for neutron transport calculations

    International Nuclear Information System (INIS)

    Askew, J.R.

    1967-07-01

    Fine-structure calculations are reported for a cylindrical natural uranium-graphite cell using different solution methods (discrete ordinate and collision probability codes) and varying the spatial mesh. It is suggested that of formulations assuming the source constant in a mesh interval the differential approach is generally to be preferred. Due to cancellation between approximations made in the derivation of the finite difference equations and the errors in neglecting source variation, the discrete ordinate code gave a more accurate estimate of fine structure for a given mesh even for unusually coarse representations. (author)

  16. 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

  17. Radiative flux calculations at UV and visible wavelengths

    International Nuclear Information System (INIS)

    Grossman, A.S.; Grant, K.E.; Wuebbles, D.J.

    1993-10-01

    A radiative transfer model to calculate the short wavelength fluxes at altitudes between 0 and 80 km has been developed at LLNL. The wavelength range extends from 175--735 nm. This spectral range covers the UV-B wavelength region, 250--350 nm, with sufficient resolution to allow comparison of UV-B measurements with theoretical predictions. Validation studies for the model have been made for both UV-B ground radiation calculations and tropospheric solar radiative forcing calculations for various ozone distributions. These studies indicate that the model produces results which agree well with respect to existing UV calculations from other published models

  18. 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

  19. Radiation damage calculations for compound materials

    International Nuclear Information System (INIS)

    Greenwood, L.R.

    1989-01-01

    Displacement damage calculations can be performed for 40 elements in the energy range up to 20 MeV with the SPECTER computer code. A recent addition to the code, called SPECOMP, can intermix atomic recoil energy distributions for any four elements to calculate the proper displacement damage for compound materials. The calculations take advantage of the atomic recoil data in the SPECTER libraries, which were determined by the DISCS computer code, using evaluated neutron cross section and angular distribution data in ENDF/B-V. Resultant damage cross sections for any compound can be added to the SPECTER libraries for the routine calculation of displacements in any given neutron field. Users do not require access to neutron cross section files. Results are presented for a variety of fusion materials and a new ceramic superconductor material. Future plans and nuclear data needs are discussed. 11 refs., 6 figs., 1 tab

  20. Generalized diffusion theory for calculating the neutron transport scalar flux

    International Nuclear Information System (INIS)

    Alcouffe, R.E.

    1975-01-01

    A generalization of the neutron diffusion equation is introduced, the solution of which is an accurate approximation to the transport scalar flux. In this generalization the auxiliary transport calculations of the system of interest are utilized to compute an accurate, pointwise diffusion coefficient. A procedure is specified to generate and improve this auxiliary information in a systematic way, leading to improvement in the calculated diffusion scalar flux. This improvement is shown to be contingent upon satisfying the condition of positive calculated-diffusion coefficients, and an algorithm that ensures this positivity is presented. The generalized diffusion theory is also shown to be compatible with conventional diffusion theory in the sense that the same methods and codes can be used to calculate a solution for both. The accuracy of the method compared to reference S/sub N/ transport calculations is demonstrated for a wide variety of examples. (U.S.)

  1. 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)

  2. 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)

  3. 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.

  4. Ionizing radiation calculations and comparisons with LDEF data

    Science.gov (United States)

    Armstrong, T. W.; Colborn, B. L.; Watts, J. W., Jr.

    1992-01-01

    In conjunction with the analysis of LDEF ionizing radiation dosimetry data, a calculational program is in progress to aid in data interpretation and to assess the accuracy of current radiation models for future mission applications. To estimate the ionizing radiation environment at the LDEF dosimeter locations, scoping calculations for a simplified (one dimensional) LDEF mass model were made of the primary and secondary radiations produced as a function of shielding thickness due to trapped proton, galactic proton, and atmospheric (neutron and proton cosmic ray albedo) exposures. Preliminary comparisons of predictions with LDEF induced radioactivity and dose measurements were made to test a recently developed model of trapped proton anisotropy.

  5. Parallel SN transport calculations on a transputer network

    International Nuclear Information System (INIS)

    Kim, Yong Hee; Cho, Nam Zin

    1994-01-01

    A parallel computing algorithm for the neutron transport problems has been implemented on a transputer network and two reactor benchmark problems (a fixed-source problem and an eigenvalue problem) are solved. We have shown that the parallel calculations provided significant reduction in execution time over the sequential calculations

  6. Neutron transport calculations of some fast critical assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Martinez-Val Penalosa, J A

    1976-07-01

    To analyse the influence of the input variables of the transport codes upon the neutronic results (eigenvalues, generation times, . . . ) four Benchmark calculations have been performed. Sensitivity analysis have been applied to express these dependences in a useful way, and also to get an unavoidable experience to carry out calculations achieving the required accuracy and doing them in practical computing times. (Author) 29 refs.

  7. Calculations of the transport properties within the PAW formalism

    Energy Technology Data Exchange (ETDEWEB)

    Mazevet, S.; Torrent, M.; Recoules, V.; Jollet, F. [CEA Bruyeres-le-Chatel, DIF, 91 (France)

    2010-07-01

    We implemented the calculation of the transport properties within the PAW formalism in the ABINIT code. This feature allows the calculation of the electrical and optical properties, including the XANES spectrum, as well as the electronic contribution to the thermal conductivity. We present here the details of the implementation and results obtained for warm dense aluminum plasma. (authors)

  8. Neutron transport calculations of some fast critical assemblies

    International Nuclear Information System (INIS)

    Martinez-Val Penalosa, J. A.

    1976-01-01

    To analyse the influence of the input variables of the transport codes upon the neutronic results (eigenvalues, generation times, . . . ) four Benchmark calculations have been performed. Sensitivity analysis have been applied to express these dependences in a useful way, and also to get an unavoidable experience to carry out calculations achieving the required accuracy and doing them in practical computing times. (Author) 29 refs

  9. Minaret, a deterministic neutron transport solver for nuclear core calculations

    International Nuclear Information System (INIS)

    Moller, J-Y.; Lautard, J-J.

    2011-01-01

    We present here MINARET a deterministic transport solver for nuclear core calculations to solve the steady state Boltzmann equation. The code follows the multi-group formalism to discretize the energy variable. It uses discrete ordinate method to deal with the angular variable and a DGFEM to solve spatially the Boltzmann equation. The mesh is unstructured in 2D and semi-unstructured in 3D (cylindrical). Curved triangles can be used to fit the exact geometry. For the curved elements, two different sets of basis functions can be used. Transport solver is accelerated with a DSA method. Diffusion and SPN calculations are made possible by skipping the transport sweep in the source iteration. The transport calculations are parallelized with respect to the angular directions. Numerical results are presented for simple geometries and for the C5G7 Benchmark, JHR reactor and the ESFR (in 2D and 3D). Straight and curved finite element results are compared. (author)

  10. Minaret, a deterministic neutron transport solver for nuclear core calculations

    Energy Technology Data Exchange (ETDEWEB)

    Moller, J-Y.; Lautard, J-J., E-mail: jean-yves.moller@cea.fr, E-mail: jean-jacques.lautard@cea.fr [CEA - Centre de Saclay , Gif sur Yvette (France)

    2011-07-01

    We present here MINARET a deterministic transport solver for nuclear core calculations to solve the steady state Boltzmann equation. The code follows the multi-group formalism to discretize the energy variable. It uses discrete ordinate method to deal with the angular variable and a DGFEM to solve spatially the Boltzmann equation. The mesh is unstructured in 2D and semi-unstructured in 3D (cylindrical). Curved triangles can be used to fit the exact geometry. For the curved elements, two different sets of basis functions can be used. Transport solver is accelerated with a DSA method. Diffusion and SPN calculations are made possible by skipping the transport sweep in the source iteration. The transport calculations are parallelized with respect to the angular directions. Numerical results are presented for simple geometries and for the C5G7 Benchmark, JHR reactor and the ESFR (in 2D and 3D). Straight and curved finite element results are compared. (author)

  11. Transport calculations for a 14.8 MeV neutron beam in a water phantom

    International Nuclear Information System (INIS)

    Goetsch, S.J.

    1981-01-01

    A coupled neutron/photon Monte Carlo radiation transport code (MORSE-CG) has been used to calculate neutron and photon doses in a water phantom irradiated by 14.8 MeV neutrons from the Gas Target Neutron Source. The source-collimator-phantom geometry was carefully simulated. Results of calculations utilizing two different statistical estimators (next-collision and track-length) are presented

  12. A grey diffusion acceleration method for time-dependent radiative transfer calculations: analysis and application

    International Nuclear Information System (INIS)

    Nowak, P.F.

    1993-01-01

    A grey diffusion acceleration method is presented and is shown by Fourier analysis and test calculations to be effective in accelerating radiative transfer calculations. The spectral radius is bounded by 0.9 for the continuous equations, but is significantly smaller for the discretized equations, especially in the optically thick regimes characteristic to radiation transport problems. The GDA method is more efficient than the multigroup DSA method because its slightly higher iteration count is more than offset by the much lower cost per iteration. A wide range of test calculations confirm the efficiency of GDA compared to multifrequency DSA. (orig.)

  13. 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.)

  14. Bivariational calculations for radiation transfer in an inhomogeneous participating media

    International Nuclear Information System (INIS)

    El Wakil, S.A.; Machali, H.M.; Haggag, M.H.; Attia, M.T.

    1986-07-01

    Equations for radiation transfer are obtained for dispersive media with space dependent albedo. Bivariational bound principle is used to calculate the reflection and transmission coefficients for such media. Numerical results are given and compared. (author)

  15. Use of implicit Monte Carlo radiation transport with hydrodynamics and compton scattering

    International Nuclear Information System (INIS)

    Fleck, J.A. Jr.

    1971-03-01

    It is shown that the combination of implicit radiation transport and hydrodynamics, Compton scattering, and any other energy transport can be simply carried out by a ''splitting'' procedure. Contributions to material energy exchange can be reckoned separately for hydrodynamics, radiation transport without scattering, Compton scattering, plus any other possible energy exchange mechanism. The radiation transport phase of the calculation would be implicit, but the hydrodynamics and Compton portions would not, leading to possible time step controls. The time step restrictions which occur on radiation transfer due to large Planck mean absorption cross-sections would not occur

  16. Parallel processing of neutron transport in fuel assembly calculation

    International Nuclear Information System (INIS)

    Song, Jae Seung

    1992-02-01

    Group constants, which are used for reactor analyses by nodal method, are generated by fuel assembly calculations based on the neutron transport theory, since one or a quarter of the fuel assembly corresponds to a unit mesh in the current nodal calculation. The group constant calculation for a fuel assembly is performed through spectrum calculations, a two-dimensional fuel assembly calculation, and depletion calculations. The purpose of this study is to develop a parallel algorithm to be used in a parallel processor for the fuel assembly calculation and the depletion calculations of the group constant generation. A serial program, which solves the neutron integral transport equation using the transmission probability method and the linear depletion equation, was prepared and verified by a benchmark calculation. Small changes from the serial program was enough to parallelize the depletion calculation which has inherent parallel characteristics. In the fuel assembly calculation, however, efficient parallelization is not simple and easy because of the many coupling parameters in the calculation and data communications among CPU's. In this study, the group distribution method is introduced for the parallel processing of the fuel assembly calculation to minimize the data communications. The parallel processing was performed on Quadputer with 4 CPU's operating in NURAD Lab. at KAIST. Efficiencies of 54.3 % and 78.0 % were obtained in the fuel assembly calculation and depletion calculation, respectively, which lead to the overall speedup of about 2.5. As a result, it is concluded that the computing time consumed for the group constant generation can be easily reduced by parallel processing on the parallel computer with small size CPU's

  17. To the development of numerical methods in problems of radiation transport

    International Nuclear Information System (INIS)

    Germogenova, T.A.

    1990-01-01

    Review of studies on the development of numerical methods and the discrete ordinate method in particular, used for solution of radiation protection physics problems is given. Consideration is given to the problems, which arise when calculating fields of penetrating radiation and when studying processes of charged-particle transport and cascade processes, generated by high-energy primary radiation

  18. Comparison of the performance of net radiation calculation models

    DEFF Research Database (Denmark)

    Kjærsgaard, Jeppe Hvelplund; Cuenca, R.H.; Martinez-Cob, A.

    2009-01-01

    . The long-wave radiation models included a physically based model, an empirical model from the literature, and a new empirical model. Both empirical models used only solar radiation as required for meteorological input. The long-wave radiation models were used with model calibration coefficients from......Daily values of net radiation are used in many applications of crop-growth modeling and agricultural water management. Measurements of net radiation are not part of the routine measurement program at many weather stations and are commonly estimated based on other meteorological parameters. Daily...... values of net radiation were calculated using three net outgoing long-wave radiation models and compared to measured values. Four meteorological datasets representing two climate regimes, a sub-humid, high-latitude environment and a semi-arid mid-latitude environment, were used to test the models...

  19. 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

  20. Application of a numerical transport correction in diffusion calculations

    International Nuclear Information System (INIS)

    Tomatis, Daniele; Dall'Osso, Aldo

    2011-01-01

    Full core calculations by ordinary transport methods can demand considerable computational time, hardly acceptable in the industrial work frame. However, the trend of next generation nuclear cores goes toward more heterogeneous systems, where transport phenomena of neutrons become very important. On the other hand, using diffusion solvers is more practical allowing faster calculations, but a specific formulation of the diffusion coefficient is requested to reproduce the scalar flux with reliable physical accuracy. In this paper, the Ronen method is used to evaluate numerically the diffusion coefficient in the slab reactor. The new diffusion solution is driven toward the solution of the integral neutron transport equation by non linear iterations. Better estimates of currents are computed and diffusion coefficients are corrected at node interfaces, still assuming Fick's law. This method enables obtaining closer results to the transport solution by a common solver in multigroup diffusion. (author)

  1. Hybrid PN-SN Calculations with SAAF for the Multiscale Transport Capability in Rattlesnake

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Yaqi; Schunert, Sebastian; DeHart, Mark; Martineau, Richard

    2016-05-01

    Two interface conditions, the Lagrange multiplier method and the upwinding method, for hybrid \\pn-\\sn calculations is proposed for the self-adjoint angular flux (SAAF) formulation of the transport equation using the continuous finite element method (FEM) for spatial discretization. These interface conditions are implemented in Rattlesnake, the radiation transport application built on MOOSE, for the on-going multiscale transport simulation effort at INL. For smoothing the solution at the interface for the Lagrange multiplier method, a method based on \\sn Lagrange interpolation on the sphere is proposed. Numerical results indicate that the interface conditions give the expected convergence.

  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. Calculation of the dose caused by internal radiation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    For the purposes of monitoring radiation exposure it is necessary to determine or to estimate the dose caused by both external and internal radiation. When comparing the value of exposure to the dose limits, account must be taken of the total dose incurred from different sources. This guide explains how to calculate the committed effective dose caused by internal radiation and gives the conversion factors required for the calculation. Application of the maximum values for radiation exposure is dealt with in ST guide 7.2, which also sets out the definitions of the quantities and concepts most commonly used in the monitoring of radiation exposure. The monitoring of exposure and recording of doses are dealt with in ST Guides 7.1 and 7.4.

  4. An Accurate Technique for Calculation of Radiation From Printed Reflectarrays

    DEFF Research Database (Denmark)

    Zhou, Min; Sorensen, Stig B.; Jorgensen, Erik

    2011-01-01

    The accuracy of various techniques for calculating the radiation from printed reflectarrays is examined, and an improved technique based on the equivalent currents approach is proposed. The equivalent currents are found from a continuous plane wave spectrum calculated by use of the spectral dyadic...... Green's function. This ensures a correct relation between the equivalent electric and magnetic currents and thus allows an accurate calculation of the radiation over the entire far-field sphere. A comparison to DTU-ESA Facility measurements of a reference offset reflectarray designed and manufactured...

  5. Measurement and calculation of radiation sources in the primary cooling system of JOYO

    International Nuclear Information System (INIS)

    Suzuki, S.; Iizawa, K.; Ohtani, N.; Kobayashi, T.; Horie, J.; Handa, H.

    1987-01-01

    Production and transfer of radiation sources in the primary cooling system are important consideration in the LMFBR plant from the viewpoint of radiation protection and shielding design. These items were evaluated with calculations and/or measurements in the Japanese experimental fast reactor JOYO. In this study, calculations were made with the DOT3.5 0 two-dimensional discrete ordinate transport code to determine the neutron flux and production rate distributions of radiation sources in the reactor vessel. Using the DOT results, the behavior in primary coolant sodium of the CP (radioactive corrosion products) which were released from the reactor structural material was also calculationally analyzed with the PSYCHE code developed by PNC. These analytical results were compared with the measured results to get the verification of analysis methods and to estimate the accuracy of calculations

  6. Shielding calculations in support of the Spallation Neutron Source (SNS) proton beam transport system

    International Nuclear Information System (INIS)

    Johnson, Jeffrey O.; Gallmeier, Franz X.; Popova, Irina

    2002-01-01

    Determining the bulk shielding requirements for accelerator environments is generally an easy task compared to analyzing the radiation transport through the complex shield configurations and penetrations typically associated with the detailed Title II design efforts of a facility. Shielding calculations for penetrations in the SNS accelerator environment are presented based on hybrid Monte Carlo and discrete ordinates particle transport methods. This methodology relies on coupling tools that map boundary surface leakage information from the Monte Carlo calculations to boundary sources for one-, two-, and three-dimensional discrete ordinates calculations. The paper will briefly introduce the coupling tools for coupling MCNPX to the one-, two-, and three-dimensional discrete ordinates codes in the DOORS code suite. The paper will briefly present typical applications of these tools in the design of complex shield configurations and penetrations in the SNS proton beam transport system

  7. 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

  8. LTRACK: Beam-transport calculation including wakefield effects

    International Nuclear Information System (INIS)

    Chan, K.C.D.; Cooper, R.K.

    1988-01-01

    LTRACK is a first-order beam-transport code that includes wakefield effects up to quadrupole modes. This paper will introduce the readers to this computer code by describing the history, the method of calculations, and a brief summary of the input/output information. Future plans for the code will also be described

  9. Load Balancing of Parallel Monte Carlo Transport Calculations

    International Nuclear Information System (INIS)

    Procassini, R J; O'Brien, M J; Taylor, J M

    2005-01-01

    The performance of parallel Monte Carlo transport calculations which use both spatial and particle parallelism is increased by dynamically assigning processors to the most worked domains. Since he particle work load varies over the course of the simulation, this algorithm determines each cycle if dynamic load balancing would speed up the calculation. If load balancing is required, a small number of particle communications are initiated in order to achieve load balance. This method has decreased the parallel run time by more than a factor of three for certain criticality calculations

  10. Dynamic Load Balancing of Parallel Monte Carlo Transport Calculations

    International Nuclear Information System (INIS)

    O'Brien, M; Taylor, J; Procassini, R

    2004-01-01

    The performance of parallel Monte Carlo transport calculations which use both spatial and particle parallelism is increased by dynamically assigning processors to the most worked domains. Since the particle work load varies over the course of the simulation, this algorithm determines each cycle if dynamic load balancing would speed up the calculation. If load balancing is required, a small number of particle communications are initiated in order to achieve load balance. This method has decreased the parallel run time by more than a factor of three for certain criticality calculations

  11. 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.

  12. Contribution to gamma ray transport calculation in heterogeneous media

    International Nuclear Information System (INIS)

    Bourdet, L.

    1985-04-01

    This thesis presents the development of gamma transport calculation codes in three dimension heterogeneous geometries. These codes allow us to define the protection against gamma-rays or verify their efficiency. The laws that govern the interactions of gamma-rays with matters are briefly revised. A library with the all necessary constants for these codes is created. TRIPOLI-2, a code that treats in exact way the neutron transport in matters using Monte-Carlo method, has been adapted to deal with the transport of gamma-rays in matters as well. TRINISHI, a code which considers only one collision, has been realized to treat heterogeneous geometries containing voids. Elaborating a formula that calculates the albedo for gamma-ray reflection (the code ALBANE) allows us to solve the problem of gamma-ray reflection on plane surfaces. NARCISSE-2 deals with gamma-rays that suffer only one reflection on the inner walls of any closed volume (rooms, halls...) [fr

  13. 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)

  14. Analysis of error in Monte Carlo transport calculations

    International Nuclear Information System (INIS)

    Booth, T.E.

    1979-01-01

    The Monte Carlo method for neutron transport calculations suffers, in part, because of the inherent statistical errors associated with the method. Without an estimate of these errors in advance of the calculation, it is difficult to decide what estimator and biasing scheme to use. Recently, integral equations have been derived that, when solved, predicted errors in Monte Carlo calculations in nonmultiplying media. The present work allows error prediction in nonanalog Monte Carlo calculations of multiplying systems, even when supercritical. Nonanalog techniques such as biased kernels, particle splitting, and Russian Roulette are incorporated. Equations derived here allow prediction of how much a specific variance reduction technique reduces the number of histories required, to be weighed against the change in time required for calculation of each history. 1 figure, 1 table

  15. Calculating the diffuse solar radiation in regions without solar radiation measurements

    International Nuclear Information System (INIS)

    Li, Huashan; Bu, Xianbiao; Long, Zhen; Zhao, Liang; Ma, Weibin

    2012-01-01

    Correlations for calculating diffuse solar radiation can be classified into models with global solar radiation (H-based method) and without it (Non-H method). The objective of the present study is to compare the performance of H-based and Non-H methods for calculating the diffuse solar radiation in regions without solar radiation measurements. The comparison is carried out at eight meteorological stations in China focusing on the monthly average daily diffuse solar radiation. Based on statistical error tests, the results show that the Non-H method that includes other readily available meteorological elements gives better estimates. Therefore, it can be concluded that the Non-H method is more appropriate than the H-based one for calculating the diffuse solar radiation in regions without solar radiation measurements. -- Highlights: ► Methods for calculating diffuse solar radiation in regions without solar radiation measurements are investigated. ► Diffuse solar radiation models can be classified into two groups according to global solar radiation. ► Two approaches are compared at the eight meteorological stations in China. ► The method without global solar radiation is recommended.

  16. Efficient calculation of dissipative quantum transport properties in semiconductor nanostructures

    Energy Technology Data Exchange (ETDEWEB)

    Greck, Peter

    2012-11-26

    We present a novel quantum transport method that follows the non-equilibrium Green's function (NEGF) framework but side steps any self-consistent calculation of lesser self-energies by replacing them by a quasi-equilibrium expression. We termed this method the multi-scattering Buettiker-Probe (MSB) method. It generalizes the so-called Buettiker-Probe model but takes into account all relevant individual scattering mechanisms. It is orders of magnitude more efficient than a fully selfconsistent non-equilibrium Green's function calculation for realistic devices, yet accurately reproduces the results of the latter method as well as experimental data. This method is fairly easy to implement and opens the path towards realistic three-dimensional quantum transport calculations. In this work, we review the fundamentals of the non-equilibrium Green's function formalism for quantum transport calculations. Then, we introduce our novel MSB method after briefly reviewing the original Buettiker-Probe model. Finally, we compare the results of the MSB method to NEGF calculations as well as to experimental data. In particular, we calculate quantum transport properties of quantum cascade lasers in the terahertz (THz) and the mid-infrared (MIR) spectral domain. With a device optimization algorithm based upon the MSB method, we propose a novel THz quantum cascade laser design. It uses a two-well period with alternating barrier heights and complete carrier thermalization for the majority of the carriers within each period. We predict THz laser operation for temperatures up to 250 K implying a new temperature record.

  17. Transport survey calculations using the spectral collocation method

    International Nuclear Information System (INIS)

    Painter, S.L.; Lyon, J.F.

    1989-01-01

    A novel transport survey code has been developed and is being used to study the sensitivity of stellarator reactor performance to various transport assumptions. Instead of following one of the usual approaches, the steady-state transport equation are solved in integral form using the spectral collocation method. This approach effectively combine the computational efficiency of global models with the general nature of 1-D solutions. A compact torsatron reactor test case was used to study the convergence properties and flexibility of the new method. The heat transport model combined Shaing's model for ripple-induced neoclassical transport, the Chang-Hinton model for axisymmetric neoclassical transport, and neoalcator scaling for anomalous electron heat flux. Alpha particle heating, radiation losses, classical electron-ion heat flow, and external heating were included. For the test problem, the method exhibited some remarkable convergence properties. As the number of basis functions was increased, the maximum, pointwise error in the integrated power balance decayed exponentially until the numerical noise level as reached. Better than 10% accuracy in the globally-averaged quantities was achieved with only 5 basis functions; better than 1% accuracy was achieved with 10 basis functions. The numerical method was also found to be very general. Extreme temperature gradients at the plasma edge which sometimes arise from the neoclassical models and are difficult to resolve with finite-difference methods were easily resolved. 8 refs., 6 figs

  18. Numerical calculation of radiation pattern of plasma channel antenna

    International Nuclear Information System (INIS)

    Xia Xinren; Yin Chengyou

    2010-01-01

    The idea of plasma channel antenna (PCA) for high power microwave weapon is presented in this paper. The radiation pattern of PCA is calculated. The directivity functions of general antenna are derived. The near electromagnetic model of PCA is created based on physical circumstances. The electromagnetic fields of PCA and surrounding air in cylindrical coordinate are given. The dispersion equation of PCA is deduced by applying the boundary conditions of electromagnetic fields. The surface wave vector of PCA is achieved. The variations of radiation characteristic with plasma density, antenna length and antenna radius are emphatically discussed. The controllability of PCA's radiation patterns is confirmed. (authors)

  19. Calculations of radiation levels during reactor operations for safeguard inspections

    International Nuclear Information System (INIS)

    Sobhy, M.

    1996-01-01

    When an internal core spent fuel storage is used in the shield tank to accommodate a large number of spent fuel baskets, physical calculations are performed to determine the number of these spent fuel elements which can be accommodated and still maintain the gamma activity outside under the permissible limit. The corresponding reactor power level is determined. The radioactivity calculations are performed for this internal storage at different axial levels to avoid the criticality of the reactor core. Transport theory is used in calculations based on collision probability for multi group cell calculations. Diffusion theory is used in three dimensions in the core calculations. The nuclear fuel history is traced and radioactive decay is calculated, since reactor fission products are very sensitive to power level. The radioactivity is calculated with a developed formula based on fuel basket loading integrity. (author)

  20. Application of maximum values for radiation exposure and principles for the calculation of radiation doses

    International Nuclear Information System (INIS)

    2007-08-01

    The guide presents the definitions of equivalent dose and effective dose, the principles for calculating these doses, and instructions for applying their maximum values. The limits (Annual Limit on Intake and Derived Air Concentration) derived from dose limits are also presented for the purpose of monitoring exposure to internal radiation. The calculation of radiation doses caused to a patient from medical research and treatment involving exposure to ionizing radiation is beyond the scope of this ST Guide

  1. Three dimensions transport calculations for PWR core; Calcul de coeur R.E.P. en transport 3D

    Energy Technology Data Exchange (ETDEWEB)

    Richebois, E

    2000-07-01

    The objective of this work is to define improved 3-D core calculation methods based on the transport theory. These methods can be particularly useful and lead to more precise computations in areas of the core where anisotropy and steep flux gradients occur, especially near interface and boundary conditions and in regions of high heterogeneity (bundle with absorbent rods). In order to apply the transport theory a new method for calculating reflector constants has been developed, since traditional methods were only suited for 2-group diffusion core calculations and could not be extrapolated to transport calculations. In this thesis work, the new method for obtaining reflector constants is derived regardless of the number of energy groups and of the operator used. The core calculations results using the reflector constants thereof obtained have been validated on the EDF's power reactor Saint Laurent B1 with MOX loading. The advantages of a 3-D core transport calculation scheme have been highlighted as opposed to diffusion methods; there are a considerable number of significant effects and potential advantages to be gained in rod worth calculations for instance. These preliminary results obtained with on particular cycle will have to be confirmed by more systematic analysis. Accidents like MSLB (main steam line break) and LOCA (loss of coolant accident) should also be investigated and constitute challenging situations where anisotropy is high and/or flux gradients are steep. This method is now being validated for others EDF's PWRs' reactors, as well as for experimental reactors and other types of commercial reactors. (author)

  2. Three dimensions transport calculations for PWR core; Calcul de coeur R.E.P. en transport 3D

    Energy Technology Data Exchange (ETDEWEB)

    Richebois, E

    2000-07-01

    The objective of this work is to define improved 3-D core calculation methods based on the transport theory. These methods can be particularly useful and lead to more precise computations in areas of the core where anisotropy and steep flux gradients occur, especially near interface and boundary conditions and in regions of high heterogeneity (bundle with absorbent rods). In order to apply the transport theory a new method for calculating reflector constants has been developed, since traditional methods were only suited for 2-group diffusion core calculations and could not be extrapolated to transport calculations. In this thesis work, the new method for obtaining reflector constants is derived regardless of the number of energy groups and of the operator used. The core calculations results using the reflector constants thereof obtained have been validated on the EDF's power reactor Saint Laurent B1 with MOX loading. The advantages of a 3-D core transport calculation scheme have been highlighted as opposed to diffusion methods; there are a considerable number of significant effects and potential advantages to be gained in rod worth calculations for instance. These preliminary results obtained with on particular cycle will have to be confirmed by more systematic analysis. Accidents like MSLB (main steam line break) and LOCA (loss of coolant accident) should also be investigated and constitute challenging situations where anisotropy is high and/or flux gradients are steep. This method is now being validated for others EDF's PWRs' reactors, as well as for experimental reactors and other types of commercial reactors. (author)

  3. 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.

  4. Lagrangian Transport Calculations Using UARS Data. Part 2; Ozone

    Science.gov (United States)

    Manney, Gloria L.; Zurek, R. W.; Froidevaux, L.; Waters, J. W.; ONeill, A.; Swinbank, R.

    1995-01-01

    Trajectory calculations are used to examine ozone transport in the polar winter stratosphere during periods of the Upper Atmosphere Research Satellite (UARS) observations. The value of these calculations for determining mass transport was demonstrated previously using UARS observations of long-lived tracers, In the middle stratosphere, the overall ozone behavior observed by the Microwave Limb Sounder in the polar vortex is reproduced by this purely dynamical model. Calculations show the evolution of ozone in the lower stratosphere during early winter to be dominated by dynamics in December 1992 in the Arctic. Calculations for June 1992 in the Antarctic show evidence of chemical ozone destruction and indicate that approx. 50% of the chemical destruction may be masked by dynamical effects, mainly diabatic descent, which bring higher ozone into the lower-stratospheric vortex. Estimating differences between calculated and observed fields suggests that dynamical changes masked approx. 20% - 35% of chemical ozone loss during late February and early March 1993 in the Arctic. In the Antarctic late winter, in late August and early September 1992, below approx. 520 K, the evolution of vortex-averaged ozone is entirely dominated by chemical effects; above this level, however, chemical ozone depletion can be partially or completely masked by dynamical effects. Our calculations for 1992 showed that chemical loss was nearly completely compensated by increases due to diabatic descent at 655 K.

  5. CALCULATING BEDLOAD TRANSPORT IN RIVERS: CONCEPTS, CALCULUS ROUTINES AND APPLICATION

    Directory of Open Access Journals (Sweden)

    Hudson de Azevedo Macedo

    2017-10-01

    Full Text Available Rivers are immensely important to human activities such as water supply, navigation, energy generation, and agriculture. They are also an important morphodynamic agent of erosion, transport and deposition. Their capacity to transport sediment depends on their hydraulic characteristics and can be predicted by mathematical models. Several mathematical models can be used to compute bed-load transport. Each one is appropriately better for certain conditions. In this paper, we present an application built in Microsoft Excel to compute the bed-load transport in rivers based on the Van Rijn mathematical model. The Van Rijn model is appropriate for rivers transporting sandy sediments in conditions of subcritical flow. Hydraulic parameters such as channel slope, stream power, and Reynolds and Froude numbers can be calculated using the application proposed here. The application was tested in the Paraná River and results from the calculations are consistent with data obtained from fieldwork surveys. The error of the application was only 20%, which shows good agreement of the model with survey values.

  6. Calculation of conversion coefficients for radiological protection against external radiation exposures

    International Nuclear Information System (INIS)

    Zankl, M.

    2001-01-01

    Calculations are essential for radiation protection practice because organ doses and effective doses cannot be measured directly. Conversion coefficients describe the numerical relationships of protection quantities and operational quantities. The latter can be measured in practical situations using suitable dosimeters. The conversion coefficients are calculated using radiation transport codes - usually based on Monte Carlo methods - that simulate the interactions of radiation with matter in computational models of the human body. A new generation of human body models, the so-called voxel models, are constructed from image data of real persons using suitable image processing systems, consequently, they represent the human anatomy more realistically than the so-called mathematical models. The numerical effects of realistic body anatomy on the calculated conversion coefficients can amount to 70% and more for external exposures. (orig.) [de

  7. TRING: a computer program for calculating radionuclide transport in groundwater

    International Nuclear Information System (INIS)

    Maul, P.R.

    1984-12-01

    The computer program TRING is described which enables the transport of radionuclides in groundwater to be calculated for use in long term radiological assessments using methods described previously. Examples of the areas of application of the program are activity transport in groundwater associated with accidental spillage or leakage of activity, the shutdown of reactors subject to delayed decommissioning, shallow land burial of intermediate level waste and geologic disposal of high level waste. Some examples of the use of the program are given, together with full details to enable users to run the program. (author)

  8. HAMMER, 1-D Multigroup Neutron Transport Infinite System Cell Calculation for Few-Group Diffusion Calculation

    International Nuclear Information System (INIS)

    Honeck, H.C.

    1984-01-01

    1 - Description of problem or function: HAMMER performs infinite lattice, one-dimensional cell multigroup calculations, followed (optionally) by one-dimensional, few-group, multi-region reactor calculations with neutron balance edits. 2 - Method of solution: Infinite lattice parameters are calculated by means of multigroup transport theory, composite reactor parameters by few-group diffusion theory. 3 - Restrictions on the complexity of the problem: - Cell calculations - maxima of: 30 thermal groups; 54 epithermal groups; 20 space points; 20 regions; 18 isotopes; 10 mixtures; 3 thermal up-scattering mixtures; 200 resonances per group; no overlap or interference; single level only. - Reactor calculations - maxima of : 40 regions; 40 mixtures; 250 space points; 4 groups

  9. Design of a transport calculation system for logging sondes simulation

    International Nuclear Information System (INIS)

    Marquez Damian, Jose Ignacio

    2005-01-01

    Analysis of available resources in earth crust is performed by different techniques, one of them is neutron logging. Design of sondes that are used to make such logging is supported by laboratory experiments as well as by numerical calculations.This work presents several calculation schemes, designed to simplify the task of whom has to planify such experiments or optimize parameters of this kind of sondes.These schemes use transport calculation codes, especially DaRT, TORT and MCNP, and cross section processing modules from SCALE system.Additionally a system for DaRT and TORT data postprocessing using OpenDX is presented.It allows scalar flux spatial distribution analysis, as wells as cross section condensation and reaction rates calculation

  10. Radiation damage calculations for the APT materials test program

    International Nuclear Information System (INIS)

    Corzine, R.K.; Wechsler, M.S.; Dudziak, D.J.; Ferguson, P.D.; James, M.R.

    1999-01-01

    A materials irradiation was performed at the Los Alamos Neutron Science Center (LANSCE) in the fall of 1996 and spring of 1997 in support of the Accelerator Production of Tritium (APT) program. Testing of the irradiated materials is underway. In the proposed APT design, materials in the target and blanket are to be exposed to protons and neutrons over a wide range of energies. The irradiation and testing program was undertaken to enlarge the very limited direct knowledge presently available of the effects of medium-energy protons (∼1 GeV) on the properties of engineering materials. APT candidate materials were placed in or near the LANSCE accelerator 800-MeV, 1-mA proton beam and received roughly the same proton current density in the center of the beam as would be the case for the APT facility. As a result, the proton fluences achieved in the irradiation were expected to approach the APT prototypic full-power-year values. To predict accurately the performance of materials in APT, radiation damage parameters for the materials experiment must be determined. By modeling the experiment, calculations for atomic displacement, helium and hydrogen cross sections and for proton and neutron fluences were done for representative samples in the 17A, 18A, and 18C areas. The LAHET code system (LCS) was used to model the irradiation program, LAHET 2.82 within LCS transports protons > 1 MeV, and neutrons >20 MeV. A modified version of MCNP for use in LCS, HMCNP 4A, was employed to tally neutrons of energies <20 MeV

  11. 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

  12. CALCULATION OF POLLUTION DYNAMICS NEAR RAILWAY TERRITORY DURING COAL TRANSPORTATION

    Directory of Open Access Journals (Sweden)

    M. M. Biliaiev

    2017-02-01

    Full Text Available Purpose. The article is aimed to develop 3D numerical model for the prediction of atmospheric pollution during transportation of bulk cargo in the railway car. Methodology.To solve this problem, it was developed three-dimensional numerical model, based on the use of the transport equation of dust pollution in the air by the wind and atmospheric turbulent diffusion. For the numerical integration of the simulating equation of the dust transport the implicit difference scheme was used. When constructing a difference scheme, it was carried out prior splitting of the original transport equation into the sequence of solutions of three equations. The first of them takes into account the transport of dust in paths, the second equation – dust transport under the influence of atmospheric turbulent diffusion, and the third equation –change of the dust concentration in the air due to its emissions from the cars.Unknown value of the pollutant concentration at every step of splitting is determined by the explicit scheme – the method of running account, which provides a simple numerical implementation of splitting equations. The developed numerical model is the basis for specialized computer program. On the basis of the constructed numerical model we carried out a computational experiment to assess the level of air pollution at the railway station during the motion of train with coal. Findings. Authors developed 3D numerical model, which belongs to the class of «screening models». This model takes into account the main physical factors affecting the process of dispersion of dust pollution in the atmosphere during coal transportation. The proposed numerical model requires low cost of computer time in the practical implementation on small and medium-power computers. This model can be used for rapid calculations of the dynamics of air pollution when transporting coal by rail. Calculations to determine the pollutant concentration and formation of the

  13. 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

  14. Axial SPN and radial MOC coupled whole core transport calculation

    International Nuclear Information System (INIS)

    Cho, Jin-Young; Kim, Kang-Seog; Lee, Chung-Chan; Zee, Sung-Quun; Joo, Han-Gyu

    2007-01-01

    The Simplified P N (SP N ) method is applied to the axial solution of the two-dimensional (2-D) method of characteristics (MOC) solution based whole core transport calculation. A sub-plane scheme and the nodal expansion method (NEM) are employed for the solution of the one-dimensional (1-D) SP N equations involving a radial transverse leakage. The SP N solver replaces the axial diffusion solver of the DeCART direct whole core transport code to provide more accurate, transport theory based axial solutions. In the sub-plane scheme, the radial equivalent homogenization parameters generated by the local MOC for a thick plane are assigned to the multiple finer planes in the subsequent global three-dimensional (3-D) coarse mesh finite difference (CMFD) calculation in which the NEM is employed for the axial solution. The sub-plane scheme induces a much less nodal error while having little impact on the axial leakage representation of the radial MOC calculation. The performance of the sub-plane scheme and SP N nodal transport solver is examined by solving a set of demonstrative problems and the C5G7MOX 3-D extension benchmark problems. It is shown in the demonstrative problems that the nodal error reaching upto 1,400 pcm in a rodded case is reduced to 10 pcm by introducing 10 sub-planes per MOC plane and the transport error is reduced from about 150 pcm to 10 pcm by using SP 3 . Also it is observed, in the C5G7MOX rodded configuration B problem, that the eigenvalues and pin power errors of 180 pcm and 2.2% of the 10 sub-planes diffusion case are reduced to 40 pcm and 1.4%, respectively, for SP 3 with only about a 15% increase in the computing time. It is shown that the SP 5 case gives very similar results to the SP 3 case. (author)

  15. Comparison of neutron transport calculations with NRC test results

    International Nuclear Information System (INIS)

    Koban, J.; Hofmann, W.

    1981-02-01

    For an exactly defined reactor arrangement (PCA = Pool Critical Assembly) neutron fluxes, neutron spectra and reaction rates for several neutron detectors were calculated by means of one and two dimensional transport codes. An international comparison proved the methods applied at KWU to be adequate. There were difficulties, however, in considering the three dimensions of the assembly which result mainly from its small dimension. This fact applies to all participants who didn't use three dimensional codes. (orig.) [de

  16. Department of Environmental and Radiation Transport Physics - Overview

    International Nuclear Information System (INIS)

    Woznicka, U.

    2001-01-01

    thermal neutron transport in heterogeneous media containing highly absorbing grains is presented in two announcements in ''Reports on research''. Our 14 MeV pulsed neutron generator, after a few years' revamp, became a valuable tool for experimental verification of theoretical solutions for the non-stationary thermal neutron diffusion in finite media. We also work on the applications of the radiation transport physics in other disciplines, namely: 1) Calculations of the neutron flux in the borehole cylindrical geometry made it possible to calibrate the geophysical well logging neutron tools. A need for measuring the thermal neutron absorption cross section of borehole brines has been proved with the experiments performed at the pulsed neutron generator. Correlations of the thermal neutron absorption cross section of rocks with their geological parameters (the K, U, Th concentration and the clay mineral content) have been theoretically suggested and experimentally confirmed. These relationships can be used as complementary methods in geophysical, geological and geochemical investigations; 2) The fields of neutrons, protons, and pions produced by electrons of energy up to 250 GeV hitting thick targets have been calculated for needs of the DESY facility, using the FLUKA Monte Carlo code. The same computer code has been used to calculate the radiation field and energy deposition in the water beam dump for the TESLA electron-positron collider; 3) Research on humidity measurements of coke, using the fast neutron and gamma transmission methods, has been finalised. The results were presented on the final IAEA meeting of the Research Programme on Bulk Hydrogen Analysis Using Neutrons. The latest research opens a possibility to build a prototype of an industrial device. (author)

  17. Uniform Gauss-Weight Quadratures for Discrete Ordinate Transport Calculations

    International Nuclear Information System (INIS)

    Carew, John F.; Hu, Kai; Zamonsky, Gabriel

    2000-01-01

    Recently, a uniform equal-weight quadrature set, UE n , and a uniform Gauss-weight quadrature set, UG n , have been derived. These quadratures have the advantage over the standard level-symmetric LQ n quadrature sets in that the weights are positive for all orders,and the transport solution may be systematically converged by increasing the order of the quadrature set. As the order of the quadrature is increased,the points approach a uniform continuous distribution on the unit sphere,and the quadrature is invariant with respect to spatial rotations. The numerical integrals converge for continuous functions as the order of the quadrature is increased.The numerical characteristics of the UE n quadrature set have been investigated previously. In this paper, numerical calculations are performed to evaluate the application of the UG n quadrature set in typical transport analyses. A series of DORT transport calculations of the >1-MeV neutron flux have been performed for a set of pressure-vessel fluence benchmark problems. These calculations employed the UG n (n = 8, 12, 16, 24, and 32) quadratures and indicate that the UG n solutions have converged to within ∼0.25%. The converged UG n solutions are found to be comparable to the UE n results and are more accurate than the level-symmetric S 16 predictions

  18. 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)

  19. Accounting for chemical kinetics in field scale transport calculations

    International Nuclear Information System (INIS)

    Bryan, N.D.

    2005-01-01

    The modelling of column experiments has shown that the humic acid mediated transport of metal ions is dominated by the non-exchangeable fraction. Metal ions enter this fraction via the exchangeable fraction, and may transfer back again. However, in both directions these chemical reactions are slow. Whether or not a kinetic description of these processes is required during transport calculations, or an assumption of local equilibrium will suffice, will depend upon the ratio of the reaction half-time to the residence time of species within the groundwater column. If the flow rate is sufficiently slow or the reaction sufficiently fast then the assumption of local equilibrium is acceptable. Alternatively, if the reaction is sufficiently slow (or the flow rate fast), then the reaction may be 'decoupled', i.e. removed from the calculation. These distinctions are important, because calculations involving chemical kinetics are computationally very expensive, and should be avoided wherever possible. In addition, column experiments have shown that the sorption of humic substances and metal-humate complexes may be significant, and that these reactions may also be slow. In this work, a set of rules is presented that dictate when the local equilibrium and decoupled assumptions may be used. In addition, it is shown that in all cases to a first approximation, the behaviour of a kinetically controlled species, and in particular its final distribution against distance at the end of a calculation, depends only upon the ratio of the reaction first order rate to the residence time, and hence, even in the region where the simplifications may not be used, the behaviour is predictable. In this way, it is possible to obtain an estimate of the migration of these species, without the need for a complex transport calculation. (orig.)

  20. Prospects in deterministic three dimensional whole-core transport calculations

    International Nuclear Information System (INIS)

    Sanchez, Richard

    2012-01-01

    The point we made in this paper is that, although detailed and precise three-dimensional (3D) whole-core transport calculations may be obtained in the future with massively parallel computers, they would have an application to only some of the problems of the nuclear industry, more precisely those regarding multiphysics or for methodology validation or nuclear safety calculations. On the other hand, typical design reactor cycle calculations comprising many one-point core calculations can have very strict constraints in computing time and will not directly benefit from the advances in computations in large scale computers. Consequently, in this paper we review some of the deterministic 3D transport methods which in the very near future may have potential for industrial applications and, even with low-order approximations such as a low resolution in energy, might represent an advantage as compared with present industrial methodology, for which one of the main approximations is due to power reconstruction. These methods comprise the response-matrix method and methods based on the two-dimensional (2D) method of characteristics, such as the fusion method.

  1. Standard Guide for Selection and Use of Mathematical Methods for Calculating Absorbed Dose in Radiation Processing Applications

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This guide describes different mathematical methods that may be used to calculate absorbed dose and criteria for their selection. Absorbed-dose calculations can determine the effectiveness of the radiation process, estimate the absorbed-dose distribution in product, or supplement or complement, or both, the measurement of absorbed dose. 1.2 Radiation processing is an evolving field and annotated examples are provided in Annex A6 to illustrate the applications where mathematical methods have been successfully applied. While not limited by the applications cited in these examples, applications specific to neutron transport, radiation therapy and shielding design are not addressed in this document. 1.3 This guide covers the calculation of radiation transport of electrons and photons with energies up to 25 MeV. 1.4 The mathematical methods described include Monte Carlo, point kernel, discrete ordinate, semi-empirical and empirical methods. 1.5 General purpose software packages are available for the calcul...

  2. 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

  3. Radiation field calculation in the vicinity of Russian radioisotope generator sources

    Energy Technology Data Exchange (ETDEWEB)

    Pretzsch, Gunter; Hummelsheim, Klemens; Bogorinski, Peter [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Kurfuerstendamm 200, 10719 Berlin (Germany)

    2005-07-01

    Germany supports the Russian Federation in the framework of the G8 Global Partnership programme to secure nuclear and radioactive materials against misuse and proliferation. In this context, GRS, on behalf of the German Foreign Office, is coordinating activities to remove disused radioisotope thermoelectric generators (RITEG) from the Baltic Sea which serve as power supply for marine lighthouses and their replacement by alternative energy sources. Further the planned project includes transportation to an interim storage, the storage equipped with radiation monitoring and physical protection measures, later transportation for reprocessing to the Mayak Production Association, where the RITEG will be dismantled in a hot cell and encapsulated radioactive source will be vitrified and stored as radioactive waste. For the whole project safety analyses are to be performed e.g. to meet radiation protection requirements. In the present paper modelling and calculation of radiation fields in the vicinity of RITEG as a basis for safety analyses is reported. (authors)

  4. Whole core transport calculation for the VHTR hexagonal core

    International Nuclear Information System (INIS)

    Cho, J. Y.; Kim, K. S.; Lee, C. C.; Joo, H. G.

    2007-01-01

    Recently, the DeCART code which performs the whole core calculation by coupling the radial MOC transport kernel with the axial nodal kernel has equipped a kernel to deal with the hexagonal geometry and applied to the VHTR hexagonal core to examine the accuracy and the computational efficiency of the implemented kernel. The implementation includes a modular ray tracing module based on the hexagonal assembly and a multi-group CMFD module to perform an efficient transport calculation. The requirements for the modular ray are: (1) the assembly based path linking and (2) the complete reflection capabilities. The first requirement is met by adjusting the azimuthal angle and the ray spacing for the modular ray to construct a core ray by the path linking. The second requirement is met by expanding the constructed azimuthal angle in the range of [0,30 degree] to the remained range to reflect completely at the core boundaries. The considered reflecting surface angles for the complete reflection are 30n's (n=1,2,1,12). The CMFD module performs the equivalent diffusion calculation to the radial MOC transport calculation based on the homogenized structure units. The structure units include the hexagonal pin cells and gap cells appearing at the assembly boundary. Therefore, the CMFD module is programmed to deal with the unstructured cells such as the gap cells. The CMFD equation consists of the two parts of (1) the conventional FDM and (2) the current corrective parts. Since the second part of the CMFD equation guarantees the reproducibility of the radial MOC transport solutions for the cell averaged reaction rate and the net current at the cell surfaces, how to build the first part of the CMFD equation is not important. Therefore, the first part of the CMFD equation is roughly built by using the normal distance from the gravity center to the surface. The VHTR core uses helium as a coolant which is realized as a void hole in a neutronics calculation. This void hole which

  5. Electron transport calculations with biomedical and environmental applications: [Progress report, FY 1987

    International Nuclear Information System (INIS)

    1987-01-01

    This project investigated radiation interactions with matter and radiation transport in bulk media, to generate basic radiological physics information. Applications include biomedical radiation dosimetry, the assessment of radiation hazards in nuclear technology, and modeling of biological radiation action. This work included the development of transport-theoretic methods, the compilation and critical evaluation of the underlying single-scattering cross sections, and the application of the transport methods to radiological physics problems. 7 refs

  6. Calculation of heat generation due to nuclear radiation in nuclear reactors

    International Nuclear Information System (INIS)

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

    1986-01-01

    The study is performed for caculating nuclear heating due to the interaction of neutrons and gamma-rays with matter. Modifications were implemented in the ANISN code, that solves the one-dimensional transport equation using the discrete ordinate method, to include nuclear heating calculations. Tests of the implemented modifications were performed in problems of nuclear heating due to radiation energy deposition in a fusion reactor. (Author) [pt

  7. Practice of calculation of neutron-physical characteristics of reactors and radiating shielding in structure SNPS with program complex MCNP

    International Nuclear Information System (INIS)

    Krotov, A.D.; Son'ko, A.V.

    2009-01-01

    Calculation of neutron-physical properties and radiation protection of space power reactor was made by means of the MCNP code allowing simulation of neutron, γ- and electron transport by the Monte Carlo method in the systems with combined geometry. Universality of the MCNP code has been demonstrated both for the calculation of reactor-converter so for the optimization of radiation protection that allows to reserve a new level of complex simulation of SNPS [ru

  8. Application of γ field theory based calculation method to the monitoring of mine nuclear radiation environment

    International Nuclear Information System (INIS)

    Du Yanjun; Liu Qingcheng; Liu Hongzhang; Qin Guoxiu

    2009-01-01

    In order to find the feasibility of calculating mine radiation dose based on γ field theory, this paper calculates the γ radiation dose of a mine by means of γ field theory based calculation method. The results show that the calculated radiation dose is of small error and can be used to monitor mine environment of nuclear radiation. (authors)

  9. Beam transport calculations for BARC-TIFR 14UD pelletron

    International Nuclear Information System (INIS)

    Prasad, K.G.

    1993-01-01

    The 14UD pelletron tandem accelerator installed at Tata Institute of Fundamental Research (TIFR) as a joint BARC-TIFR project, is supplied by National Electrostatic Corporation (NEC), U.S.A. To optimise the parameters of various elements along the beam path, it is essential to work out the beam optics of the entire system. There are various computer codes in use for such calculations. All these codes, except the detailed ray tracing programs, use matrix formulation. Thus each ion optical element is characterised in terms of a transport matrix, whose elements are assumed to be independent of particle trajectory. We have performed only the first order calculations, meaning thereby that no aberrations are included. Further, all calculations are carried out assuming ideal conditions like axial beam injection, perfectly aligned beam line elements, etc. The main code that has been employed in our calculations is based on the one at the Australian National University, Canberra, suitably modified for use with CYBER 170/730 computer at TIFR. However, codes at NEC and Stony Brook were also used for the checking the results. The results of calculations are given and discussed. (author). 2 figs

  10. Modeling Dynamic Objects in Monte Carlo Particle Transport Calculations

    International Nuclear Information System (INIS)

    Yegin, G.

    2008-01-01

    In this study, the Multi-Geometry geometry modeling technique was improved in order to handle moving objects in a Monte Carlo particle transport calculation. In the Multi-Geometry technique, the geometry is a superposition of objects not surfaces. By using this feature, we developed a new algorithm which allows a user to make enable or disable geometry elements during particle transport. A disabled object can be ignored at a certain stage of a calculation and switching among identical copies of the same object located adjacent poins during a particle simulation corresponds to the movement of that object in space. We called this powerfull feature as Dynamic Multi-Geometry technique (DMG) which is used for the first time in Brachy Dose Monte Carlo code to simulate HDR brachytherapy treatment systems. Our results showed that having disabled objects in a geometry does not effect calculated dose values. This technique is also suitable to be used in other areas such as IMRT treatment planning systems

  11. Radiation shielding calculation for the MOX fuel fabrication plant Melox

    International Nuclear Information System (INIS)

    Lee, Y.K.; Nimal, J.C.; Chiron, M.

    1994-01-01

    Radiation shielding calculation is an important engineering work in the design of the MOX fuel fabrication plant MELOX. Due to the recycle of plutonium and uranium from UO2 spent fuel reprocessing and the large capacity of production (120t HM/yr.), the shielding design requires more attention in this LWR fuel plant. In MELOX, besides several temporary storage facilities of massive fissile material, about one thousand radioactive sources with different geometries, forms, densities, quantities and Pu concentrations, are distributed through different workshops from the PuO 2 powder reception unit to the fuel assembly packing room. These sources, with or without close shield, stay temporarily in different locations, containers and glove boxes. In order to optimize the dimensions, the material and the cost of shield as well as to limit the calculation work in a reasonable engineer-hours, a calculation scheme for shielding design of MELOX is developed. This calculation scheme has been proved to be useful in consideration of the feedback from the evolutionary design and construction. The validated shielding calculations give a predictive but reliable radiation doses information. (authors). 2 figs., 10 refs

  12. Ab Initio Calculations of Transport Properties of Vanadium Oxides

    Science.gov (United States)

    Lamsal, Chiranjivi; Ravindra, N. M.

    2018-04-01

    The temperature-dependent transport properties of vanadium oxides have been studied near the Fermi energy using the Kohn-Sham band structure approach combined with Boltzmann transport equations. V2O5 exhibits significant thermoelectric properties, which can be attributed to its layered structure and stability. Highly anisotropic electrical conduction in V2O5 is clearly manifested in the calculations. Due to specific details of the band structure and anisotropic electron-phonon interactions, maxima and crossovers are also seen in the temperature-dependent Seebeck coefficient of V2O5. During the phase transition of VO2, the Seebeck coefficient changes by 18.9 µV/K, which is close to (within 10% of) the observed discontinuity of 17.3 µV/K.

  13. The calculation of maximum permissible exposure levels for laser radiation

    International Nuclear Information System (INIS)

    Tozer, B.A.

    1979-01-01

    The maximum permissible exposure data of the revised standard BS 4803 are presented as a set of decision charts which ensure that the user automatically takes into account such details as pulse length and pulse pattern, limiting angular subtense, combinations of multiple wavelength and/or multiple pulse lengths, etc. The two decision charts given are for the calculation of radiation hazards to skin and eye respectively. (author)

  14. Detriment calculations resulting from occupational radiation exposures in Egypt

    International Nuclear Information System (INIS)

    Abdel-Ghani, A.H.

    2000-01-01

    The application of the nominal probability coefficient to evaluate the detriment after the annual occupational exposures of workers from radiation sources and radioactive material have been calculated for workers in medical practices, industrial applications, atomic energy activities and those involved in exploration and mining of radioactive ores and phosphates. The aim of detriment calculations is to provide a foresight for the future occurrence of stochastic effects among the exposed workers. The calculated detriment can be classified into three classes. The first includes workers in diagnostic radiology and atomic energy activities who received the higher doses and consequently represent the higher detriment. The second class comprises workers in radiotherapy and nuclear medicine whose detriment is for times lesser than that of the first class. The third one concerns workers in industrial applications and in exploration and mining of radioactive ores and phosphates, their detriments ten times lesser than that of the second class. The occupational radiation doses are endorsed by the united nation scientific committee on efects of atomic radiation (UNSCEAR) for the period january 1995 to december 1998

  15. Transport and hydrodynamic calculations of direct photons at FAIR

    International Nuclear Information System (INIS)

    Baeuchle, Bjorn; Bleicher, Marcus

    2011-01-01

    The microscopic transport model UrQMD and a micro + macro hybrid model are used to calculate direct photon spectra from U+U-collisions at E lab =35 A GeV as will be measured by the CBM Collaboration at FAIR. In the hybrid model, the intermediate high-density part of the nuclear interaction is described with ideal 3+1-dimensional hydrodynamics. Different equations of state of the matter created in the heavy-ion collisions are investigated and the resulting spectra of direct photons are predicted. The emission patterns of direct photons in space and time are discussed.

  16. 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

  17. Two-dimensional radiation shielding optimization analysis of spent fuel transport container

    International Nuclear Information System (INIS)

    Tian Yingnan; Chen Yixue; Yang Shouhai

    2013-01-01

    The intelligent radiation shielding optimization design software platform is a one-dimensional multi-target radiation shielding optimization program which is developed on the basis of the genetic algorithm program and one-dimensional discrete ordinate program-ANISN. This program was applied in the optimization design analysis of the spent fuel transport container radiation shielding. The multi-objective optimization calculation model of the spent fuel transport container radiation shielding was established, and the optimization calculation of the spent fuel transport container weight and radiation dose rate was carried by this program. The calculation results were checked by Monte-Carlo program-MCNP/4C. The results show that the weight of the optimized spent fuel transport container decreases to 81.1% of the origin and the radiation dose rate decreases to below 65.4% of the origin. The maximum deviation between the calculated values from the program and the MCNP is below 5%. The results show that the optimization design scheme is feasible and the calculation result is correct. (authors)

  18. Global aerosol transport and consequences for the radiation budget

    International Nuclear Information System (INIS)

    Newiger, M.; Grassl, H.; Schussel, P.; Rehkopf, J.

    1984-01-01

    Man's activities may influence global climate by changing the atmospheric composition and surface characteristics and by waste heat. Most prominent within this discussion is the increase or decrease of radiatively active trace gases like CO/sub 2/, N/sub 2/O, O/sub 3/, and others. The general opinion is converging towards a greenhouse effect as a combined action of all trace gases, whose exact magnitude is uncertain mainly because of the unknown reaction of water cycle. The aim of our global 2-D (resolving latitude and height) aerosol transport model is the calculation of aerosol particle number density profiles as a function of latitude for present natural plus anthropogenic emissions. The aerosol transport model uses prescribed meridonal circulation, diffusivity factors and cloud climatology for January as well as July. All these latitude and height dependent input parameters were taken from well known sources. The fixed climatology excludes the feedback of aerosol particle parameter changes on mean circulation. However, the radiative parameters of six clouds types are modified, although they possess by adoption of the Telegadas and London (1954) cloud climatology prescribed amount and height. The inclusion of the feedback on mean circulation seems premature at present. Adding particles either accounting for natural emissions or natural anthropogenic emission and removing particles by all known sinks outside and within clouds gives us - for the stationary state - vertical profiles of aerosol number density in three sizes classes as a function of latitude. These profiles in turn are input for radiation flux calculations in clear and cloudy areas in order to assess net flux changes caused by the present aerosol load in comparison to a scenario without anthropogenic emissions. The net flux changes finally are compared to those calculated for increased CO/sub 2/ levels

  19. 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

  20. 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

  1. Repair for scattering expansion truncation errors in transport calculations

    International Nuclear Information System (INIS)

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

    1980-01-01

    Legendre expansion of angular scattering distributions is usually limited to P 3 in practical transport calculations. This truncation often results in non-trivial errors, especially alternating negative and positive lateral scattering peaks. The effect is especially prominent in forward-peaked situations such as the within-group component of the Compton Scattering of gammas. Increasing the expansion to P 7 often makes the peaks larger and narrower. Ward demonstrated an accurate repair, but his method requires special cross section sets and codes. The DOT IV code provides fully-compatible, but heuristic, repair of the erroneous scattering. An analytical Klein-Nishina estimator, newly available in the MORSE code, allows a test of this method. In the MORSE calculation, particle scattering histories are calculated in the usual way, with scoring by an estimator routine at each collision site. Results for both the conventional P 3 estimator and the analytical estimator were obtained. In the DOT calculation, the source moments are expanded into the directional representation at each iteration. Optionally a sorting procedure removes all negatives, and removes enough small positive values to restore particle conservation. The effect of this is to replace the alternating positive and negative values with positive values of plausible magnitude. The accuracy of those values is examined herein

  2. Parallel processing of two-dimensional Sn transport calculations

    International Nuclear Information System (INIS)

    Uematsu, M.

    1997-01-01

    A parallel processing method for the two-dimensional S n transport code DOT3.5 has been developed to achieve a drastic reduction in computation time. In the proposed method, parallelization is achieved with angular domain decomposition and/or space domain decomposition. The calculational speed of parallel processing by angular domain decomposition is largely influenced by frequent communications between processing elements. To assess parallelization efficiency, sample problems with up to 32 x 32 spatial meshes were solved with a Sun workstation using the PVM message-passing library. As a result, parallel calculation using 16 processing elements, for example, was found to be nine times as fast as that with one processing element. As for parallel processing by geometry segmentation, the influence of processing element communications on computation time is small; however, discontinuity at the segment boundary degrades convergence speed. To accelerate the convergence, an alternate sweep of angular flux in conjunction with space domain decomposition and a two-step rescaling method consisting of segmentwise rescaling and ordinary pointwise rescaling have been developed. By applying the developed method, the number of iterations needed to obtain a converged flux solution was reduced by a factor of 2. As a result, parallel calculation using 16 processing elements was found to be 5.98 times as fast as the original DOT3.5 calculation

  3. Neutron and photon transport calculations in fusion system. 2

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Satoshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1998-03-01

    On the application of MCNP to the neutron and {gamma}-ray transport calculations for fusion reactor system, the wide range design calculation has been carried out in the engineering design activities for the international thermonuclear fusion experimental reactor (ITER) being developed jointly by Japan, USA, EU and Russia. As the objects of shielding calculation for fusion reactors, there are the assessment of dose equivalent rate for living body shielding and the assessment of the nuclear response for the soundness of in-core structures. In the case that the detailed analysis of complicated three-dimensional shapes is required, the assessment using MCNP has been carried out. Also when the nuclear response of peripheral equipment due to the gap streaming between blanket modules is evaluated with good accuracy, the calculation with MCNP has been carried out. The analyses of the shieldings for blanket modules and NBI port are explained, and the examples of the results of analyses are shown. In the blanket modules, there are penetrating holes and continuous gap. In the case of the NBI port, shielding plug cannot be installed. These facts necessitate the MCNP analysis with high accuracy. (K.I.)

  4. Effects of the scattering anisotropy approximation in multigroup radiation shielding calculations

    International Nuclear Information System (INIS)

    Altiparmarkov, D.

    1983-01-01

    Expansion of the scattering cross-sections into Legendre series is the usual way of solving the neutron transport problem. Because of the large space gradients of the neutron flux, the effects of that approximations become especially remarkable in the radiation shielding calculations. In this paper, a method taking into account scattering anisotropy is presented. From the point of view of the accuracy and computing speed, the optimal approximation of the scattering anisotropy is established for the basic protective materials on the basis of simple problem calculations (author) [sr

  5. Robust volume calculations for Constructive Solid Geometry (CSG) components in Monte Carlo transport calculations

    Energy Technology Data Exchange (ETDEWEB)

    Millman, D. L. [Dept. of Computer Science, Univ. of North Carolina at Chapel Hill (United States); Griesheimer, D. P.; Nease, B. R. [Bechtel Marine Propulsion Corporation, Bertis Atomic Power Laboratory (United States); Snoeyink, J. [Dept. of Computer Science, Univ. of North Carolina at Chapel Hill (United States)

    2012-07-01

    In this paper we consider a new generalized algorithm for the efficient calculation of component object volumes given their equivalent constructive solid geometry (CSG) definition. The new method relies on domain decomposition to recursively subdivide the original component into smaller pieces with volumes that can be computed analytically or stochastically, if needed. Unlike simpler brute-force approaches, the proposed decomposition scheme is guaranteed to be robust and accurate to within a user-defined tolerance. The new algorithm is also fully general and can handle any valid CSG component definition, without the need for additional input from the user. The new technique has been specifically optimized to calculate volumes of component definitions commonly found in models used for Monte Carlo particle transport simulations for criticality safety and reactor analysis applications. However, the algorithm can be easily extended to any application which uses CSG representations for component objects. The paper provides a complete description of the novel volume calculation algorithm, along with a discussion of the conjectured error bounds on volumes calculated within the method. In addition, numerical results comparing the new algorithm with a standard stochastic volume calculation algorithm are presented for a series of problems spanning a range of representative component sizes and complexities. (authors)

  6. Robust volume calculations for Constructive Solid Geometry (CSG) components in Monte Carlo transport calculations

    International Nuclear Information System (INIS)

    Millman, D. L.; Griesheimer, D. P.; Nease, B. R.; Snoeyink, J.

    2012-01-01

    In this paper we consider a new generalized algorithm for the efficient calculation of component object volumes given their equivalent constructive solid geometry (CSG) definition. The new method relies on domain decomposition to recursively subdivide the original component into smaller pieces with volumes that can be computed analytically or stochastically, if needed. Unlike simpler brute-force approaches, the proposed decomposition scheme is guaranteed to be robust and accurate to within a user-defined tolerance. The new algorithm is also fully general and can handle any valid CSG component definition, without the need for additional input from the user. The new technique has been specifically optimized to calculate volumes of component definitions commonly found in models used for Monte Carlo particle transport simulations for criticality safety and reactor analysis applications. However, the algorithm can be easily extended to any application which uses CSG representations for component objects. The paper provides a complete description of the novel volume calculation algorithm, along with a discussion of the conjectured error bounds on volumes calculated within the method. In addition, numerical results comparing the new algorithm with a standard stochastic volume calculation algorithm are presented for a series of problems spanning a range of representative component sizes and complexities. (authors)

  7. Radiation calculations using LAHET/MCNP/CINDER90

    International Nuclear Information System (INIS)

    Waters, L.

    1994-01-01

    The LAHET monte carlo code system has recently been expanded to include high energy hadronic interactions via the FLUKA code, while retaining the original Los Alamos versions of HETC and ISABEL at lower energies. Electrons and photons are transported with EGS4 or ITS, while the MCNP coupled neutron/photon monte carlo code provides analysis of neutrons with kinetic energies less than 20 MeV. An interface with the CINDER activation code is now in common use. Various other changes have been made to facilitate analysis of high energy accelerator radiation environments and experimental physics apparatus, such as those found at SSC and RHIC. Current code developments and applications are reviewed

  8. Simplified models for radiational losses calculating a tokamak plasma

    International Nuclear Information System (INIS)

    Arutiunov, A.B.; Krasheninnikov, S.I.; Prokhorov, D.Yu.

    1990-01-01

    To determine the magnitudes and profiles of radiational losses in a Tokamak plasma, particularly for high plasma densities, when formation of MARFE or detached-plasma takes place, it is necessary to know impurity distribution over the ionization states. Equations describing time evolution of this distribution are rather cumbersome, besides that, transport coefficients as well as rate constants of the processes involving complex ions are known nowadays with high degree of uncertainty, thus it is believed necessary to develop simplified, half-analytical models describing time evolution of the impurities analysis of physical processes taking place in a Tokamak plasma on the base of the experimental data. (author) 6 refs., 2 figs

  9. Calculation of the quantities of radiation risk in Japanese population

    International Nuclear Information System (INIS)

    Nakamura, Yuji

    1993-01-01

    The purpose of this study was to reevaluate various kinds of indicators of radiation risks using additive projection and multiplicative projection models, as proposed by ICRP. Total death probability rate (1985) and probability rate of cancer death (1983 to 1987) were used as data base. The following indicators were calculated: total conditional death probability rate and conditional death probability rate; normalized death age probability density and unconditional death probability rate; attributable life-time probability of cancer death; and other risk indicators, including mean loss of life expectancy, reduction of life expectancy, mean annually committed probability of attributable cancer deaths, annual extra probability of cancer death, probability density of the age of death, maximum relative death probability rate (age at maximum relative rate), and probabilistic aging. In terms of calculations of these risk indicators for the comprehensive cancer death, there was no great difference between the Japanese population and ICRP. When calculating according to sites of cancer, calculations of indicators for cancer mortality (or cancer cure rate) in the Japanese population might bedifferent from ICRP's calculation. (N.K.) different from ICRP's calculations. (N.K.)

  10. Calculation of the line shapes of radiators immersed in plasma

    International Nuclear Information System (INIS)

    Hayrapetian, A.S.

    1987-01-01

    This work reports the results of theoretical calculations of line shapes of radiators immersed in plasma. The fluctuating electric field of the plasma perturbs the atomic structure of the immersed ions or atoms. The perturbed line shape is an important diagnostic tool for the temperature and density measurements of plasma. The line-shape calculation here is done by first deriving the line-shape expression, then it is shown that the problem is equivalent to calculating the temperature Green's function of the bound electron. The total Hamiltonian of the system includes the atomic, plasma, and atom-plasma parts. The temperature Green's function is calculated perturbatively by expanding in orders of atom-plasma interaction. By solving a Dyson equation, the dressed Green's functions of the bound electrons are obtained. At this point, the line shape is calculated by an analytic continuation from the complex frequency plane to real line. To derive the atomic electron Green's function, the momentum integral in the self-energy is approximated by a Riemann sum. With this approximation, the algebraic form of the line shape is obtained for an undetermined number of terms in the Riemann sum. Numerical calculation of line shape is done by using this result

  11. Error reduction techniques for Monte Carlo neutron transport calculations

    International Nuclear Information System (INIS)

    Ju, J.H.W.

    1981-01-01

    Monte Carlo methods have been widely applied to problems in nuclear physics, mathematical reliability, communication theory, and other areas. The work in this thesis is developed mainly with neutron transport applications in mind. For nuclear reactor and many other applications, random walk processes have been used to estimate multi-dimensional integrals and obtain information about the solution of integral equations. When the analysis is statistically based such calculations are often costly, and the development of efficient estimation techniques plays a critical role in these applications. All of the error reduction techniques developed in this work are applied to model problems. It is found that the nearly optimal parameters selected by the analytic method for use with GWAN estimator are nearly identical to parameters selected by the multistage method. Modified path length estimation (based on the path length importance measure) leads to excellent error reduction in all model problems examined. Finally, it should be pointed out that techniques used for neutron transport problems may be transferred easily to other application areas which are based on random walk processes. The transport problems studied in this dissertation provide exceptionally severe tests of the error reduction potential of any sampling procedure. It is therefore expected that the methods of this dissertation will prove useful in many other application areas

  12. 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

  13. Calculated characteristics of subcritical assembly with anisotropic transport of neutrons

    International Nuclear Information System (INIS)

    Gorin, N.V.; Lipilina, E.N.; Lyutov, V.D.; Saukov, A.I.

    2003-01-01

    There was considered possibility of creating enough sub-critical system that multiply neutron fluence from a primary source by many orders. For assemblies with high neutron tie between parts, it is impossible. That is why there was developed a construction consisting of many units (cascades) having weak feedback with preceding cascades. The feedback attenuation was obtained placing layers of slow neutron absorber and moderators between the cascades of fission material. Anisotropy of fast neutron transport through the layers was used. The system consisted of many identical cascades aligning one by another. Each cascade consists of layers of moderator, fissile material and absorber of slow neutrons. The calculations were carried out using the code MCNP.4a with nuclear data library ENDF/B5. In this construction neutrons spread predominantly in one direction multiplying in each next fissile layer, and they attenuate considerably in the opposite direction. In a calculated construction, multiplication factor of one cascade is about 1.5 and multiplication factor of whole construction composed of n cascades is 1.5 n . Calculated keff value is 0.9 for one cascade and does not exceed 0.98 for a system containing any number of cascades. Therefore the assembly is always sub-critical and therefore it is safe in respect of criticality. There was considered using such a sub-critical assembly to create a powerful neutron fluence for neutron boron-capturing therapy. The system merits and demerits were discussed. (authors)

  14. Calculation and Measurement of Low-Energy Radiative Moller Scattering

    Science.gov (United States)

    Epstein, Charles; DarkLight Collaboration

    2017-09-01

    A number of current nuclear physics experiments have come to rely on precise knowledge of electron-electron (Moller) and positron-electron (Bhabha) scattering. Some of these experiments, having lepton beams on targets containing atomic electrons, use these purely-QED processes as normalization. In other scenarios, with electron beams at low energy and very high intensity, Moller scattering and radiative Moller scattering have such enormous cross-sections that the backgrounds they produce must be understood. In this low-energy regime, the electron mass is also not negligible in the calculation of the cross section. This is important, for example, in the DarkLight experiment (100 MeV). As a result, we have developed a new event generator for the radiative Moller and Bhabha processes, with new calculations that keep all terms of the electron mass. The MIT High Voltage Research Laboratory provides us a unique opportunity to study this process experimentally and compare it with our work, at a low beam energy of 2.5 MeV where the effects of the electron mass are significant. We are preparing a dedicated apparatus consisting of a magnetic spectrometer in order to directly measure this process. An overview of the calculation and the status of the experiment will be presented.

  15. Calculation of Transport Coefficients in Dense Plasma Mixtures

    Science.gov (United States)

    Haxhimali, T.; Cabot, W. H.; Caspersen, K. J.; Greenough, J.; Miller, P. L.; Rudd, R. E.; Schwegler, E. R.

    2011-10-01

    We use classical molecular dynamics (MD) to estimate species diffusivity and viscosity in mixed dense plasmas. The Yukawa potential is used to describe the screened Coulomb interaction between the ions. This potential has been used widely, providing the basis for models of dense stellar materials, inertial confined plasmas, and colloidal particles in electrolytes. We calculate transport coefficients in equilibrium simulations using the Green- Kubo relation over a range of thermodynamic conditions including the viscosity and the self - diffusivity for each component of the mixture. The interdiffusivity (or mutual diffusivity) can then be related to the self-diffusivities by using a generalization of the Darken equation. We have also employed non-equilibrium MD to estimate interdiffusivity during the broadening of the interface between two regions each with a high concentration of either species. Here we present results for an asymmetric mixture between Ar and H. These can easily be extended to other plasma mixtures. A main motivation for this study is to develop accurate transport models that can be incorporated into the hydrodynamic codes to study hydrodynamic instabilities. We use classical molecular dynamics (MD) to estimate species diffusivity and viscosity in mixed dense plasmas. The Yukawa potential is used to describe the screened Coulomb interaction between the ions. This potential has been used widely, providing the basis for models of dense stellar materials, inertial confined plasmas, and colloidal particles in electrolytes. We calculate transport coefficients in equilibrium simulations using the Green- Kubo relation over a range of thermodynamic conditions including the viscosity and the self - diffusivity for each component of the mixture. The interdiffusivity (or mutual diffusivity) can then be related to the self-diffusivities by using a generalization of the Darken equation. We have also employed non-equilibrium MD to estimate interdiffusivity during

  16. Utilization of MAX and FAX human phantoms for space radiation exposure calculations using HZETRN

    Science.gov (United States)

    Qualls, Garry; Slaba, Tony; Clowdsley, Martha; Blattnig, Steve; Walker, Steven; Simonsen, Lisa

    To estimate astronaut health risk due to space radiation, one must have the ability to calculate, for known radiation environments external to the body, particle spectra, LET spectra, dose, dose equivalent, or gray equivalent that are averaged over specific organs or tissue types. This may be accomplished using radiation transport software and computational human body tissue models. Historically, NASA scientists have used the HZETRN software to calculate radiation transport through both vehicle shielding materials and body tissue. The Computerized Anatomical Man (CAM) and the Computerized Anatomical Female (CAF) body models, combined with the CAMERA software, have been used for body tissue self-shielding calculations. The CAM and CAF, which were developed in 1973 and 1992, respectively, model the 50th percentile U.S. Air Force male and female and are constructed using individual quadric surfaces that combine to form thousands of solid regions that represent specific tissues and structures within the body. In order to transport an external radiation environment to a point within one of the body models using HZETRN, a directional distribution of the tissues surrounding that point is needed. The CAMERA software is used to "ray trace" the CAM and CAF models, providing the thickness of each tissue type traversed along each of a large number of rays originating at a dose point. More recently, R. Kramer of the Departmento de Energia Nuclear, Universidade Federal de Pernambuco in Brazil and his co-workers developed the Male Adult voXel (MAX) model and the Female Adult voXel (FAX). These voxel-based body models were developed using segmented Computed Tomography (CT) scans of adult cadavers, and the quantities and distributions of various body tissues have been adjusted to match those specified in the International Commission on Radiological Protection (ICRP) reference adult male and female. A new set of tools has been developed to facilitate space radiation exposure

  17. Numerical shoves and countershoves in electron transport calculations

    International Nuclear Information System (INIS)

    Filippone, W.L.

    1986-01-01

    The justification for applying the relatively complex (compared to S/sub n/) streaming ray (SR) algorithm to electron transport problems is its potential for doing rapid and accurate calculations. Because of the Lagrangian treatment of the cell-uncollided electrons, the only significant sources of error are the numerical treatment of the scattering kernel and the spatial differencing scheme used for the cell-collided electrons. Considerable progress has been made in reducing the former source of error. If one is willing to pay the price, the latter source of error can be reduced to any desired level by refining the mesh size or by using high-order differencing schemes. Here the method of numerical shoves and countershoves is introduced, which reduces spatial differencing errors using relatively little additional computational effort

  18. 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

  19. 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

  20. 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

  1. Radiation Resistance Test of Wireless Sensor Node and the Radiation Shielding Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Li, Liqan; Sur, Bhaskar [Atomic Energy of Canada Limited, Ontario (Canada); Wang, Quan [University of Western Ontario, Ontario (Canada); Deng, Changjian [The University of Electronic Science and Technology, Chengdu (China); Chen, Dongyi; Jiang, Jin [Applied Physics Branch, Ontario (Korea, Republic of)

    2014-08-15

    A wireless sensor network (WSN) is being developed for nuclear power plants. Amongst others, ionizing radiation resistance is one essential requirement for WSN to be successful. This paper documents the work done in Chalk River Laboratories of Atomic Energy of Canada Limited (AECL) to test the resistance to neutron and gamma radiation of some WSN nodes. The recorded dose limit that the nodes can withstand before being damaged by the radiation is compared with the radiation environment inside a typical CANDU (CANada Deuterium Uranium) power plant reactor building. Shielding effects of polyethylene, cadmium and lead to neutron and gamma radiations are also analyzed using MCNP simulation. The shielding calculation can be a reference for the node case design when high dose rate or accidental condition (like Fukushima) is to be considered.

  2. Calculation of radiation dose received in computed tomography examinations

    International Nuclear Information System (INIS)

    Abed Elseed, Eslam Mustafa

    2014-07-01

    Diagnostic computed tomography (CT) examinations play an important role in the health care of the population. These examination may involve significant irradiation of the patient and probably represent the largest man-made source of radiation exposure for the population. This study was performed to assess the effective dose (ED) received in brain CT examination ( base of skull and cerebrum) and to analyze effective dose distributions among radiological departments under study. The study was performed at Elnileen Medical Center, coverage one CT unit and a sample of 51 patients (25 cerebrum sample and 26 base of skull sample). The following parameters were recorded age, weight, height body mass index (BMI) derived from weight (kg) and height ( m) and exposure factor and CTDI voi , DLP value. The effective dose was measured for brain CT examination. The ED values were calculated from the obtained DLP values using AAPM report No 96 calculation methods. The results of ED values calculated showed that patient exposure were within the normal range of exposure. The mean ED values calculated were 0.35±0.15 for base of skull of brain CT examinations and 0.70±0.32 for cerebrum of brain CT examination, respectively. Further studies are recommended with more number of pa.(Author)

  3. Calculation of Selected Emissions from Transport Services in Road Public Transport

    Directory of Open Access Journals (Sweden)

    Konečný Vladimír

    2017-01-01

    Full Text Available The article deals with road public transport and its impact on the environment. According to the methodology given in EN 16258, CO2 emission value has been calculated. The input data for the calculation and the results are shown in the tables. The declaration is created according to STN CEN / TR 14310, which contains recommendations for compiling environmental reports. Finally, the comparison of the environmental impact of a bus and a passenger car, when converted to one passenger, bus has a lower CO2 emission than a passenger car in that section.

  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. Status of the solar and infrared radiation submodels in the LLNL 1-D and 2-D chemical-transport models

    International Nuclear Information System (INIS)

    Grant, K.E.; Taylor, K.E.; Ellis, J.S.; Wuebbles, D.J.

    1987-07-01

    The authors have implemented a series of state of the art radiation transport submodels in previously developed one dimensional and two dimensional chemical transport models of the troposphere and stratosphere. These submodels provide the capability of calculating accurate solar and infrared heating rates. They are a firm basis for further radiation submodel development as well as for studying interactions between radiation and model dynamics under varying conditions of clear sky, clouds, and aerosols. 37 refs., 3 figs

  6. Charge transport properties of CdMnTe radiation detectors

    Directory of Open Access Journals (Sweden)

    Prokopovich D. A.

    2012-10-01

    Full Text Available Growth, fabrication and characterization of indium-doped cadmium manganese telluride (CdMnTe radiation detectors have been described. Alpha-particle spectroscopy measurements and time resolved current transient measurements have yielded an average charge collection efficiency approaching 100 %. Spatially resolved charge collection efficiency maps have been produced for a range of detector bias voltages. Inhomogeneities in the charge transport of the CdMnTe crystals have been associated with chains of tellurium inclusions within the detector bulk. Further, it has been shown that the role of tellurium inclusions in degrading charge collection is reduced with increasing values of bias voltage. The electron drift velocity was calculated from the rise time distribution of the preamplifier output pulses at each measured bias. From the dependence of drift velocity on applied electric field the electron mobility was found to be μn = (718 ± 55 cm2/Vs at room temperature.

  7. Stationary neutrino radiation transport by maximum entropy closure

    International Nuclear Information System (INIS)

    Bludman, S.A.

    1994-11-01

    The authors obtain the angular distributions that maximize the entropy functional for Maxwell-Boltzmann (classical), Bose-Einstein, and Fermi-Dirac radiation. In the low and high occupancy limits, the maximum entropy closure is bounded by previously known variable Eddington factors that depend only on the flux. For intermediate occupancy, the maximum entropy closure depends on both the occupation density and the flux. The Fermi-Dirac maximum entropy variable Eddington factor shows a scale invariance, which leads to a simple, exact analytic closure for fermions. This two-dimensional variable Eddington factor gives results that agree well with exact (Monte Carlo) neutrino transport calculations out of a collapse residue during early phases of hydrostatic neutron star formation

  8. Calculation of the radiative properties of photosynthetic microorganisms

    International Nuclear Information System (INIS)

    Dauchet, Jérémi; Blanco, Stéphane; Cornet, Jean-François; Fournier, Richard

    2015-01-01

    A generic methodological chain for the predictive calculation of the light-scattering and absorption properties of photosynthetic microorganisms within the visible spectrum is presented here. This methodology has been developed in order to provide the radiative properties needed for the analysis of radiative transfer within photobioreactor processes, with a view to enable their optimization for large-scale sustainable production of chemicals for energy and chemistry. It gathers an electromagnetic model of light-particle interaction along with detailed and validated protocols for the determination of input parameters: morphological and structural characteristics of the studied microorganisms as well as their photosynthetic-pigment content. The microorganisms are described as homogeneous equivalent-particles whose shape and size distribution is characterized by image analysis. The imaginary part of their refractive index is obtained thanks to a new and quite extended database of the in vivo absorption spectra of photosynthetic pigments (that is made available to the reader). The real part of the refractive index is then calculated by using the singly subtractive Kramers–Krönig approximation, for which the anchor point is determined with the Bruggeman mixing rule, based on the volume fraction of the microorganism internal-structures and their refractive indices (extracted from a database). Afterwards, the radiative properties are estimated using the Schiff approximation for spheroidal or cylindrical particles, as a first step toward the description of the complexity and diversity of the shapes encountered within the microbial world. Finally, these predictive results are confronted to experimental normal-hemispherical transmittance spectra for validation. This entire procedure is implemented for Rhodospirillum rubrum, Arthrospira platensis and Chlamydomonas reinhardtii, each representative of the main three kinds of photosynthetic microorganisms, i.e. respectively

  9. Calculation of the radiative properties of photosynthetic microorganisms

    Science.gov (United States)

    Dauchet, Jérémi; Blanco, Stéphane; Cornet, Jean-François; Fournier, Richard

    2015-08-01

    A generic methodological chain for the predictive calculation of the light-scattering and absorption properties of photosynthetic microorganisms within the visible spectrum is presented here. This methodology has been developed in order to provide the radiative properties needed for the analysis of radiative transfer within photobioreactor processes, with a view to enable their optimization for large-scale sustainable production of chemicals for energy and chemistry. It gathers an electromagnetic model of light-particle interaction along with detailed and validated protocols for the determination of input parameters: morphological and structural characteristics of the studied microorganisms as well as their photosynthetic-pigment content. The microorganisms are described as homogeneous equivalent-particles whose shape and size distribution is characterized by image analysis. The imaginary part of their refractive index is obtained thanks to a new and quite extended database of the in vivo absorption spectra of photosynthetic pigments (that is made available to the reader). The real part of the refractive index is then calculated by using the singly subtractive Kramers-Krönig approximation, for which the anchor point is determined with the Bruggeman mixing rule, based on the volume fraction of the microorganism internal-structures and their refractive indices (extracted from a database). Afterwards, the radiative properties are estimated using the Schiff approximation for spheroidal or cylindrical particles, as a first step toward the description of the complexity and diversity of the shapes encountered within the microbial world. Finally, these predictive results are confronted to experimental normal-hemispherical transmittance spectra for validation. This entire procedure is implemented for Rhodospirillum rubrum, Arthrospira platensis and Chlamydomonas reinhardtii, each representative of the main three kinds of photosynthetic microorganisms, i.e. respectively

  10. Tritium transport calculations for the IFMIF Tritium Release Test Module

    Energy Technology Data Exchange (ETDEWEB)

    Freund, Jana, E-mail: jana.freund@kit.edu; Arbeiter, Frederik; Abou-Sena, Ali; Franza, Fabrizio; Kondo, Keitaro

    2014-10-15

    Highlights: • Delivery of material data for the tritium balance in the IFMIF Tritium Release Test Module. • Description of the topological models in TMAP and the adapted fusion-devoted Tritium Permeation Code (FUS-TPC). • Computation of release of tritium from the breeder solid material into the purge gas. • Computation of the loss of tritium over the capsule wall, rig hull, container wall and purge gas return line. - Abstract: The IFMIF Tritium Release Test Module (TRTM) is projected to measure online the tritium release from breeder ceramics and beryllium pebble beds under high energy neutron irradiation. Tritium produced in the pebble bed of TRTM is swept out continuously by a purge gas flow, but can also permeate into the module's metal structures, and can be lost by permeation to the environment. According analyses on the tritium inventory are performed to support IFMIF plant safety studies, and to support the experiment planning. This paper describes the necessary elements for calculation of the tritium transport in the Tritium Release Test Module as follows: (i) applied equations for the tritium balance, (ii) material data from literature and (iii) the topological models and the computation of the five different cases; namely release of tritium from the breeder solid material into the purge gas, loss of tritium over the capsule wall, rig hull, container wall and purge gas return line in detail. The problem of tritium transport in the TRTM has been studied and analyzed by the Tritium Migration Analysis Program (TMAP) and the adapted fusion-devoted Tritium Permeation Code (FUS-TPC). TMAP has been developed at INEEL and now exists in Version 7. FUS-TPC Code was written in MATLAB with the original purpose to study the tritium transport in Helium Cooled Lead Lithium (HCLL) blanket and in a later version the Helium Cooled Pebble Bed (HCPB) blanket by [6] (Franza, 2012). This code has been further modified to be applicable to the TRTM. Results from the

  11. Tritium transport calculations for the IFMIF Tritium Release Test Module

    International Nuclear Information System (INIS)

    Freund, Jana; Arbeiter, Frederik; Abou-Sena, Ali; Franza, Fabrizio; Kondo, Keitaro

    2014-01-01

    Highlights: • Delivery of material data for the tritium balance in the IFMIF Tritium Release Test Module. • Description of the topological models in TMAP and the adapted fusion-devoted Tritium Permeation Code (FUS-TPC). • Computation of release of tritium from the breeder solid material into the purge gas. • Computation of the loss of tritium over the capsule wall, rig hull, container wall and purge gas return line. - Abstract: The IFMIF Tritium Release Test Module (TRTM) is projected to measure online the tritium release from breeder ceramics and beryllium pebble beds under high energy neutron irradiation. Tritium produced in the pebble bed of TRTM is swept out continuously by a purge gas flow, but can also permeate into the module's metal structures, and can be lost by permeation to the environment. According analyses on the tritium inventory are performed to support IFMIF plant safety studies, and to support the experiment planning. This paper describes the necessary elements for calculation of the tritium transport in the Tritium Release Test Module as follows: (i) applied equations for the tritium balance, (ii) material data from literature and (iii) the topological models and the computation of the five different cases; namely release of tritium from the breeder solid material into the purge gas, loss of tritium over the capsule wall, rig hull, container wall and purge gas return line in detail. The problem of tritium transport in the TRTM has been studied and analyzed by the Tritium Migration Analysis Program (TMAP) and the adapted fusion-devoted Tritium Permeation Code (FUS-TPC). TMAP has been developed at INEEL and now exists in Version 7. FUS-TPC Code was written in MATLAB with the original purpose to study the tritium transport in Helium Cooled Lead Lithium (HCLL) blanket and in a later version the Helium Cooled Pebble Bed (HCPB) blanket by [6] (Franza, 2012). This code has been further modified to be applicable to the TRTM. Results from the

  12. Synergism of the method of characteristics and CAD technology for neutron transport calculation

    International Nuclear Information System (INIS)

    Chen, Z.; Wang, D.; He, T.; Wang, G.; Zheng, H.

    2013-01-01

    The method of characteristics (MOC) is a very popular methodology in neutron transport calculation and numerical simulation in recent decades for its unique advantages. One of the key problems determining whether the MOC can be applied in complicated and highly heterogeneous geometry is how to combine an effective geometry processing method with MOC. Most of the existing MOC codes describe the geometry by lines and arcs with extensive input data, such as circles, ellipses, regular polygons and combination of them. Thus they have difficulty in geometry modeling, background meshing and ray tracing for complicated geometry domains. In this study, a new idea making use of a CAD solid modeler MCAM which is a CAD/Image-based Automatic Modeling Program for Neutronics and Radiation Transport developed by FDS Team in China was introduced for geometry modeling and ray tracing of particle transport to remove these geometrical limitations mentioned above. The diamond-difference scheme was applied to MOC to reduce the spatial discretization error of the flat flux approximation in theory. Based on MCAM and MOC, a new MOC code was developed and integrated into SuperMC system, which is a Super Multi-function Computational system for neutronics and radiation simulation. The numerical testing results demonstrated the feasibility and effectiveness of the new idea for geometry treatment in SuperMC. (authors)

  13. 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)

  14. Compendium of Material Composition Data for Radiation Transport Modeling

    Energy Technology Data Exchange (ETDEWEB)

    McConn, Ronald J.; Gesh, Christopher J.; Pagh, Richard T.; Rucker, Robert A.; Williams III, Robert

    2011-03-04

    Introduction Meaningful simulations of radiation transport applications require realistic definitions of material composition and densities. When seeking that information for applications in fields such as homeland security, radiation shielding and protection, and criticality safety, researchers usually encounter a variety of materials for which elemental compositions are not readily available or densities are not defined. Publication of the Compendium of Material Composition Data for Radiation Transport Modeling, Revision 0, in 2006 was the first step toward mitigating this problem. Revision 0 of this document listed 121 materials, selected mostly from the combined personal libraries of staff at the Pacific Northwest National Laboratory (PNNL), and thus had a scope that was recognized at the time to be limited. Nevertheless, its creation did provide a well-referenced source of some unique or hard-to-define material data in a format that could be used directly in radiation transport calculations being performed at PNNL. Moreover, having a single common set of material definitions also helped to standardize at least one aspect of the various modeling efforts across the laboratory by providing separate researchers the ability to compare different model results using a common basis of materials. The authors of the 2006 compendium understood that, depending on its use and feedback, the compendium would need to be revised to correct errors or inconsistencies in the data for the original 121 materials, as well as to increase (per users suggestions) the number of materials listed. This 2010 revision of the compendium has accomplished both of those objectives. The most obvious change is the increased number of materials from 121 to 372. The not-so-obvious change is the mechanism used to produce the data listed here. The data listed in the 2006 document were compiled, evaluated, entered, and error-checked by a group of individuals essentially by hand, providing no library

  15. 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

  16. 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

  17. Radiation transport effects in divertor plasmas generated during a tokamak reactor disruption

    International Nuclear Information System (INIS)

    Peterson, R.R.; MacFarlane, J.J.; Wang, P.

    1994-01-01

    Vaporization of material from tokamak divertors during disruptions is a critical issue for tokamak reactors from ITER to commercial power plants. Radiation transport from the vaporized material onto the remaining divertor surface plays an important role in the total mass loss to the divertor. Radiation transport in such a vapor is very difficult to calculate in full detail, and this paper quantifies the sensitivity of the divertor mass loss to uncertainties in the radiation transport. Specifically, the paper presents the results of computer simulations of the vaporization of a graphite coated divertor during a tokamak disruption with ITER CDA parameters. The results show that a factor of 100 change in the radiation conductivity changes the mass loss by more than a factor of two

  18. 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

  19. Factors affecting radiation doses from dedicated rail transport of spent reactor fuel

    International Nuclear Information System (INIS)

    Martin, J.E.

    1988-01-01

    This paper reports there are two exposure control concerns associated with the shipment of spent reactor fuel in dedicated trains -- compliance with transportation regulations for maximum allowable radiation levels, and minimizing the dose received by the general public. This article examines the methods used to calculate the dose equivalent rates alongside stationary (transport regulations) and moving trains (public exposure) of various lengths. The factors examined include the source term, the effect of overlapping radiation fields, the speed of the train, and the location of the population relative to the train. Trains made up of series of cars that individually meet transport regulations can, as a whole, exceed transport vehicle dose equivalent rate limits by up to 23% due to overlapping radiation fields. For moving trains and the worst case analyzed -- a person located 20 feet from the tracks and a train speed of 5 mph --- 141 rail cars would have to pass by to deliver a dose equivalent of 1 mrem

  20. Calculation of radiation dose rates from a spent nuclear fuel shipping cask

    International Nuclear Information System (INIS)

    Chen, S.Y.; Yuan, Y.C.

    1988-01-01

    Radiation doses from a spent nuclear fuel cask are usually from various phases of operations during handling, shipping, and storage of the casks. Assessment of such doses requires knowledge of external radiation dose rates at various locations surrounding a cask. Under current practices, dose rates from gamma photons are usually estimated by means of point- or line-source approaches incorporating the conventional buildup factors. Although such simplified approaches may at times be easy to use, their accuracy has not been verified. For example, those simplified methods have not taken into account influencing factors such as the geometry of the cask and the presence of the ground surface, and the effects of these factors on the calculated dose rates are largely unknown. Moreover, similar empirical equations for buildup factors currently do not exist for neutrons. The objective of this study is to use a more accurate approach in calculating radiation dose rates for both neutrons and gamma photons from a spent fuel cask. The calculation utilizes the more sophisticated transport method and takes into account the geometry of the cask and the presence of the ground surface. The results of a detailed study of dose rates in the near field (within 20 meters) are presented and, for easy application, the cask centerline dose rates are fitted into empirical equations at cask centerline distances up to 2000 meters from the surface of the cask

  1. User's manual for sustainable transportation performance measures calculator

    Science.gov (United States)

    2010-08-01

    Sustainable transportation can be viewed as the provision of safe, effective, and efficient : access and mobility into the future while considering economic, social, and environmental : needs. For the Texas Department of Transportation (TxDOT) to ass...

  2. 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

  3. 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.

  4. 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

  5. Numerical simulations for radiation hydrodynamics. 2: Transport limit

    International Nuclear Information System (INIS)

    Dai, W.W.; Woodward, P.R.

    2000-01-01

    A finite difference scheme is proposed for two-dimensional radiation hydrodynamical equations in the transport limit. The scheme is of Godunov-type, in which the set of time-averaged flux needed in the scheme is calculated through Riemann problems solved. In the scheme, flow signals are explicitly treated, while radiation signals are implicitly treated. Flow fields and radiation fields are updated simultaneously. An iterative approach is proposed to solve the set of nonlinear algebraic equations arising from the implicitness of the scheme. The sweeping method used in the scheme significantly reduces the number of iterations or computer CPU time needed. A new approach to further accelerate the convergence is proposed, which further reduces the number of iterations needed by more than one order. No matter how many cells radiation signals propagate in one time step, only an extremely small number of iterations are needed in the scheme, and each iteration costs only about 0.8% of computer CPU time which is needed for one time step of a second order accurate and fully explicit scheme. Two-dimensional problems are treated through a dimensionally split technique. Therefore, iterations for solving the set of algebraic equations are carried out only in each one-dimensional sweep. Through numerical examples it is shown that the scheme keeps the principle advantages of Godunov schemes for flow motion. In the time scale of flow motion numerical results are the same as those obtained from a second order accurate and fully explicit scheme. The acceleration of the convergence proposed in this paper may be directly applied to other hyperbolic systems. This study is important for laser fusion and astrophysics

  6. SPACETRAN, Radiation Leakage from Cylinder with ANISN Flux Calculation

    International Nuclear Information System (INIS)

    Cramer, S.N.; Solomito, M.

    1974-01-01

    1 - Nature of physical problem solved: SPACETRAN is designed to calculate the energy-dependent total flux or some proportional quantity such as kerma, due to the radiation leakage from the surface of a right-circular cylinder at detector positions located at arbitrary distances from the surface. The assumptions are made that the radiation emerging from the finite cylinder has no spatial dependence and that a vacuum surrounds the cylinder. 2 - Method of solution: There are three versions of the program in the code package. SPACETRAN-I uses the surface angular fluxes calculated by the discrete ordinates SN code ANISN, as input. SPACETRAN-II assumes that the surface angular flux for all energies can be represented as a function (Cos(PHI))**N, where PHI is the angle between surface outward normal and radiation direction, and N is an integer specified by the user. For both versions the energy group structure and the number and location of detectors is arbitrary. The flux (or response function) for a given energy group at some detection point is computed by summing the contributions from each surface area element over the entire surface. The surface area elements are defined by input data. SPACETRAN-III uses surface angular fluxes from DOT-3. SPACETRAN-I handles contributions either from a cylinder 'end' or 'side', so the total contributions must be obtained by adding the results of separate end and side runs. ANISN angular fluxes are specified for discrete directions. In general, the direction between the detector and contributing area will not exactly coincide with one of these discrete directions. In this case, the ANISN angular flux for the 'closest' discrete direction is used to approximate the contribution to the detector. SPACETRAN-II handles contributions from both the side and end of a cylinder in a single run. Since the assumed angular distribution is specified by a continuous function, it is not necessary to perform the angle selection described above. For

  7. Considerations of beta and electron transport in internal dose calculations. Progress report

    Energy Technology Data Exchange (ETDEWEB)

    Bolch, W.E.

    1994-11-01

    The goal of this particular task is to consider, for the first time, the explicit transport of beta particles and photon-generated electrons in the series of six phantoms developed by Cristy and Eckerman (1987) at the Oak Ridge National Laboratory. In their report, ORNL/TM-8381, specific absorbed fractions of energy are reported for phantoms representing the newborn (3.4 kg), the one-year-old (9.8 kg), the five-year-old (19 kg), the ten-year-old (32 kg), the fifteen-year-old/adult female (55-58 kg), and the adult male (70 kg). Radiation transport calculations were performed with the Monte Carlo code ALGAMP which allows photon transport only. In subsequent calculations of radionuclide S values as is done in the MIRDOSE2 computer program, electron absorbed fractions are thus considered to be either unity or zero depending upon whether the source region does or does not equal the target region, respectively.

  8. Harmonizing carbon footprint calculation for freight transport chains

    NARCIS (Netherlands)

    Lewis, A.; Ehrler, V.; Auvinen, H.; Maurer, H.; Davydenko, I.; Burmeister, A.; Seidel, S.; Lischke, A.; Kiel, J.

    2016-01-01

    The European Commission has set as a target a reduction of 60% in transport greenhouse gas emissions by 2050 [EC 11]. This includes freight transport emissions, which present a particular challenge due to the forecast increase in goods transport linked to future economic growth, the current trend of

  9. Advanced nuclear data for radiation-damage calculations

    International Nuclear Information System (INIS)

    MacFarlane, R.E.; Foster, D.G. Jr.

    1983-01-01

    Accurate calculations of atomic displacement damage in materials exposed to neutrons require detailed spectra for primary recoil nuclei. Such data are not available from direct experimental measurements. Moreover, they cannot always be computed accurately starting from evaluated nuclear data libraries such as ENDF/B-V that were developed primarily for neutron transport applications, because these libraries lack detailed energy-and-angle distributions for outgoing charged particles. Fortunately, a new generation of nuclear model codes is now available that can be used to fill in the missing spectra. One example is the preequilibrium statistical-model code GNASH. For heating and damage applications, a supplementary code called RECOIL has been developed. RECOIL uses detailed reaction data from GNASH, together with angular distributions based on Kalbach-Mann systematics to compute the energy and angle distributions of recoil nuclei. The energy-angle distributions for recoil nuclei and outgoing particles are written out in the new ENDF/B File 6 format. The result is a complete set of nuclear data that can be used to calculate displacement-energy production, heat production, gas production, transmutation, and activation. Sample results for iron are given and compared to the results of conventional damage models such as those used in NJOY

  10. Calculation of Radiation Protection Quantities and Analysis of Astronaut Orientation Dependence

    Science.gov (United States)

    Clowdsley, Martha S.; Nealy, John E.; Atwell, William; Anderson, Brooke M.; Luetke, Nathan J.; Wilson, John W.

    2006-01-01

    Health risk to astronauts due to exposure to ionizing radiation is a primary concern for exploration missions and may become the limiting factor for long duration missions. Methodologies for evaluating this risk in terms of radiation protection quantities such as dose, dose equivalent, gray equivalent, and effective dose are described. Environment models (galactic cosmic ray and solar particle event), vehicle/habitat geometry models, human geometry models, and transport codes are discussed and sample calculations for possible lunar and Mars missions are used as demonstrations. The dependence of astronaut health risk, in terms of dosimetric quantities, on astronaut orientation within a habitat is also examined. Previous work using a space station type module exposed to a proton spectrum modeling the October 1989 solar particle event showed that reorienting the astronaut within the module could change the calculated dose equivalent by a factor of two or more. Here the dose equivalent to various body tissues and the whole body effective dose due to both galactic cosmic rays and a solar particle event are calculated for a male astronaut in two different orientations, vertical and horizontal, in a representative lunar habitat. These calculations also show that the dose equivalent at some body locations resulting from a solar particle event can vary by a factor of two or more, but that the dose equivalent due to galactic cosmic rays has a much smaller (<15%) dependence on astronaut orientation.

  11. 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.

  12. New dose-mortality data based on 3-D radiation shielding calculation for concrete buildings at Nagasaki

    International Nuclear Information System (INIS)

    Rhoades, W.A.; Childs, R.L.; Ingersoll, D.T.

    1988-01-01

    The analysis of radiation doses received during the World War II attack on Nagasaki provides an important source of biochemical information. More than 40 years after the war, it has been possible to make a satisfactory calculation of the doses to personnel inside reinforced concrete buildings by use of a 3-dimensional discrete ordinates code, TORT. The results were used to deduce a new value of the LD50 parameter that is in good agreement with traditional values. The new discrete ordinates software appears to have potential application to conventional radiation transport calculations as well. 9 refs., 3 figs., 2 tabs

  13. Application of maximum values for radiation exposure and principles for the calculation of radiation dose

    International Nuclear Information System (INIS)

    2000-01-01

    The guide sets out the mathematical definitions and principles involved in the calculation of the equivalent dose and the effective dose, and the instructions concerning the application of the maximum values of these quantities. further, for monitoring the dose caused by internal radiation, the guide defines the limits derived from annual dose limits (the Annual Limit on Intake and the Derived Air Concentration). Finally, the guide defines the operational quantities to be used in estimating the equivalent dose and the effective dose, and also sets out the definitions of some other quantities and concepts to be used in monitoring radiation exposure. The guide does not include the calculation of patient doses carried out for the purposes of quality assurance

  14. Application of maximum values for radiation exposure and principles for the calculation of radiation dose

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    The guide sets out the mathematical definitions and principles involved in the calculation of the equivalent dose and the effective dose, and the instructions concerning the application of the maximum values of these quantities. further, for monitoring the dose caused by internal radiation, the guide defines the limits derived from annual dose limits (the Annual Limit on Intake and the Derived Air Concentration). Finally, the guide defines the operational quantities to be used in estimating the equivalent dose and the effective dose, and also sets out the definitions of some other quantities and concepts to be used in monitoring radiation exposure. The guide does not include the calculation of patient doses carried out for the purposes of quality assurance.

  15. Transport analysis of high radiation and high density plasmas in the ASDEX Upgrade tokamak

    Directory of Open Access Journals (Sweden)

    Casali L.

    2014-01-01

    Full Text Available Future fusion reactors, foreseen in the “European road map” such as DEMO, will operate under more demanding conditions compared to present devices. They will require high divertor and core radiation by impurity seeding to reduce heat loads on divertor target plates. In addition, DEMO will have to work at high core densities to reach adequate fusion performance. The performance of fusion reactors depends on three essential parameters: temperature, density and energy confinement time. The latter characterizes the loss rate due to both radiation and transport processes. The DEMO foreseen scenarios described above were not investigated so far, but are now addressed at the ASDEX Upgrade tokamak. In this work we present the transport analysis of such scenarios. Plasma with high radiation by impurity seeding: transport analysis taking into account the radiation distribution shows no change in transport during impurity seeding. The observed confinement improvement is an effect of higher pedestal temperatures which extend to the core via stiffness. A non coronal radiation model was developed and compared to the bolometric measurements in order to provide a reliable radiation profile for transport calculations. High density plasmas with pellets: the analysis of kinetic profiles reveals a transient phase at the start of the pellet fuelling due to a slower density build up compared to the temperature decrease. The low particle diffusion can explain the confinement behaviour.

  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. Use of realistic anthropomorphic models for calculation of radiation dose in nuclear medicine

    International Nuclear Information System (INIS)

    Stabin, Michael G.; Emmons, Mary A.; Fernald, Michael J.; Brill, A.B.; Segars, W.Paul

    2008-01-01

    Anthropomorphic phantoms based on simple geometric structures have been used in radiation dose calculations for many years. We have now developed a series of anatomically realistic phantoms representing adults and children using body models based on non-uniform rational B-spline (NURBS), with organ and body masses based on the reference values given in ICRP Publication 89. Age-dependent models were scaled and shaped to represent the reference individuals described in ICRP 89 (male and female adults, newborns, 1-, 5-, 10- and 15-year-olds), using a software tool developed in Visual C++. Voxel-based versions of these models were used with GEANT4 radiation transport codes for calculation of specific absorbed fractions (SAFs) for internal sources of photons and electrons, using standard starting energy values. Organ masses in the models were within a few % of ICRP reference masses, and physicians reviewed the models for anatomical realism. Development of individual phantoms was much faster than manual segmentation of medical images, and resulted in a very uniform standardized phantom series. SAFs were calculated on the Vanderbilt multi node computing network (ACCRE). Photon and electron SAFs were calculated for all organs in all models, and were compared to values from similar phantoms developed by others. Agreement was very good in most cases; some differences were seen, due to differences in organ mass and geometry. This realistic phantom series represents a possible replacement for the Cristy/Eckerman series of the 1980's. Both phantom sets will be included in the next release of the OLINDA/EXM personal computer code, and the new phantoms will be made generally available to the research community for other uses. Calculated radiation doses for diagnostic and therapeutic radiopharmaceuticals will be compared with previous values. (author)

  18. 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.

  19. Numerical procedure for the calculation of nonsteady spherical shock fronts with radiation

    International Nuclear Information System (INIS)

    Winkler, K.H.

    The basis of the numerical method is an implicit difference scheme with time backward differences to a freely moving coordinate system. The coordinate system itself is determined simultaneously with the iterative solution of the physical equations as a function of the physical variables. Shock fronts, even nonsteady ones, are calculated as discontinuities according to the Rankine--Hugoniot equations. The radiation field is obtained from the two-dimensional, static, spherically symmetric transport equation in conjunction with the time-dependent one-dimensional moment equations. No artificial viscosity of any type is ever used. The applicability of the method developed is demonstrated by an example involving the calculation of protostar collapse. 11 figures

  20. Flux and brightness calculations for various synchrotron radiation sources

    International Nuclear Information System (INIS)

    Weber, J.M.; Hulbert, S.L.

    1991-11-01

    Synchrotron radiation (SR) storage rings are powerful scientific and technological tools. The first generation of storage rings in the US., e.g., SURF (Washington, D.C.), Tantalus (Wisconsin), SSRL (Stanford), and CHESS (Cornell), revolutionized VUV, soft X-ray, and hard X-ray science. The second (present) generation of storage rings, e.g. the NSLS VUV and XRAY rings and Aladdin (Wisconsin), have sustained the revolution by providing higher stored currents and up to a factor of ten smaller electron beam sizes than the first generation sources. This has made possible a large number of experiments that could not performed using first generation sources. In addition, the NSLS XRAY ring design optimizes the performance of wigglers (high field periodic magnetic insertion devices). The third generation storage rings, e.g. ALS (Berkeley) and APS (Argonne), are being designed to optimize the performance of undulators (low field periodic magnetic insertion devices). These extremely high brightness sources will further revolutionize x-ray science by providing diffraction-limited x-ray beams. The output of undulators and wigglers is distinct from that of bending magnets in magnitude, spectral shape, and in spatial and angular size. Using published equations, we have developed computer programs to calculate the flux, central intensity, and brightness output bending magnets and selected wigglers and undulators of the NSLS VUV and XRAY rings, the Advanced Light Source (ALS), and the Advanced Photon Source (APS). Following is a summary of the equations used, the graphs and data produced, and the computer codes written. These codes, written in the C programming language, can be used to calculate the flux, central intensity, and brightness curves for bending magnets and insertion devices on any storage ring

  1. Computer program for calculating thermodynamic and transport properties of fluids

    Science.gov (United States)

    Hendricks, R. C.; Braon, A. K.; Peller, I. C.

    1975-01-01

    Computer code has been developed to provide thermodynamic and transport properties of liquid argon, carbon dioxide, carbon monoxide, fluorine, helium, methane, neon, nitrogen, oxygen, and parahydrogen. Equation of state and transport coefficients are updated and other fluids added as new material becomes available.

  2. 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.

  3. 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.

  4. Measurement and Calculation of Gamma Radiation from HWZPR Reactor

    International Nuclear Information System (INIS)

    Jalali, Majid

    2006-01-01

    HWZPR is a research reactor with natural uranium fuel, D 2 O moderator and graphite reflector with maximum power of 100 W. It is a suitable means for theoretical research and heavy water reactor experiments. Neutrons from the core participate in different nuclear reactions by interactions with fuel, moderator, graphite and the concrete around the reactor. The results of these interactions are the production of prompt gammas in the environment. Useful information is gained by the reactor gamma spectrum measurement from point of view of relative quantity and energy distribution of direct and scattered radiations. Reactor gamma ray spectrum has been gathered in different places around the reactor by HPGe detector. In analysis of these spectra, 1 H(n,γ) 2 H, 16 O(n,n'γ) 16 O, 2 H(n,γ) 3 H and 238 U(n,γ) 239 U reactions occurring in reactor moderator and fuel, are important. The measured spectrum has been primarily estimated by the MCNP code. There is agreement between the code and the experiments in some points. The scattered gamma rays from 27 Al (n,γ) 28 Al reaction in the reactor tank, are the most among the gammas scattered in the reactor environment. Also the dose calculations by MCNP code show that 72% of gamma dose belongs to the energy range 3-11 MeV from reactor gamma spectrum and the danger of exposure from the reactor high-energy photons is serious. (author)

  5. Tips for deciphering and quick calculation of radiation spectra

    Science.gov (United States)

    Bondarenco, M. V.

    2018-04-01

    Radiation spectra from ultra-relativistic electrons in thin [Tll lf(ω)] and thick [Tgg lf(ω)] targets are discussed. The method of simplified averaging is described by examples of Landau-Pomeranchuk-Migdal effect and radiation at doughnut scattering. General infrared and ultraviolet asymptotic properties of radiation spectra are discussed.

  6. 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

  7. Density functional theory calculations of charge transport properties ...

    Indian Academy of Sciences (India)

    ZIRAN CHEN

    2017-08-04

    Aug 4, 2017 ... properties of 'plate-like' coronene topological structures ... Keywords. Organic semiconductors; density functional theory; charge carrier mobility; ambipolar transport; ..... nology Department of Sichuan Province (Grant Number.

  8. 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.

  9. 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.

  10. 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

  11. 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.

  12. Quasi solution of radiation transport equation

    International Nuclear Information System (INIS)

    Pogosbekyan, L.R.; Lysov, D.A.

    1995-01-01

    There is uncertainty with experimental data as well as with input data of theoretical calculations. The neutron distribution from the variational principle, which takes into account both theoretical and experimental data, is obtained to increase the accuracy and speed of neutronic calculations. The neutron imbalance in mesh cells and the discrepancy between experimentally measured and calculated functional of the neutron distribution are simultaneously minimized. A fast-working and simple-programming iteration method is developed to minimize the objective functional. The method can be used in the core monitoring and control system for (a) power distribution calculations, (b) in- and ex-core detector calibration, (c) macro-cross sections or isotope distribution correction by experimental data, and (d) core and detector diagnostics

  13. MORSE-C, Neutron Transport, Gamma Transport for Criticality Calculation by Monte-Carlo Method

    International Nuclear Information System (INIS)

    2002-01-01

    1 - Description of program or function: MORSE-C is a Monte-Carlo code to solve the multiple energy group form of the Boltzmann transport equation in order to obtain the eigenvalue (multiplication) when fissionable materials are present. Cross sections for up to 100 energy groups may be employed. The angular scattering is treated by the usual Legendre expansion as used in the discrete ordinates codes. Up-scattering may be specified. The geometry is defined by relationships to general 1. or 2. degree surfaces. Array units may be specified. Output includes, besides the usual values of input quantities, plots of the geometry, calculated volumes and masses, and graphs of results to assist the user in determining the correctness of the problem's solution

  14. 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

  15. Monte Carlo perturbation theory in neutron transport calculations

    International Nuclear Information System (INIS)

    Hall, M.C.G.

    1980-01-01

    The need to obtain sensitivities in complicated geometrical configurations has resulted in the development of Monte Carlo sensitivity estimation. A new method has been developed to calculate energy-dependent sensitivities of any number of responses in a single Monte Carlo calculation with a very small time penalty. This estimation typically increases the tracking time per source particle by about 30%. The method of estimation is explained. Sensitivities obtained are compared with those calculated by discrete ordinates methods. Further theoretical developments, such as second-order perturbation theory and application to k/sub eff/ calculations, are discussed. The application of the method to uncertainty analysis and to the analysis of benchmark experiments is illustrated. 5 figures

  16. Lecture note on neutron and photon transport calculation with MCNP

    International Nuclear Information System (INIS)

    Sakurai, Kiyoshi

    2003-01-01

    This paper is a lecture note on the continuous energy Monte Carlo method. The contents are as follows; history of the Monte Carlo study, continuous energy Monte Carlo codes, libraries, evaluation method for calculation results, integral emergent particle density equation, pseudorandom number, random walk, variance reduction techniques, MCNP weight window method, MCNP weight window generator, exponential transform, estimators, criticality problem and research subjects. This paper is a textbook for beginners on the Monte Carlo calculation. (author)

  17. 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

  18. Beam transport calculations for the EN tandem installation

    International Nuclear Information System (INIS)

    Sparks, R.J.

    1980-12-01

    Transport of a charged particle beam through the new EN tandem accelerator installation of the Institute of Nuclear Sciences has been analysed using simplified mathematical models. The purpose is to identify the factors affecting transmission of the beam, and to arrive at a design for the system to inject the beam into the accelerator

  19. Cyclic machine scheduling with tool transportation - additional calculations

    NARCIS (Netherlands)

    Kuijpers, C.M.H.

    2001-01-01

    In the PhD Thesis of Kuijpers a cyclic machine scheduling problem with tool transportation is considered. For the problem with two machines, it is shown that there always exists an optimal schedule with a certain structure. This is done by means of an elaborate case study. For a number of cases some

  20. New physical model calculates airline crews' radiation exposure

    Science.gov (United States)

    Schultz, Colin

    2013-12-01

    Airline pilots and crews, who spend hundreds of hours each year flying at high altitude, are exposed to increased doses of radiation from galactic cosmic rays and solar energy particles, enough that airline crew members are actually considered radiation workers by the International Commission on Radiological Protection.

  1. 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.

  2. 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

  3. 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

  4. 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

  5. Global transport calculations with an equivalent barotropic system

    Science.gov (United States)

    Salby, Murry L.; O'Sullivan, Donal; Garcia, Rolando R.; Tribbia, Joseph

    1990-01-01

    Transport properties of the two-dimensional equations governing equivalent barotropic motion are investigated on the sphere. This system has ingredients such as forcing, equivalent depth, and thermal dissipation explicitly represented, and takes into account compression effects associated with vertical motion along isentropic surfaces. Horizontal transport properties of this system are investigated under adiabatic and diabatic conditions for different forms of dissipation, and over a range of resolutions. It is shown that forcing represetative of time-mean and amplified conditions at 10 mb leads to the behavior typical of observations at this level. The displacement of the polar night vortex and its distortion into a comma shape are evident, as is irreversible mixing under sufficiently strong forcing amplitude. It is shown that thermal dissipation influences the behavior significantly by inhibiting the amplification of unstable eddies and thereby the horizontal stirring of air.

  6. 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.)

  7. Concise four-vector scheme for neutron transport calculations

    International Nuclear Information System (INIS)

    Kulacsy, K.; Lukacs, B.; Racz, A.

    1995-01-01

    An explicit Riemannian geometrical form or the vectorial Neutron Streaming Term is presented. The method applies the full Riemannian technique of general covariance. There are cases when the symmetry of the neutron flux must be smaller than that of the arrangement. However, in coordinate space there are always solutions of the Neutron Transport Equation as symmetric as the arrangement, if the latter's symmetry is at least an affine collineation of the Euclidian 3-space. (author). 7 refs

  8. Recently developed methods in neutral-particle transport calculations: overview

    International Nuclear Information System (INIS)

    Alcouffe, R.E.

    1982-01-01

    It has become increasingly apparent that successful, general methods for the solution of the neutral particle transport equation involve a close connection between the spatial-discretization method used and the source-acceleration method chosen. The first form of the transport equation, angular discretization which is discrete ordinates is considered as well as spatial discretization based upon a mesh arrangement. Characteristic methods are considered briefly in the context of future, desirable developments. The ideal spatial-discretization method is described as having the following attributes: (1) positive-positive boundary data yields a positive angular flux within the mesh including its boundaries; (2) satisfies the particle balance equation over the mesh, that is, the method is conservative; (3) possesses the diffusion limit independent of spatial mesh size, that is, for a linearly isotropic flux assumption, the transport differencing reduces to a suitable diffusion equation differencing; (4) the method is unconditionally acceleratable, i.e., for each mesh size, the method is unconditionally convergent with a source iteration acceleration. It is doubtful that a single method possesses all these attributes for a general problem. Some commonly used methods are outlined and their computational performance and usefulness are compared; recommendations for future development are detailed, which include practical computational considerations

  9. Calculation of the Coherent Synchrotron Radiation Impedance from a Wiggler

    International Nuclear Information System (INIS)

    Wu, Juhao

    2003-01-01

    Most studies of Coherent Synchrotron Radiation (CSR) have only considered the radiation from independent dipole magnets. However, in the damping rings of future linear colliders, a large fraction of the radiation power will be emitted in damping wigglers. In this paper, the longitudinal wakefield and impedance due to CSR in a wiggler are derived in the limit of a large wiggler parameter K. After an appropriate scaling, the results can be expressed in terms of universal functions, which are independent of K. Analytical asymptotic results are obtained for the wakefield in the limit of large and small distances, and for the impedance in the limit of small and high frequencies

  10. PABLM: a computer program to calculate accumulated radiation doses from radionuclides in the environment

    International Nuclear Information System (INIS)

    Napier, B.A.; Kennedy, W.E. Jr.; Soldat, J.K.

    1980-03-01

    A computer program, PABLM, was written to facilitate the calculation of internal radiation doses to man from radionuclides in food products and external radiation doses from radionuclides in the environment. This report contains details of mathematical models used and calculational procedures required to run the computer program. Radiation doses from radionuclides in the environment may be calculated from deposition on the soil or plants during an atmospheric or liquid release, or from exposure to residual radionuclides in the environment after the releases have ended. Radioactive decay is considered during the release of radionuclides, after they are deposited on the plants or ground, and during holdup of food after harvest. The radiation dose models consider several exposure pathways. Doses may be calculated for either a maximum-exposed individual or for a population group. The doses calculated are accumulated doses from continuous chronic exposure. A first-year committed dose is calculated as well as an integrated dose for a selected number of years. The equations for calculating internal radiation doses are derived from those given by the International Commission on Radiological Protection (ICRP) for body burdens and MPC's of each radionuclide. The radiation doses from external exposure to contaminated water and soil are calculated using the basic assumption that the contaminated medium is large enough to be considered an infinite volume or plane relative to the range of the emitted radiations. The equations for calculations of the radiation dose from external exposure to shoreline sediments include a correction for the finite width of the contaminated beach

  11. Calculation of double energy angle differential neutron albedos for radiation shielding applications

    International Nuclear Information System (INIS)

    Litaize, O.; Diop, C.M.; Nimal, J.C.

    2000-01-01

    Void radiation shielding problems can be dealt with albedo concept which is an alternative to the complex bringing into operation of the 'exact' transport method calculations (SN, Monte Carlo). Up to here, differential albedos are used for single reflections from walls in the NARCISSE-3 propagation albedo code developed at CEA and used for project calculations. For taking into account the neutron multiple reflections on lacunar medium walls, double energy-angle differential albedos are needed. TRIPOLI-4 neutral particle transport Monte Carlo code in three dimensional geometries, has been chosen to implement a double differential albedo calculus routine and therefore to generate albedo data for different kinds of medium. The surfacic estimator, which could be used, is not enough efficient because all neutrons do not contribute to the result. A new estimator is carried out. At each collision site, during the neutron history simulation, it allows to compute the probability of the neutron to go through the medium and to come through the reflection surface in the direction and at the energy considered. This estimator is about hundred times more efficient than the surfacic estimator. (author)

  12. Ab Initio Calculations of Transport in Titanium and Aluminum Mixtures

    Science.gov (United States)

    Walker, Nicholas; Novak, Brian; Tam, Ka Ming; Moldovan, Dorel; Jarrell, Mark

    In classical molecular dynamics simulations, the self-diffusion and shear viscosity of titanium about the melting point have fallen within the ranges provided by experimental data. However, the experimental data is difficult to collect and has been rather scattered, making it of limited value for the validation of these calculations. By using ab initio molecular dynamics simulations within the density functional theory framework, the classical molecular dynamics data can be validated. The dynamical data from the ab initio molecular dynamics can also be used to calculate new potentials for use in classical molecular dynamics, allowing for more accurate classical dynamics simulations for the liquid phase. For metallic materials such as titanium and aluminum alloys, these calculations are very valuable due to an increasing demand for the knowledge of their thermophysical properties that drive the development of new materials. For example, alongside knowledge of the surface tension, viscosity is an important input for modeling the additive manufacturing process at the continuum level. We are developing calculations of the viscosity along with the self-diffusion for aluminum, titanium, and titanium-aluminum alloys with ab initio molecular dynamics. Supported by the National Science Foundation through cooperative agreement OIA-1541079 and the Louisiana Board of Regents.

  13. 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.

  14. Calculation of radiation fields for Hanford tank riser modifications

    International Nuclear Information System (INIS)

    Schwarz, R.A.; Carter, L.L.; Hillesland, K.E.

    1994-10-01

    A visualization process has been created to animate the dose rates from radiation sources as the source and shielding configuration are modified. This process has been illustrated with the removal of a mixer pump from a Hanford Site waste tank

  15. 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.

  16. Development of Monte Carlo decay gamma-ray transport calculation system

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Satoshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Kawasaki, Nobuo [Fujitsu Ltd., Tokyo (Japan); Kume, Etsuo [Japan Atomic Energy Research Inst., Center for Promotion of Computational Science and Engineering, Tokai, Ibaraki (Japan)

    2001-06-01

    In the DT fusion reactor, it is critical concern to evaluate the decay gamma-ray biological dose rates after the reactor shutdown exactly. In order to evaluate the decay gamma-ray biological dose rates exactly, three dimensional Monte Carlo decay gamma-ray transport calculation system have been developed by connecting the three dimensional Monte Carlo particle transport calculation code and the induced activity calculation code. The developed calculation system consists of the following four functions. (1) The operational neutron flux distribution is calculated by the three dimensional Monte Carlo particle transport calculation code. (2) The induced activities are calculated by the induced activity calculation code. (3) The decay gamma-ray source distribution is obtained from the induced activities. (4) The decay gamma-rays are generated by using the decay gamma-ray source distribution, and the decay gamma-ray transport calculation is conducted by the three dimensional Monte Carlo particle transport calculation code. In order to reduce the calculation time drastically, a biasing system for the decay gamma-ray source distribution has been developed, and the function is also included in the present system. In this paper, the outline and the detail of the system, and the execution example are reported. The evaluation for the effect of the biasing system is also reported. (author)

  17. An improved filtered spherical harmonic method for transport calculations

    International Nuclear Information System (INIS)

    Ahrens, C.; Merton, S.

    2013-01-01

    Motivated by the work of R. G. McClarren, C. D. Hauck, and R. B. Lowrie on a filtered spherical harmonic method, we present a new filter for such numerical approximations to the multi-dimensional transport equation. In several test problems, we demonstrate that the new filter produces results with significantly less Gibbs phenomena than the filter used by McClarren, Hauck and Lowrie. This reduction in Gibbs phenomena translates into propagation speeds that more closely match the correct propagation speed and solutions that have fewer regions where the scalar flux is negative. (authors)

  18. PKI, Gamma Radiation Reactor Shielding Calculation by Point-Kernel Method

    International Nuclear Information System (INIS)

    Li Chunhuai; Zhang Liwu; Zhang Yuqin; Zhang Chuanxu; Niu Xihua

    1990-01-01

    1 - Description of program or function: This code calculates radiation shielding problem of gamma-ray in geometric space. 2 - Method of solution: PKI uses a point kernel integration technique, describes radiation shielding geometric space by using geometric space configuration method and coordinate conversion, and makes use of calculation result of reactor primary shielding and flow regularity in loop system for coolant

  19. Calculation of radiation losses in cylinder symmetric high pressure discharges by means of a digital computer

    International Nuclear Information System (INIS)

    Andriessen, F.J.; Boerman, W.; Holtz, I.F.E.M.

    1973-08-01

    Computer calculations have been made of radiative energy losses in a cylindrically symmetric high pressure discharge. The calculations show that the radiation losses which occur in discharges at pressures of a few atmospheres and central temperatures of about 20000degK when compared with the electrical energy supplied, are only of importance in the neighbourhood of the centre of discharge

  20. 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.

  1. 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.

  2. Classical calculation of radiative lifetimes of atomic hydrogen in a homogeneous magnetic field

    International Nuclear Information System (INIS)

    Horbatsch, M.W.; Hessels, E.A.; Horbatsch, M.

    2005-01-01

    Radiative lifetimes of hydrogenic atoms in a homogeneous magnetic field of moderate strength are calculated on the basis of classical radiation. The modifications of the Keplerian orbits due to the magnetic field are incorporated by classical perturbation theory. The model is complemented by a classical radiative decay calculation using the radiated Larmor power. A recently derived highly accurate formula for the transition rate of a field-free hydrogenic state is averaged over the angular momentum oscillations caused by the magnetic field. The resulting radiative lifetimes for diamagnetic eigenstates classified by n,m and the diamagnetic energy shift C compare well with quantum results

  3. 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...

  4. Aberration analysis calculations for synchrotron radiation beamline design

    International Nuclear Information System (INIS)

    McKinney, W.R.; Howells, M.; Padmore, H.A.

    1997-09-01

    The application of ray deviation calculations based on aberration coefficients for a single optical surface for the design of beamline optical systems is reviewed. A systematic development is presented which allows insight into which aberration may be causing the rays to deviate from perfect focus. A new development allowing analytical calculation of line shape is presented

  5. CALCULATION OF THE COHERENT RADIATION IMPDENACE FROM A WIGGLER

    International Nuclear Information System (INIS)

    Wu, J

    2004-01-01

    Most studies of coherent synchrotron radiation (CSR) have considered only the radiation from independent dipole magnets. However, in the damping rings of future linear colliders and many high luminosity factories, a large fraction of the radiation power will be emitted in damping wigglers. In this paper, the longitudinal wakefield and impedance due to CSR in a wiggler are derived in the limit of a large wiggler parameter K. After an appropriate scaling, the results can be expressed in terms of universal functions, which are independent of K. Analytical asymptotic results are obtained for the wakefield in the limit of large and small distances, and for the impedance in the limit of small and high frequencies

  6. 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

  7. One-group transport theory calculation for three slabs cells

    International Nuclear Information System (INIS)

    Maia, C.R.M.

    1979-01-01

    As an idealized model of plate type fuel assemblies for nuclear reactors, three-slab cells are analysed numerically based on the exact solution of the transport equation in the one-group isotropic scattering model. From the equations describing the interface conditions, a set of regular integral equations for the coefficients of the singular eigenfunctions expansions is derived using the half-range orthogonality relations of the eigenfunctions and the recently developed method of regularization. Numerical solutions are obtained by solving this set of equations iteratively. The thermal utilization factor and thermal disadvantage factors as well as flux and current distributions are reported for the first time for various sets of parameters. The accuracy of the P sub(N) approximations is also analysed compared to the exact results. (Author) [pt

  8. Approximate models for neutral particle transport calculations in ducts

    International Nuclear Information System (INIS)

    Ono, Shizuca

    2000-01-01

    The problem of neutral particle transport in evacuated ducts of arbitrary, but axially uniform, cross-sectional geometry and isotropic reflection at the wall is studied. The model makes use of basis functions to represent the transverse and azimuthal dependences of the particle angular flux in the duct. For the approximation in terms of two basis functions, an improvement in the method is implemented by decomposing the problem into uncollided and collided components. A new quadrature set, more suitable to the problem, is developed and generated by one of the techniques of the constructive theory of orthogonal polynomials. The approximation in terms of three basis functions is developed and implemented to improve the precision of the results. For both models of two and three basis functions, the energy dependence of the problem is introduced through the multigroup formalism. The results of sample problems are compared to literature results and to results of the Monte Carlo code, MCNP. (author)

  9. Benchmark calculations in multigroup and multidimensional time-dependent transport

    International Nuclear Information System (INIS)

    Ganapol, B.D.; Musso, E.; Ravetto, P.; Sumini, M.

    1990-01-01

    It is widely recognized that reliable benchmarks are essential in many technical fields in order to assess the response of any approximation to the physics of the problem to be treated and to verify the performance of the numerical methods used. The best possible benchmarks are analytical solutions to paradigmatic problems where no approximations are actually introduced and the only error encountered is connected to the limitations of computational algorithms. Another major advantage of analytical solutions is that they allow a deeper understanding of the physical features of the model, which is essential for the intelligent use of complicated codes. In neutron transport theory, the need for benchmarks is particularly great. In this paper, the authors propose to establish accurate numerical solutions to some problems concerning the migration of neutron pulses. Use will be made of the space asymptotic theory, coupled with a Laplace transformation inverted by a numerical technique directly evaluating the inversion integral

  10. The Suppression of Energy Discretization Errors in Multigroup Transport Calculations

    International Nuclear Information System (INIS)

    Larsen, Edward

    2013-01-01

    The Objective of this project is to develop, implement, and test new deterministric methods to solve, as efficiently as possible, multigroup neutron transport problems having an extremely large number of groups. Our approach was to (i) use the standard CMFD method to 'coarsen' the space-angle grid, yielding a multigroup diffusion equation, and (ii) use a new multigrid-in-space-and-energy technique to efficiently solve the multigroup diffusion problem. The overall strategy of (i) how to coarsen the spatial an energy grids, and (ii) how to navigate through the various grids, has the goal of minimizing the overall computational effort. This approach yields not only the fine-grid solution, but also coarse-group flux-weighted cross sections that can be used for other related problems.

  11. DANTSYS: a system for deterministic, neutral particle transport calculations

    Energy Technology Data Exchange (ETDEWEB)

    Alcouffe, R.E.; Baker, R.S.

    1996-12-31

    The THREEDANT code is the latest addition to our system of codes, DANTSYS, which perform neutral particle transport computations on a given system of interest. The system of codes is distinguished by geometrical or symmetry considerations. For example, ONEDANT and TWODANT are designed for one and two dimensional geometries respectively. We have TWOHEX for hexagonal geometries, TWODANT/GQ for arbitrary quadrilaterals in XY and RZ geometry, and THREEDANT for three-dimensional geometries. The design of this system of codes is such that they share the same input and edit module and hence the input and output is uniform for all the codes (with the obvious additions needed to specify each type of geometry). The codes in this system are also designed to be general purpose solving both eigenvalue and source driven problems. In this paper we concentrate on the THREEDANT module since there are special considerations that need to be taken into account when designing such a module. The main issues that need to be addressed in a three-dimensional transport solver are those of the computational time needed to solve a problem and the amount of storage needed to accomplish that solution. Of course both these issues are directly related to the number of spatial mesh cells required to obtain a solution to a specified accuracy, but is also related to the spatial discretization method chosen and the requirements of the iteration acceleration scheme employed as will be noted below. Another related consideration is the robustness of the resulting algorithms as implemented; because insistence on complete robustness has a significant impact upon the computation time. We address each of these issues in the following through which we give reasons for the choices we have made in our approach to this code. And this is useful in outlining how the code is evolving to better address the shortcomings that presently exist.

  12. Radiation inactivation target size of rat adipocyte glucose transporters in the plasma membrane and intracellular pools

    International Nuclear Information System (INIS)

    Jacobs, D.B.; Berenski, C.J.; Spangler, R.A.; Jung, C.Y.

    1987-01-01

    The in situ assembly states of the glucose transport carrier protein in the plasma membrane and in the intracellular (microsomal) storage pool of rat adipocytes were assessed by studying radiation-induced inactivation of the D-glucose-sensitive cytochalasin B binding activities. High energy radiation inactivated the glucose-sensitive cytochalasin B binding of each of these membrane preparations by reducing the total number of the binding sites without affecting the dissociation constant. The reduction in total number of binding sites was analyzed as a function of radiation dose based on target theory, from which a radiation-sensitive mass (target size) was calculated. When the plasma membranes of insulin-treated adipocytes were used, a target size of approximately 58,000 daltons was obtained. For adipocyte microsomal membranes, we obtained target sizes of approximately 112,000 and 109,000 daltons prior to and after insulin treatment, respectively. In the case of microsomal membranes, however, inactivation data showed anomalously low radiation sensitivities at low radiation doses, which may be interpreted as indicating the presence of a radiation-sensitive inhibitor. These results suggest that the adipocyte glucose transporter occurs as a monomer in the plasma membrane while existing in the intracellular reserve pool either as a homodimer or as a stoichiometric complex with a protein of an approximately equal size

  13. 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.

  14. Monte Carlo calculations of electron transport on microcomputers

    International Nuclear Information System (INIS)

    Chung, Manho; Jester, W.A.; Levine, S.H.; Foderaro, A.H.

    1990-01-01

    In the work described in this paper, the Monte Carlo program ZEBRA, developed by Berber and Buxton, was converted to run on the Macintosh computer using Microsoft BASIC to reduce the cost of Monte Carlo calculations using microcomputers. Then the Eltran2 program was transferred to an IBM-compatible computer. Turbo BASIC and Microsoft Quick BASIC have been used on the IBM-compatible Tandy 4000SX computer. The paper shows the running speed of the Monte Carlo programs on the different computers, normalized to one for Eltran2 on the Macintosh-SE or Macintosh-Plus computer. Higher values refer to faster running times proportionally. Since Eltran2 is a one-dimensional program, it calculates energy deposited in a semi-infinite multilayer slab. Eltran2 has been modified to a two-dimensional program called Eltran3 to computer more accurately the case with a point source, a small detector, and a short source-to-detector distance. The running time of Eltran3 is about twice as long as that of Eltran2 for a similar case

  15. Interface methods for hybrid Monte Carlo-diffusion radiation-transport simulations

    International Nuclear Information System (INIS)

    Densmore, Jeffery D.

    2006-01-01

    Discrete diffusion Monte Carlo (DDMC) is a technique for increasing the efficiency of Monte Carlo simulations in diffusive media. An important aspect of DDMC is the treatment of interfaces between diffusive regions, where DDMC is used, and transport regions, where standard Monte Carlo is employed. Three previously developed methods exist for treating transport-diffusion interfaces: the Marshak interface method, based on the Marshak boundary condition, the asymptotic interface method, based on the asymptotic diffusion-limit boundary condition, and the Nth-collided source technique, a scheme that allows Monte Carlo particles to undergo several collisions in a diffusive region before DDMC is used. Numerical calculations have shown that each of these interface methods gives reasonable results as part of larger radiation-transport simulations. In this paper, we use both analytic and numerical examples to compare the ability of these three interface techniques to treat simpler, transport-diffusion interface problems outside of a more complex radiation-transport calculation. We find that the asymptotic interface method is accurate regardless of the angular distribution of Monte Carlo particles incident on the interface surface. In contrast, the Marshak boundary condition only produces correct solutions if the incident particles are isotropic. We also show that the Nth-collided source technique has the capacity to yield accurate results if spatial cells are optically small and Monte Carlo particles are allowed to undergo many collisions within a diffusive region before DDMC is employed. These requirements make the Nth-collided source technique impractical for realistic radiation-transport calculations

  16. Calculation of three-dimensional groundwater transport using second-order moments

    International Nuclear Information System (INIS)

    Pepper, D.W.; Stephenson, D.E.

    1987-01-01

    Groundwater transport of contaminants from the F-Area seepage basin at the Savannah River Plant (SRP) was calculated using a three-dimensional, second-order moment technique. The numerical method calculates the zero, first, and second moment distributions of concentration within a cell volume. By summing the moments over the entire solution domain, and using a Lagrangian advection scheme, concentrations are transported without numerical dispersion errors. Velocities obtained from field tests are extrapolated and interpolated to all nodal points; a variational analysis is performed over the three-dimensional velocity field to ensure mass consistency. Transport predictions are calculated out to 12,000 days. 28 refs., 9 figs

  17. 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

  18. 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

  19. Calculations in the Wheeler-Feynman absorber theory of radiation

    International Nuclear Information System (INIS)

    Balaji, K.S.

    1986-01-01

    One dimensional computer aided calculations were done to find the self consistent solutions for various absorber configurations in the context of the Wheeler-Feynman absorber theory, wherein every accelerating charge is assumed to produce a time symmetric combination of advanced and retarded fields. These calculations picked out the so called outerface solution for incomplete absorbers and showed that advanced as well as retarded signals interact with matter in the same manner as in the full retarded theory. Based on these calculations, the Partridge experiment and the Schmidt-Newman experiment were ruled out as tests of the absorber theory. An experiment designed to produce and detect advanced effects is proposed, based on more one-dimensional calculations

  20. A sub-structure method for multidimensional integral transport calculations

    International Nuclear Information System (INIS)

    Kavenoky, A.; Stankovski, Z.

    1983-03-01

    A new method has been developed for fine structure burn-up calculations of very heterogeneous large size media. It is a generalization of the well-known surface-source method, allowing coupling actual two-dimensional heterogeneous assemblies, called sub-structures. The method has been applied to a rectangular medium, divided into sub-structures, containing rectangular and/or cylindrical fuel, moderator and structure elements. The sub-structures are divided into homogeneous zones. A zone-wise flux expansion is used to formulate a direct collision probability problem within it (linear or flat flux expansion in the rectangular zones, flat flux in the others). The coupling of the sub-structures is performed by making extra assumptions on the currents entering and leaving the interfaces. The accuracies and computing times achieved are illustrated by numerical results on two benchmark problems

  1. Nonlinear acceleration of S_n transport calculations

    International Nuclear Information System (INIS)

    Fichtl, Erin D.; Warsa, James S.; Calef, Matthew T.

    2011-01-01

    The use of nonlinear iterative methods, Jacobian-Free Newton-Krylov (JFNK) in particular, for solving eigenvalue problems in transport applications has recently become an active subject of research. While JFNK has been shown to be effective for k-eigenvalue problems, there are a number of input parameters that impact computational efficiency, making it difficult to implement efficiently in a production code using a single set of default parameters. We show that different selections for the forcing parameter in particular can lead to large variations in the amount of computational work for a given problem. In contrast, we employ a nonlinear subspace method that sits outside and effectively accelerates nonlinear iterations of a given form and requires only a single input parameter, the subspace size. It is shown to consistently and significantly reduce the amount of computational work when applied to fixed-point iteration, and this combination of methods is shown to be more efficient than JFNK for our application. (author)

  2. Discontinuous finite element treatment of duct problems in transport calculations

    International Nuclear Information System (INIS)

    Mirza, A. M.; Qamar, S.

    1998-01-01

    A discontinuous finite element approach is presented to solve the even-parity Boltzmann transport equation for duct problems. Presence of ducts in a system results in the streaming of particles and hence requires the employment of higher order angular approximations to model the angular flux. Conventional schemes based on the use of continuous trial functions require the same order of angular approximations to be used everywhere in the system, resulting in wastage of computational resources. Numerical investigations for the test problems presented in this paper indicate that the discontinuous finite elements eliminate the above problems and leads to computationally efficient and economical methods. They are also found to be more suitable for treating the sharp changes in the angular flux at duct-observer interfaces. The new approach provides a single-pass alternate to extrapolation and interactive schemes which need multiple passes of the solution strategy to acquire convergence. The method has been tested with the help of two case studies, namely straight and dog-leg duct problems. All results have been verified against those obtained from Monte Carlo simulations and K/sup +/ continuous finite element method. (author)

  3. Radiation doses by radiation diagnostics at the border of a hospital. Calculation model for Nuclear Energy Law regulations

    International Nuclear Information System (INIS)

    Shapiro, B.; Thijssen, T.; De Jong, R.

    2000-01-01

    According to the Nuclear Energy Law in the Netherlands radiation doses at the border of a specific institute (e.g. hospitals) must be determined which can not simply be done by measurements. In this article a model calculation for radiation diagnostics is described

  4. Modeling of laser radiation transport in powder beds with high-dispersive metal particles

    Energy Technology Data Exchange (ETDEWEB)

    Kharanzhevskiy, Evgeny, E-mail: eh@udsu.ru [Udmurt State University, 426034 Universitetskaya St., 1, Izhevsk (Russian Federation); Kostenkov, Sergey [Udmurt State University, 426034 Universitetskaya St., 1, Izhevsk (Russian Federation)

    2014-02-15

    Highlights: ► Transport of laser energy in dispersive powder beds was numerically simulated. ► The results of simulating are compared with physicals experiments. ► We established the dependence of the extinction coefficient from powder properties. ► A confirmation of a geometric optic approach for monodisperse powders was proposed. -- Abstract: Two-dimensional transfer of laser radiation in a high-dispersive powder heterogeneous media is numerically calculated. The size of particles is comparable with the wave length of laser radiation so the model takes into account all known physical effects that are occurred on the vacuum–metal surface interface. It is shown that in case of small particles size both morphology of powder particles and porosity of beds influence on absorptance by the solid phase and laser radiation penetrate deep into the area of geometric shadow. Intensity of laser radiation may be described as a function corresponded to the Beer–Lambert–Bouguer law.

  5. Modeling of laser radiation transport in powder beds with high-dispersive metal particles

    International Nuclear Information System (INIS)

    Kharanzhevskiy, Evgeny; Kostenkov, Sergey

    2014-01-01

    Highlights: ► Transport of laser energy in dispersive powder beds was numerically simulated. ► The results of simulating are compared with physicals experiments. ► We established the dependence of the extinction coefficient from powder properties. ► A confirmation of a geometric optic approach for monodisperse powders was proposed. -- Abstract: Two-dimensional transfer of laser radiation in a high-dispersive powder heterogeneous media is numerically calculated. The size of particles is comparable with the wave length of laser radiation so the model takes into account all known physical effects that are occurred on the vacuum–metal surface interface. It is shown that in case of small particles size both morphology of powder particles and porosity of beds influence on absorptance by the solid phase and laser radiation penetrate deep into the area of geometric shadow. Intensity of laser radiation may be described as a function corresponded to the Beer–Lambert–Bouguer law

  6. Radiation effect calculation means of the Crisis Technical Center of the Nuclear Safety and Protection Institut

    International Nuclear Information System (INIS)

    Crabol, B.; Manesse, D.; Robeau, D.

    1989-07-01

    The available calculation tools of the Crisis Technical Center (CTC), for the analysis and evaluation of radiation effects from a nuclear accident, are presented. The CTC calculation unit depends on local means, and on the National Meteorology system, in order to collect the data needed for the atmospheric waste diffusion evaluation. For the radiation dose calculations, plotters and software allowing the analysis of all waste Kinetics and all the meteorological conditions are available. The work developed by CTC calculation unit enables an easy application of the calculation tools as well as the results obtention. Images from data bases are provided to complete the obtained results [fr

  7. 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

  8. 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

  9. Monte Carlo Calculation of the Radiation Field at Aircraft Altitudes

    Energy Technology Data Exchange (ETDEWEB)

    Roesler, Stefan

    2001-08-24

    Energy spectra of secondary cosmic rays are calculated for aircraft altitudes and a discrete set of solar modulation parameters and rigidity cutoff values covering all possible conditions. The calculations are based on the Monte Carlo code FLUKA and on the most recent information on the interstellar cosmic ray flux including a detailed model of solar modulation. Results are compared to a large variety of experimental data obtained on ground and aboard of aircrafts and balloons, such as neutron, proton, and muon spectra and yields of charged particles. Furthermore, particle fluence is converted into ambient dose equivalent and effective dose and the dependence of these quantities on height above sea level, solar modulation, and geographic location is studied. Finally, calculated dose equivalent is compared to results of comprehensive measurements performed aboard of aircrafts.

  10. Calculation of radiation dose rate arisen from radionuclide contained in building materials

    International Nuclear Information System (INIS)

    Lai Tien Thinh; Nguyen Hao Quang

    2008-01-01

    This paper presents some results that we used MCNP5 program to calculate radiation dose rate arisen from radionuclide in building materials. Since then, the limits of radionuclide content in building materials are discussed. The calculation results by MCNP are compared with those calculated by analytical method. (author)

  11. 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.

  12. Calculation of radiation exposure in diagnostic radiology. Method and surveys

    International Nuclear Information System (INIS)

    Duvauferrier, R.; Ramee, A.; Ezzeldin, K.; Guibert, J.L.

    1984-01-01

    A computerized method for evaluating the radiation exposure of the main target organs during various diagnostic radiologic procedures is described. This technique was used for educational purposes: study of exposure variations according to the technical modalities of a given procedure, and study of exposure variations according to various technical protocols (IVU, EGD barium study, etc.). This method was also used for studying exposure of patients during hospitalization in the Rennes Regional Hospital Center (France) in 1982, according to departments (urology, neurology, etc.). This method and results of these three studies are discussed [fr

  13. Computer program for calculation of complex chemical equilibrium compositions and applications. Supplement 1: Transport properties

    Science.gov (United States)

    Gordon, S.; Mcbride, B.; Zeleznik, F. J.

    1984-01-01

    An addition to the computer program of NASA SP-273 is given that permits transport property calculations for the gaseous phase. Approximate mixture formulas are used to obtain viscosity and frozen thermal conductivity. Reaction thermal conductivity is obtained by the same method as in NASA TN D-7056. Transport properties for 154 gaseous species were selected for use with the program.

  14. Transport calculation of medium-energy protons and neutrons by Monte Carlo method

    International Nuclear Information System (INIS)

    Ban, Syuuichi; Hirayama, Hideo; Katoh, Kazuaki.

    1978-09-01

    A Monte Carlo transport code, ARIES, has been developed for protons and neutrons at medium energy (25 -- 500 MeV). Nuclear data provided by R.G. Alsmiller, Jr. were used for the calculation. To simulate the cascade development in the medium, each generation was represented by a single weighted particle and an average number of emitted particles was used as the weight. Neutron fluxes were stored by the collisions density method. The cutoff energy was set to 25 MeV. Neutrons below the cutoff were stored to be used as the source for the low energy neutron transport calculation upon the discrete ordinates method. Then transport calculations were performed for both low energy neutrons (thermal -- 25 MeV) and secondary gamma-rays. Energy spectra of emitted neutrons were calculated and compared with those of published experimental and calculated results. The agreement was good for the incident particles of energy between 100 and 500 MeV. (author)

  15. A model for calculating hourly global solar radiation from satellite data in the tropics

    International Nuclear Information System (INIS)

    Janjai, S.; Pankaew, P.; Laksanaboonsong, J.

    2009-01-01

    A model for calculating global solar radiation from geostationary satellite data is presented. The model is designed to calculate the monthly average hourly global radiation in the tropics with high aerosol load. This model represents a physical relation between the earth-atmospheric albedo derived from GMS5 satellite data and the absorption and scattering coefficients of various atmospheric constituents. The absorption of solar radiation by water vapour which is important for the tropics, was calculated from ambient temperature and relative humidity. The relationship between the visibility and solar radiation depletion due to aerosols was developed for a high aerosol load environment. This relationship was used to calculate solar radiation depletion by aerosols in the model. The total column ozone from TOMS/EP satellite was employed for the determination of solar radiation absorbed by ozone. Solar radiation from four pyranometer stations was used to formulate the relationship between the satellite band earth-atmospheric albedo and broadband earth-atmospheric albedo required by the model. To test its performance, the model was used to compute the monthly average hourly global radiation at 25 solar radiation monitoring stations in tropical areas in Thailand. It was found that the values of monthly average of hourly global radiations calculated from the model were in good agreement with those obtained from the measurements, with the root mean square difference of 10%. After the validation the model was employed to generate hourly solar radiation maps of Thailand. These maps reveal the diurnal and season variation of solar radiation over the country.

  16. A model for calculating hourly global solar radiation from satellite data in the tropics

    Energy Technology Data Exchange (ETDEWEB)

    Janjai, S.; Pankaew, P.; Laksanaboonsong, J. [Solar Energy Research Laboratory, Department of Physics, Faculty of Science, Silpakorn University, Nakhon Pathom 73000 (Thailand)

    2009-09-15

    A model for calculating global solar radiation from geostationary satellite data is presented. The model is designed to calculate the monthly average hourly global radiation in the tropics with high aerosol load. This model represents a physical relation between the earth-atmospheric albedo derived from GMS5 satellite data and the absorption and scattering coefficients of various atmospheric constituents. The absorption of solar radiation by water vapour which is important for the tropics, was calculated from ambient temperature and relative humidity. The relationship between the visibility and solar radiation depletion due to aerosols was developed for a high aerosol load environment. This relationship was used to calculate solar radiation depletion by aerosols in the model. The total column ozone from TOMS/EP satellite was employed for the determination of solar radiation absorbed by ozone. Solar radiation from four pyranometer stations was used to formulate the relationship between the satellite band earth-atmospheric albedo and broadband earth-atmospheric albedo required by the model. To test its performance, the model was used to compute the monthly average hourly global radiation at 25 solar radiation monitoring stations in tropical areas in Thailand. It was found that the values of monthly average of hourly global radiations calculated from the model were in good agreement with those obtained from the measurements, with the root mean square difference of 10%. After the validation the model was employed to generate hourly solar radiation maps of Thailand. These maps reveal the diurnal and season variation of solar radiation over the country. (author)

  17. Measurements and theoretical calculations of diffused radiation and atmosphere lucidity

    International Nuclear Information System (INIS)

    Pelece, I.; Iljins, U.; Ziemelis, I.

    2009-01-01

    Align with other environment friendly renewable energy sources solar energy is widely used in the world. Also in Latvia solar collectors are used. However, in Latvia because of its geographical and climatic conditions there are some specific features in comparison with traditional solar energy using countries. These features lead to the necessity to pay more attention to diffused irradiance. Another factor affecting the received irradiance of any surface is lucidity of atmosphere. This factor has not been studied in Latvia yet. This article deals with evaluation of diffused irradiance, and also of lucidity of atmosphere. The diffused irradiance can be measured directly or as a difference between the global irradiance and the beam one. The lucidity of atmosphere can be calculated from the measurements of both global and beam irradiance, if the height of the sun is known. Therefore, measurements of both global and beam irradiance have been carried out, and the diffused irradiance calculated as a difference between the global irradiance and the beam one. For measuring of the global irradiance the dome solarimeter has been used. For measuring of the direct irradiance tracking to sun pirheliometer has been used. The measurements were performed in Riga from October 2008 till March 2009. The measurements were executed automatically after every 5 minutes. The obtained results have been analyzed taking into account also the data on nebulosity from the State agency Latvian Environment, Geology and Meteorology Agency. Also efforts to calculate theoretically the diffused irradiance from the height of the sun and the data of the nebulosity have been done. These calculated values have been compared with the measured ones. Good accordance is obtained. (author)

  18. PABLM: a computer program to calculate accumulated radiation doses from radionuclides in the environment

    Energy Technology Data Exchange (ETDEWEB)

    Napier, B.A.; Kennedy, W.E. Jr.; Soldat, J.K.

    1980-03-01

    A computer program, PABLM, was written to facilitate the calculation of internal radiation doses to man from radionuclides in food products and external radiation doses from radionuclides in the environment. This report contains details of mathematical models used and calculational procedures required to run the computer program. Radiation doses from radionuclides in the environment may be calculated from deposition on the soil or plants during an atmospheric or liquid release, or from exposure to residual radionuclides in the environment after the releases have ended. Radioactive decay is considered during the release of radionuclides, after they are deposited on the plants or ground, and during holdup of food after harvest. The radiation dose models consider several exposure pathways. Doses may be calculated for either a maximum-exposed individual or for a population group. The doses calculated are accumulated doses from continuous chronic exposure. A first-year committed dose is calculated as well as an integrated dose for a selected number of years. The equations for calculating internal radiation doses are derived from those given by the International Commission on Radiological Protection (ICRP) for body burdens and MPC's of each radionuclide. The radiation doses from external exposure to contaminated water and soil are calculated using the basic assumption that the contaminated medium is large enough to be considered an infinite volume or plane relative to the range of the emitted radiations. The equations for calculations of the radiation dose from external exposure to shoreline sediments include a correction for the finite width of the contaminated beach.

  19. Advances in radiation modeling in ALEGRA :a final report for LDRD-67120, efficient implicit mulitgroup radiation calculations.

    Energy Technology Data Exchange (ETDEWEB)

    Mehlhorn, Thomas Alan; Kurecka, Christopher J. (University of Michigan, Ann Arbor, MI); McClarren, Ryan (University of Michigan, Ann Arbor, MI); Brunner, Thomas A.; Holloway, James Paul (University of Michigan, Ann Arbor, MI)

    2005-11-01

    The original LDRD proposal was to use a nonlinear diffusion solver to compute estimates for the material temperature that could then be used in a Implicit Monte Carlo (IMC) calculation. At the end of the first year of the project, it was determined that this was not going to be effective, partially due to the concept, and partially due to the fact that the radiation diffusion package was not as efficient as it could be. The second, and final year, of the project focused on improving the robustness and computational efficiency of the radiation diffusion package in ALEGRA. To this end, several new multigroup diffusion methods have been developed and implemented in ALEGRA. While these methods have been implemented, their effectiveness of reducing overall simulation run time has not been fully tested. Additionally a comprehensive suite of verification problems has been developed for the diffusion package to ensure that it has been implemented correctly. This process took considerable time, but exposed significant bugs in both the previous and new diffusion packages, the linear solve packages, and even the NEVADA Framework's parser. In order to manage this large suite of problem, a new tool called Tampa has been developed. It is a general tool for automating the process of running and analyzing many simulations. Ryan McClarren, at the University of Michigan has been developing a Spherical Harmonics capability for unstructured meshes. While still in the early phases of development, this promises to bridge the gap in accuracy between a full transport solution using IMC and the diffusion approximation.

  20. Advances in radiation modeling in ALEGRA: a final report for LDRD-67120, efficient implicit multigroup radiation calculations

    International Nuclear Information System (INIS)

    Mehlhorn, Thomas Alan; Kurecka, Christopher J.; McClarren, Ryan; Brunner, Thomas A.; Holloway, James Paul

    2005-01-01

    The original LDRD proposal was to use a nonlinear diffusion solver to compute estimates for the material temperature that could then be used in a Implicit Monte Carlo (IMC) calculation. At the end of the first year of the project, it was determined that this was not going to be effective, partially due to the concept, and partially due to the fact that the radiation diffusion package was not as efficient as it could be. The second, and final year, of the project focused on improving the robustness and computational efficiency of the radiation diffusion package in ALEGRA. To this end, several new multigroup diffusion methods have been developed and implemented in ALEGRA. While these methods have been implemented, their effectiveness of reducing overall simulation run time has not been fully tested. Additionally a comprehensive suite of verification problems has been developed for the diffusion package to ensure that it has been implemented correctly. This process took considerable time, but exposed significant bugs in both the previous and new diffusion packages, the linear solve packages, and even the NEVADA Framework's parser. In order to manage this large suite of problem, a new tool called Tampa has been developed. It is a general tool for automating the process of running and analyzing many simulations. Ryan McClarren, at the University of Michigan has been developing a Spherical Harmonics capability for unstructured meshes. While still in the early phases of development, this promises to bridge the gap in accuracy between a full transport solution using IMC and the diffusion approximation

  1. 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.

  2. Time step size selection for radiation diffusion calculations

    International Nuclear Information System (INIS)

    Rider, W.J.; Knoll, D.A.

    1999-01-01

    The purpose of this note is to describe a time step control technique as applied to radiation diffusion. Standard practice only provides a heuristic criteria related to the relative change in the dependent variables. The authors propose an alternative based on relatively simple physical principles. This time step control applies to methods of solution that are unconditionally stable and converges nonlinearities within a time step in the governing equations. Commonly, nonlinearities in the governing equations are evaluated using existing (old time) data. The authors refer to this as the semi-implicit (SI) method. When a method converges nonlinearities within a time step, the entire governing equation including all nonlinearities is self-consistently evaluated using advance time data (with appropriate time centering for accuracy)

  3. Selection of distribution coefficients for contaminant fate and transport calculations: Strontium as a case study

    International Nuclear Information System (INIS)

    Kaplan, D.I.; Krupka, K.M.; Serne, R.J.

    1997-01-01

    As part of an ongoing project funded by a cooperative effort involving the Office of Radiation and Indoor Air (ORIA) of the U.S. Environmental Protection Agency (EPA), the Office of Environmental Restoration (EM-40) of the Department of Energy (DOE), and the Nuclear Regulatory Agency (NRC), distribution coefficient (K d ) values are being compiled from the literature to develop provisional tables for cadmium, cesium, chromium, lead, plutonium, strontium, thorium, and uranium. The tables are organized according to important aqueous- and solid-phase parameters affecting the sorption of these contaminants. These parameters, which vary with contaminant, include pH and redox conditions; cation exchange capacity (CEC); presence of iron-oxide, aluminum-oxide, clay, and mica minerals; organic matter content; and solution concentrations of contaminants, competing ions, and complexing ligands. Sorption information compiled for strontium is used to illustrate our approach. The strontium data show how selected geochemical parameters (i.e., CEC, pH, and clay content) affect Strontium K d values and the selection of open-quote default close-quote K d values needed for modeling contaminant transport and risks at sites for which site specific data are lacking. Results of our evaluation may be used by site management and technical staff to assess contaminant fate, migration, and risk calculations in support of site remediation and waste management decisions

  4. Deterministic calculations of radiation doses from brachytherapy seeds

    International Nuclear Information System (INIS)

    Reis, Sergio Carneiro dos; Vasconcelos, Vanderley de; Santos, Ana Maria Matildes dos

    2009-01-01

    Brachytherapy is used for treating certain types of cancer by inserting radioactive sources into tumours. CDTN/CNEN is developing brachytherapy seeds to be used mainly in prostate cancer treatment. Dose calculations play a very significant role in the characterization of the developed seeds. The current state-of-the-art of computation dosimetry relies on Monte Carlo methods using, for instance, MCNP codes. However, deterministic calculations have some advantages, as, for example, short computer time to find solutions. This paper presents a software developed to calculate doses in a two-dimensional space surrounding the seed, using a deterministic algorithm. The analysed seeds consist of capsules similar to IMC6711 (OncoSeed), that are commercially available. The exposure rates and absorbed doses are computed using the Sievert integral and the Meisberger third order polynomial, respectively. The software also allows the isodose visualization at the surface plan. The user can choose between four different radionuclides ( 192 Ir, 198 Au, 137 Cs and 60 Co). He also have to enter as input data: the exposure rate constant; the source activity; the active length of the source; the number of segments in which the source will be divided; the total source length; the source diameter; and the actual and effective source thickness. The computed results were benchmarked against results from literature and developed software will be used to support the characterization process of the source that is being developed at CDTN. The software was implemented using Borland Delphi in Windows environment and is an alternative to Monte Carlo based codes. (author)

  5. Local transport method for hybrid diffusion-transport calculations in 2-D cylindrical (R, THETA) geometry

    International Nuclear Information System (INIS)

    Zhang, Dingkang; Rahnema, Farzad; Ougouag, Abderrfi M.

    2011-01-01

    A response-based local transport method has been developed in 2-D (r, θ) geometry for coupling to any coarse-mesh (nodal) diffusion method/code. Monte Carlo method is first used to generate a (pre-computed) the response function library for each unique coarse mesh in the transport domain (e.g., the outer reflector region of the Pebble Bed Reactor). The scalar flux and net current at the diffusion/transport interface provided by the diffusion method are used as an incoming surface source to the transport domain. A deterministic iterative sweeping method together with the response function library is utilized to compute the local transport solution within all transport coarse meshes. After the partial angular currents crossing the coarse mesh surfaces are converged, albedo coefficients are computed as boundary conditions for the diffusion methods. The iteration on the albedo boundary condition (for the diffusion method via transport) and the incoming angular flux boundary condition (for the transport via diffusion) is continued until convergence is achieved. The method was tested for in a simplified 2-D (r, θ) pebble bed reactor problem consisting of an inner reflector, an annular fuel region and a controlled outer reflector. The comparisons have shown that the results of the response-function-based transport method agree very well with a direct MCNP whole core solution. The agreement in coarse mesh averaged flux was found to be excellent: relative difference of about 0.18% and a maximum difference of about 0.55%. Note that the MCNP uncertainty was less than 0.1%. (author)

  6. 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

  7. Preliminary integrated calculation of radionuclide cation and anion transport at Yucca Mountain using a geochemical model

    International Nuclear Information System (INIS)

    Birdsell, K.H.; Campbell, K.; Eggert, K.G.; Travis, B.J.

    1989-01-01

    This paper presents preliminary transport calculations for radionuclide movement at Yucca Mountain using preliminary data for mineral distributions, retardation parameter distributions, and hypothetical recharge scenarios. These calculations are not performance assessments, but are used to study the effectiveness of the geochemical barriers at the site at mechanistic level. The preliminary calculations presented have many shortcomings and should be viewed only as a demonstration of the modeling methodology. The simulations were run with TRACRN, a finite-difference porous flow and radionuclide transport code developed for the Yucca Mountain Project. Approximately 30,000 finite-difference nodes are used to represent the unsaturated and saturated zones underlying the repository in three dimensions. Sorption ratios for the radionuclides modeled are assumed to be functions of mineralogic assemblages of the underlying rock. These transport calculations present a representative radionuclide cation, 135 Cs and anion, 99 Tc. The effects on transport of many of the processes thought to be active at Yucca Mountain may be examined using this approach. The model provides a method for examining the integration of flow scenarios, transport, and retardation processes as currently understood for the site. It will also form the basis for estimates of the sensitivity of transport calculations to retardation processes. 11 refs., 17 figs., 1 tab

  8. Iterative resonance self-shielding methods using resonance integral table in heterogeneous transport lattice calculations

    International Nuclear Information System (INIS)

    Hong, Ser Gi; Kim, Kang-Seog

    2011-01-01

    This paper describes the iteration methods using resonance integral tables to estimate the effective resonance cross sections in heterogeneous transport lattice calculations. Basically, these methods have been devised to reduce an effort to convert resonance integral table into subgroup data to be used in the physical subgroup method. Since these methods do not use subgroup data but only use resonance integral tables directly, these methods do not include an error in converting resonance integral into subgroup data. The effective resonance cross sections are estimated iteratively for each resonance nuclide through the heterogeneous fixed source calculations for the whole problem domain to obtain the background cross sections. These methods have been implemented in the transport lattice code KARMA which uses the method of characteristics (MOC) to solve the transport equation. The computational results show that these iteration methods are quite promising in the practical transport lattice calculations.

  9. 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)

  10. A new spherical harmonics scheme for multi-dimensional radiation transport I. Static matter configurations

    International Nuclear Information System (INIS)

    Radice, David; Abdikamalov, Ernazar; Rezzolla, Luciano; Ott, Christian D.

    2013-01-01

    Recent work by McClarren and Hauck (2010) [31] suggests that the filtered spherical harmonics method represents an efficient, robust, and accurate method for radiation transport, at least in the two-dimensional (2D) case. We extend their work to the three-dimensional (3D) case and find that all of the advantages of the filtering approach identified in 2D are present also in the 3D case. We reformulate the filter operation in a way that is independent of the timestep and of the spatial discretization. We also explore different second- and fourth-order filters and find that the second-order ones yield significantly better results. Overall, our findings suggest that the filtered spherical harmonics approach represents a very promising method for 3D radiation transport calculations

  11. Radiation Transport Around Axisymmetric Blunt Body Vehicles Using a Modified Differential Approximation

    Science.gov (United States)

    Hartung, Lin C.; Hassan, H. A.

    1992-01-01

    A moment method for computing 3-D radiative transport is applied to axisymmetric flows in thermochemical nonequilibrium. Such flows are representative of proposed aerobrake missions. The method uses the P-1 approximation to reduce the governing system of integro-di erential equations to a coupled set of partial di erential equations. A numerical solution method for these equations given actual variations of the radiation properties in thermochemical nonequilibrium blunt body flows is developed. Initial results from the method are shown and compared to tangent slab calculations. The agreement between the transport methods is found to be about 10 percent in the stagnation region, with the difference increasing along the flank of the vehicle.

  12. A new spherical harmonics scheme for multi-dimensional radiation transport I. Static matter configurations

    Energy Technology Data Exchange (ETDEWEB)

    Radice, David, E-mail: david.radice@aei.mpg.de [Max Planck Institute für Gravitationsphysik, Albert Einstein Institute, Potsdam (Germany); Abdikamalov, Ernazar [TAPIR, California Institute of Technology, Pasadena, CA (United States); Rezzolla, Luciano [Max Planck Institute für Gravitationsphysik, Albert Einstein Institute, Potsdam (Germany); Department of Physics and Astronomy, Louisiana State University, Baton Rouge, LA (United States); Ott, Christian D. [TAPIR, California Institute of Technology, Pasadena, CA (United States)

    2013-06-01

    Recent work by McClarren and Hauck (2010) [31] suggests that the filtered spherical harmonics method represents an efficient, robust, and accurate method for radiation transport, at least in the two-dimensional (2D) case. We extend their work to the three-dimensional (3D) case and find that all of the advantages of the filtering approach identified in 2D are present also in the 3D case. We reformulate the filter operation in a way that is independent of the timestep and of the spatial discretization. We also explore different second- and fourth-order filters and find that the second-order ones yield significantly better results. Overall, our findings suggest that the filtered spherical harmonics approach represents a very promising method for 3D radiation transport calculations.

  13. Development of a database system for the calculation of indicators of environmental pressure caused by transport

    DEFF Research Database (Denmark)

    Giannouli, Myrsini; Samaras, Zissis; Keller, Mario

    2006-01-01

    The scope of this paper is to summarise a methodology developed for TRENDS (TRansport and ENvironment Database System-TRENDS). The main objective of TRENDS was the calculation of environmental pressure indicators caused by transport. The environmental pressures considered are associated with air...... emissions from the four main transport modes, i.e. road, rail, ships and air. In order to determine these indicators a system for calculating a range of environmental pressures due to transport was developed within a PC-based MS Access environment. Emphasis is given oil the latest features incorporated...... the production of collective results for all transport modes as well as a comparative assessment of air emissions produced by the various modes. Traffic activity and emission data obtained according to a basic (reference) scenario are displayed for the time period 1970-2020. In addition, a detailed assessment...

  14. Performing three-dimensional neutral particle transport calculations on tera scale computers

    International Nuclear Information System (INIS)

    Woodward, C.S.; Brown, P.N.; Chang, B.; Dorr, M.R.; Hanebutte, U.R.

    1999-01-01

    A scalable, parallel code system to perform neutral particle transport calculations in three dimensions is presented. To utilize the hyper-cluster architecture of emerging tera scale computers, the parallel code successfully combines the MPI message passing and paradigms. The code's capabilities are demonstrated by a shielding calculation containing over 14 billion unknowns. This calculation was accomplished on the IBM SP ''ASCI-Blue-Pacific computer located at Lawrence Livermore National Laboratory (LLNL)

  15. Measurement and calculation of radiation fields of the Sandia irradiator for dried sewage solids

    International Nuclear Information System (INIS)

    Morris, M.E.

    1981-03-01

    The radiation field of the Sandia Irradiator for Dried Sewage Solids was measured. The results of the measurement are given in this report. In addition, theoretical calculations of the fields are given and then compared with the measured values. Elementary models of the radiation source geometry and irradiated product are found to be adequate and thus allow us to duplicate (through calculation) the important features of the measured fields

  16. 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

  17. Biomedical applications of two- and three-dimensional deterministic radiation transport methods

    International Nuclear Information System (INIS)

    Nigg, D.W.

    1992-01-01

    Multidimensional deterministic radiation transport methods are routinely used in support of the Boron Neutron Capture Therapy (BNCT) Program at the Idaho National Engineering Laboratory (INEL). Typical applications of two-dimensional discrete-ordinates methods include neutron filter design, as well as phantom dosimetry. The epithermal-neutron filter for BNCT that is currently available at the Brookhaven Medical Research Reactor (BMRR) was designed using such methods. Good agreement between calculated and measured neutron fluxes was observed for this filter. Three-dimensional discrete-ordinates calculations are used routinely for dose-distribution calculations in three-dimensional phantoms placed in the BMRR beam, as well as for treatment planning verification for live canine subjects. Again, good agreement between calculated and measured neutron fluxes and dose levels is obtained

  18. 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...

  19. Computational complexity in multidimensional neutron transport theory calculations. Progress report, September 1, 1975--August 31, 1976

    International Nuclear Information System (INIS)

    Bareiss, E.H.

    1976-05-01

    The objectives of the work are to develop mathematically and computationally founded for the design of highly efficient and reliable multidimensional neutron transport codes to solve a variety of neutron migration and radiation problems, and to analyze existing and new methods for performance. As new analytical insights are gained, new numerical methods are developed and tested. Significant results obtained include implementation of the integer-preserving Gaussian elimination method (two-step method) in a CDC 6400 computer code, modes analysis for one-dimensional transport solutions, and a new method for solving the 1-T transport equation. Some of the work dealt with the interface and corner problem in diffusion theory

  20. Effects of radiation transport on mass ablation rate and conversion efficiency in numerical simulations of inertial confinement fusion

    International Nuclear Information System (INIS)

    Gupta, N.K.

    2002-01-01

    The effects of radiation transport on hydrodynamic parameters of laser produced plasmas are studied. LTE and non-LTE atomic models are used to calculate multi group opacities and emissivities. Screened hydrogenic atom model is used to calculate the energy levels. The population densities of neutral to fully ionized ions are obtained by solving the steady state rate equations. Radiation transport is treated in multi-group diffusion or Sn method. A comparison is made between 1 and 100 group radiation transport and LTE and non-LTE models. For aluminium, multi group radiation transport leads to much higher mass ablation as compared to the 1 group and no radiation transport cases. This in turn leads to higher ablation pressures. However, for gold gray approximation gives higher mass ablation as compared to multi group simulations. LTE conversion efficiency of laser light into x-rays is more than the non-LTE estimates. For LTE as well as non-LTE cases, the one group approximation over-predicts the conversion efficiency Multi group non-LTE simulations predict that the conversion efficiency increases with laser intensity up to a maximum and then it decreases. (author)

  1. Radiation transport analyses in support of the SNS Target Station Neutron Beam Line Shutters Title I Design

    International Nuclear Information System (INIS)

    Miller, T.M.; Pevey, R.E.; Lillie, R.A.; Johnson, J.O.

    2000-01-01

    A detailed radiation transport analysis of the Spallation Neutron Source (SNS) shutters is important for the construction of the SNS because of its impact on conventional facility design, normal operation of the facility, and maintenance operations. Thus far the analysis of the SNS shutter travel gaps has been completed. This analysis was performed using coupled Monte Carlo and multi-dimensional discrete ordinates calculations

  2. 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)

  3. 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

  4. Comparison and application study on cosmic radiation dose calculation received by air crew

    International Nuclear Information System (INIS)

    Zhou Qiang; Xu Cuihua; Ren Tianshan; Li Wenhong; Zhang Jing; Lu Xu

    2009-01-01

    Objective: To facilitate evaluation on Cosmic radiation dose received by flight crew by developing a convenient and effective measuring method. Methods: In comparison with several commonly used evaluating methods, this research employs CARI-6 software issued by FAA (Federal Aviation Administration) to measure Cosmic radiation dose for flight crew members exposed to. Results: Compared with other methods, CARI-6 is capable of providing reliable calculating results on radiation dose and applicable to all flight crew of different airlines. Conclusion: Cosmic radiation received by flight crew is on the list of occupational radiation. For a smooth running of Standards for controlling exposure to cosmic radiation of air crew, CARI software may be a widely applied tool in radiation close estimation of for flight crew. (authors)

  5. Radiation by the numbers: developing an on-line Canadian radiation dose calculator as a public engagement and education tool

    Energy Technology Data Exchange (ETDEWEB)

    Dalzell, M.T.J. [Sylvia Fedoruk Canadian Centre for Nuclear Innovation, Saskatoon, Saskatchewan (Canada)

    2016-06-15

    Concerns arising from misunderstandings about radiation are often cited as a main reason for public antipathy towards nuclear development and impede decision-making by governments and individuals. A lack of information about everyday sources of radiation exposure that is accessible, relatable and factual contributes to the problem. As part of its efforts to be a fact-based source of information on nuclear issues, the Sylvia Fedoruk Canadian Centre for Nuclear Innovation has developed an on-line Canadian Radiation Dose Calculator as a tool to provide context about common sources of radiation. This paper discusses the development of the calculator and describes how the Fedoruk Centre is using it and other tools to support public engagement on nuclear topics. (author)

  6. Computational complexity in multidimensional neutron transport theory calculations. Progress report, September 1, 1974--August 31, 1975

    International Nuclear Information System (INIS)

    Bareiss, E.H.

    1975-01-01

    The objectives of the research remain the same as outlined in the original proposal. They are in short as follows: Develop mathematically and computationally founded criteria for the design of highly efficient and reliable multi-dimensional neutron transport codes to solve a variety of neutron migration and radiation problems and analyze existing and new methods for performance. (U.S.)

  7. Computational complexity in multidimensional neutron transport theory calculations. Progress report, September 1976--November 30, 1977

    International Nuclear Information System (INIS)

    Bareiss, E.H.

    1977-08-01

    The objectives of this research are to develop mathematically and computationally founded criteria for the design of highly efficient and reliable multidimensional neutron transport codes to solve a variety of neutron migration and radiation problems, and to analyze existing and new methods for performance

  8. 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)

  9. RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

    International Nuclear Information System (INIS)

    Yuan, Y.C.; Chen, S.Y.; LePoire, D.J.

    1993-02-01

    This report presents the technical details of RISIUND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, semiinteractive program that can be run on an IBM or equivalent personal computer. The program language is FORTRAN-77. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incidentfree models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionudide inventory and dose conversion factors

  10. RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, Y.C. [Square Y, Orchard Park, NY (United States); Chen, S.Y.; LePoire, D.J. [Argonne National Lab., IL (United States). Environmental Assessment and Information Sciences Div.; Rothman, R. [USDOE Idaho Field Office, Idaho Falls, ID (United States)

    1993-02-01

    This report presents the technical details of RISIUND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, semiinteractive program that can be run on an IBM or equivalent personal computer. The program language is FORTRAN-77. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incidentfree models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionudide inventory and dose conversion factors.

  11. Guideline of Monte Carlo calculation. Neutron/gamma ray transport simulation by Monte Carlo method

    CERN Document Server

    2002-01-01

    This report condenses basic theories and advanced applications of neutron/gamma ray transport calculations in many fields of nuclear energy research. Chapters 1 through 5 treat historical progress of Monte Carlo methods, general issues of variance reduction technique, cross section libraries used in continuous energy Monte Carlo codes. In chapter 6, the following issues are discussed: fusion benchmark experiments, design of ITER, experiment analyses of fast critical assembly, core analyses of JMTR, simulation of pulsed neutron experiment, core analyses of HTTR, duct streaming calculations, bulk shielding calculations, neutron/gamma ray transport calculations of the Hiroshima atomic bomb. Chapters 8 and 9 treat function enhancements of MCNP and MVP codes, and a parallel processing of Monte Carlo calculation, respectively. An important references are attached at the end of this report.

  12. Development of comprehensive models for opacities and radiation transport for IFE systems

    International Nuclear Information System (INIS)

    Tolkach, V.; Morozov, V.; Hassanein, A.

    2003-01-01

    An ignition in an inertial confinement fusion (ICF) reactor results in X-ray spectra and ion fluxes moving toward the chamber wall with different velocities. During flight, parts of the energy will be deposited either in the residual and/or protective chamber gas or in the initial vapor cloud developed near the wall surface from vaporization. The deposited energy will be re-radiated to the chamber wall long after the ignition process. The exact amount of energy deposited/radiated and time of deposition are key issues in evaluating the chamber response and the economical feasibility of an ICF reactor. The radiation processes in the protective gas layer or in the vapor cloud developed above the first wall play an important role in the overall dynamics of the ICF chamber. A self-consistent field method has been developed to calculate ionization potentials, atom and ion energy levels, transition probabilities, and other atomic properties used to calculate thermodynamic and optical characteristics of the plasma by means of collisional-radiation equilibrium (CRE). The methodology of solving radiation transport equations in spherical geometry and the dependence of results on the chosen theoretical model are demonstrated using the method of inward/outward directions

  13. A NEW MONTE CARLO METHOD FOR TIME-DEPENDENT NEUTRINO RADIATION TRANSPORT

    International Nuclear Information System (INIS)

    Abdikamalov, Ernazar; Ott, Christian D.; O'Connor, Evan; Burrows, Adam; Dolence, Joshua C.; Löffler, Frank; Schnetter, Erik

    2012-01-01

    Monte Carlo approaches to radiation transport have several attractive properties such as simplicity of implementation, high accuracy, and good parallel scaling. Moreover, Monte Carlo methods can handle complicated geometries and are relatively easy to extend to multiple spatial dimensions, which makes them potentially interesting in modeling complex multi-dimensional astrophysical phenomena such as core-collapse supernovae. The aim of this paper is to explore Monte Carlo methods for modeling neutrino transport in core-collapse supernovae. We generalize the Implicit Monte Carlo photon transport scheme of Fleck and Cummings and gray discrete-diffusion scheme of Densmore et al. to energy-, time-, and velocity-dependent neutrino transport. Using our 1D spherically-symmetric implementation, we show that, similar to the photon transport case, the implicit scheme enables significantly larger timesteps compared with explicit time discretization, without sacrificing accuracy, while the discrete-diffusion method leads to significant speed-ups at high optical depth. Our results suggest that a combination of spectral, velocity-dependent, Implicit Monte Carlo and discrete-diffusion Monte Carlo methods represents a robust approach for use in neutrino transport calculations in core-collapse supernovae. Our velocity-dependent scheme can easily be adapted to photon transport.

  14. A NEW MONTE CARLO METHOD FOR TIME-DEPENDENT NEUTRINO RADIATION TRANSPORT

    Energy Technology Data Exchange (ETDEWEB)

    Abdikamalov, Ernazar; Ott, Christian D.; O' Connor, Evan [TAPIR, California Institute of Technology, MC 350-17, 1200 E California Blvd., Pasadena, CA 91125 (United States); Burrows, Adam; Dolence, Joshua C. [Department of Astrophysical Sciences, Princeton University, Peyton Hall, Ivy Lane, Princeton, NJ 08544 (United States); Loeffler, Frank; Schnetter, Erik, E-mail: abdik@tapir.caltech.edu [Center for Computation and Technology, Louisiana State University, 216 Johnston Hall, Baton Rouge, LA 70803 (United States)

    2012-08-20

    Monte Carlo approaches to radiation transport have several attractive properties such as simplicity of implementation, high accuracy, and good parallel scaling. Moreover, Monte Carlo methods can handle complicated geometries and are relatively easy to extend to multiple spatial dimensions, which makes them potentially interesting in modeling complex multi-dimensional astrophysical phenomena such as core-collapse supernovae. The aim of this paper is to explore Monte Carlo methods for modeling neutrino transport in core-collapse supernovae. We generalize the Implicit Monte Carlo photon transport scheme of Fleck and Cummings and gray discrete-diffusion scheme of Densmore et al. to energy-, time-, and velocity-dependent neutrino transport. Using our 1D spherically-symmetric implementation, we show that, similar to the photon transport case, the implicit scheme enables significantly larger timesteps compared with explicit time discretization, without sacrificing accuracy, while the discrete-diffusion method leads to significant speed-ups at high optical depth. Our results suggest that a combination of spectral, velocity-dependent, Implicit Monte Carlo and discrete-diffusion Monte Carlo methods represents a robust approach for use in neutrino transport calculations in core-collapse supernovae. Our velocity-dependent scheme can easily be adapted to photon transport.

  15. Radiation Re-solution Calculation in Uranium-Silicide Fuels

    International Nuclear Information System (INIS)

    Matthews, Christopher; Andersson, Anders David Ragnar; Unal, Cetin

    2017-01-01

    The release of fission gas from nuclear fuels is of primary concern for safe operation of nuclear power plants. Although the production of fission gas atoms can be easily calculated from the fission rate in the fuel and the average yield of fission gas, the actual diffusion, behavior, and ultimate escape of fission gas from nuclear fuel depends on many other variables. As fission gas diffuses through the fuel grain, it tends to collect into intra-granular bubbles, as portrayed in Figure 1.1. These bubbles continue to grow due to absorption of single gas atoms. Simultaneously, passing fission fragments can cause collisions in the bubble that result in gas atoms being knocked back into the grain. This so called ''re-solution'' event results in a transient equilibrium of single gas atoms within the grain. As single gas atoms progress through the grain, they will eventually collect along grain boundaries, creating inter-granular bubbles. As the inter-granular bubbles grow over time, they will interconnect with other grain-face bubbles until a pathway is created to the outside of the fuel surface, at which point the highly pressurized inter-granular bubbles will expel their contents into the fuel plenum. This last process is the primary cause of fission gas release. From the simple description above, it is clear there are several parameters that ultimately affect fission gas release, including the diffusivity of single gas atoms, the absorption and knockout rate of single gas atoms in intra-granular bubbles, and the growth and interlinkage of intergranular bubbles. Of these, the knockout, or re-solution rate has an particularly important role in determining the transient concentration of single gas atoms in the grain. The re-solution rate will be explored in the following sections with regards to uranium-silicide fuels in order to support future models of fission gas bubble behavior.

  16. Program system for calculating streaming neutron radiation field in reactor cavity

    International Nuclear Information System (INIS)

    He Zhongliang; Zhao Shu.

    1986-01-01

    The A23 neutron albedo data base based on Monte Carlo method well agrees with SAIL albedo data base. RSCAM program system, using Monte Carlo method with albedo approach, is used to calculate streaming neutron radiation field in reactor cavity and containment operating hall. The dose rate distributions calculated with RSCAM in square concrete duct well agree with experiments

  17. Calculation of the radiation force on a cylinder in a standing wave acoustic field

    Energy Technology Data Exchange (ETDEWEB)

    Haydock, David [Unilever R and D Colworth, Sharnbrook, Bedford MK44 1LQ (United Kingdom); Department of Physics, Theoretical Physics, University of Oxford, 1 Keble Road, Oxford OX1 3NP (United Kingdom)

    2005-04-15

    We present a new calculation of the radiation force on a cylinder in a standing wave acoustic field. We use the formula to calculate the force on a cylinder which is free to move in the field and one which is fixed in space.

  18. Calculation of the radiation force on a cylinder in a standing wave acoustic field

    International Nuclear Information System (INIS)

    Haydock, David

    2005-01-01

    We present a new calculation of the radiation force on a cylinder in a standing wave acoustic field. We use the formula to calculate the force on a cylinder which is free to move in the field and one which is fixed in space

  19. Department of Environmental and Radiation Transport Physics - Overview

    International Nuclear Information System (INIS)

    Woznicka, U.

    2002-01-01

    geological structures, in collaboration with the Institute of Geological Sciences of Polish Academy of Sciences. The geological fault system which surrounds the ''Las Wolski'' horst is covered with loess overburden. An evident increase in radon concentration in the upper loess layer is observed over the fault position. This may have important environmental implications. Several samples of soil taken from those areas were analysed for the concentration of natural isotopes (U, Th, and K). Natural radioactivity measurements in various samples (soils, rocks, raw, and building materials, etc.) have been carried out using low background spectrometers (with NaI(Tl) and HPGe detectors). We took part in the national inter-comparison concerning the methodology of ''radon-in-water'' measurements. The results are to be published. A joint project ''The Radon Centre - Non- Governmental International Scientific Network'' has been started in co-operation with the Central Mining Institute in Katowice. The main goals are to prepare and execute joint research projects and programmes, and to disseminate and put into practice the results of research activities of particular Network members. Neutron methods are an important part of nuclear geophysics and are also used in medical modalities. Investigations of the neutron transport parameters require usually the detection and/or calculation of spatial, time, and energy distributions of fast, epithermal and thermal neutrons, and of the accompanying γ radiation. The research has been directed into several aspects: - Basic theoretical and experimental investigation for the thermal neutron transport: a) the temperature behaviour of the pulsed parameters in a hydrogenous moderator, b) diffusion cooling in small two-region systems containing substances of different types of energy characteristics of thermal neutron scattering. - Calculations of the radiation field and energy deposition in the water beam dump for the TESLA electron-positron collider for

  20. The Eddington approximation calculation of radiation flux in the atmosphere–ocean system

    International Nuclear Information System (INIS)

    Shi, Chong; Nakajima, Teruyuki

    2015-01-01

    An analytical approximation method is presented to calculate the radiation flux in the atmosphere–ocean system using the Eddington approximation when the upwelling radiation from the ocean body is negligibly small. Numerical experiments were carried out to investigate the feasibility of the method in two cases: flat and rough ocean surfaces. The results show good consistency for the reflectivity at the top of atmosphere and transmissivity just above the ocean surface, in comparison with the exact values calculated by radiative transfer models in each case. Moreover, an obvious error might be introduced for the calculation of radiation flux at larger solar zenith angles when the roughness of the ocean surface is neglected. - Highlights: • The Eddington approximation method is extended to the atmosphere–ocean system. • The roughness of ocean surface cannot be neglected at lager solar zenith angles. • Unidirectional reflectivity for rough ocean surface is proposed

  1. The simple analytical method for scattered radiation calculation in contrast X-ray diagnostic techniques

    Energy Technology Data Exchange (ETDEWEB)

    Markovic, S; Pavlovic, R [Inst. of Nuclear Science Vinca, Belgrade (Yugoslavia). Radiation and Environmental Protection Lab.; Boreli, F [Fac. of Electrical Engineering, Belgrade (Yugoslavia)

    1996-12-31

    In realization of radiation protection measures for medical staff present during diagnostic procedures, the necessary condition is knowledge of the space - energy distributions of the scattered radiation from the patient. In this paper, the simple calculation procedure for the scattered radiation field of the actual diagnostic energies is presented. Starting from the single Compton scattering model and using the justified transformations the final equations in elementary form are derived. For numerical calculations the computer code ANGIO was created. The calculated results were confirmed by detailed dosimetric measurements of the scattered field around patient (the water phantom) in SSDL in the Institute of nuclear sciences `Vinca`, Belgrade. These results are good base for assessment of irradiation. The main irradiation source for the physician and the other members of the medical team is the back scattered radiation from patient - albedo. (author). 3 figs., 3 refs.

  2. Deterministic and stochastic methods of calculation of polarization characteristics of radiation in natural environment

    Science.gov (United States)

    Strelkov, S. A.; Sushkevich, T. A.; Maksakova, S. V.

    2017-11-01

    We are talking about russian achievements of the world level in the theory of radiation transfer, taking into account its polarization in natural media and the current scientific potential developing in Russia, which adequately provides the methodological basis for theoretically-calculated research of radiation processes and radiation fields in natural media using supercomputers and mass parallelism. A new version of the matrix transfer operator is proposed for solving problems of polarized radiation transfer in heterogeneous media by the method of influence functions, when deterministic and stochastic methods can be combined.

  3. 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.

  4. 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

  5. Structural instability of atmospheric flows under perturbations of the mass balance and effect in transport calculations

    International Nuclear Information System (INIS)

    Núñez, M A; Mendoza, R

    2015-01-01

    Several methods to estimate the velocity field of atmospheric flows, have been proposed to the date for applications such as emergency response systems, transport calculations and for budget studies of all kinds. These applications require a wind field that satisfies the conservation of mass but, in general, estimated wind fields do not satisfy exactly the continuity equation. An approach to reduce the effect of using a divergent wind field as input in the transport-diffusion equations, was proposed in the literature. In this work, a linear local analysis of a wind field, is used to show analytically that the perturbation of a large-scale nondivergent flow can yield a divergent flow with a substantially different structure. The effects of these structural changes in transport calculations are illustrated by means of analytic solutions of the transport equation

  6. CLEAR (Calculates Logical Evacuation And Response): A Generic Transportation Network Model for the Calculation of Evacuation Time Estimates

    Energy Technology Data Exchange (ETDEWEB)

    Moeller, M. P.; Urbanik, II, T.; Desrosiers, A. E.

    1982-03-01

    This paper describes the methodology and application of the computer model CLEAR (Calculates Logical Evacuation And Response) which estimates the time required for a specific population density and distribution to evacuate an area using a specific transportation network. The CLEAR model simulates vehicle departure and movement on a transportation network according to the conditions and consequences of traffic flow. These include handling vehicles at intersecting road segments, calculating the velocity of travel on a road segment as a function of its vehicle density, and accounting for the delay of vehicles in traffic queues. The program also models the distribution of times required by individuals to prepare for an evacuation. In order to test its accuracy, the CLEAR model was used to estimate evacuatlon tlmes for the emergency planning zone surrounding the Beaver Valley Nuclear Power Plant. The Beaver Valley site was selected because evacuation time estimates had previously been prepared by the licensee, Duquesne Light, as well as by the Federal Emergency Management Agency and the Pennsylvania Emergency Management Agency. A lack of documentation prevented a detailed comparison of the estimates based on the CLEAR model and those obtained by Duquesne Light. However, the CLEAR model results compared favorably with the estimates prepared by the other two agencies.

  7. CLEAR (Calculates Logical Evacuation And Response): A generic transportation network model for the calculation of evacuation time estimates

    International Nuclear Information System (INIS)

    Moeller, M.P.; Desrosiers, A.E.; Urbanik, T. II

    1982-03-01

    This paper describes the methodology and application of the computer model CLEAR (Calculates Logical Evacuation And Response) which estimates the time required for a specific population density and distribution to evacuate an area using a specific transportation network. The CLEAR model simulates vehicle departure and movement on a transportation network according to the conditions and consequences of traffic flow. These include handling vehicles at intersecting road segments, calculating the velocity of travel on a road segment as a function of its vehicle density, and accounting for the delay of vehicles in traffic queues. The program also models the distribution of times required by individuals to prepare for an evacuation. In order to test its accuracy, the CLEAR model was used to estimate evacuation times for the emergency planning zone surrounding the Beaver Valley Nuclear Power Plant. The Beaver Valley site was selected because evacuation time estimates had previously been prepared by the licensee, Duquesne Light, as well as by the Federal Emergency Management Agency and the Pennsylvania Emergency Management Agency. A lack of documentation prevented a detailed comparison of the estimates based on the CLEAR model and those obtained by Duquesne Light. However, the CLEAR model results compared favorably with the estimates prepared by the other two agencies. (author)

  8. Calculation analysis of the thickness of radiation shield for the RIA equipment IP10

    International Nuclear Information System (INIS)

    Benar Bukit; Kristiyanti; Hari Nurcahyadi

    2011-01-01

    Calculation Analysis has been performed on the thickness of radiation shield for the design of the Radioimmunoassay (RIA) IP10 counters using five detectors arranged in parallel. The calculation is intended to ensure that the radiation on each detector does not influence each other. The radiation shield is made of lead. The calculation of lead thickness was based on the principle of the lead plates absorptive power toward the gamma ray of a certain energy. which is the function of linear absorption coefficient and the material thickness. Assuming the use of Iodium-125(I-125) source with an activity 10 µCi, and expecting an absorptive power of 95%, calculations showed that the required lead thickness is equal to 0,013 cm. Since lead is soft and its availability in the market is limited, lead plate of 2 mm thickness are used instead, so that counting result for the detectors do not influence each other. (author)

  9. ALMA. Calculations of diffusion and radiation doses in connection with final storage of radioactive wastes

    International Nuclear Information System (INIS)

    Gelin, R.; Kjellbert, N.; Stenquist, C.

    1978-09-01

    Calculations of diffusion and radiation doses in connection with final storage of low-lavel and intermediate-level radioactive wastes. The results show that the doses obtained with realistic values of parameters used in the calculations are very low. However, substantially simplified assumption have been applied in the calculations. Thus more detailed models for the description of the diffusion process have to be developed. (E.R.)

  10. Radiation doses due to long-range transport of airborne radionuclides

    International Nuclear Information System (INIS)

    Nordlund, G.; Valkama, I.; Rossi, J.; Savolainen, I.

    1985-12-01

    Within the framework of this study a model for estimating the long range transport of radioactive material and for calculating the resultant doses is developed. In the model initially the dispersion paths, i.e. trajectories, of the radioactive matter are calculated from the assumed source areas as well as the dispersion conditions along the trajectories. The trajectories are calculated at three-hour intervals in a two-dimensional grid using numerically analysed winds at a constant pressure level of 850 mb. The dispersion condition parameters applied are: the stability of the atmospheric boundary layer, the so-called mixing height, occurrence of precipitation and the character of the terrain. For each trajectory a type-index value is computed, describing the severity of the possible effects of radioactivity transported by the particular trajectory. The dispersion model uses the information on dispersion conditions provided by the trajectory model to compute the remaining radioactivity in the cloud, the deposition, as well as the doses due to different dose pathways. The pathways used are the external radiation from the cloud and from the activity deposited on the ground, inhalation of radioactive material and ingestion of contaminated food products (milk, meat, green vegetables, grain and roots). In addition to the effects of individual transport incidents, the cumulative probability distributions of the effects of accidental releases of radioactive matter can also be calculated using trajectory statistics and the trajectory type index

  11. 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.

  12. 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

  13. 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.

  14. 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

  15. Radiation transport methods for nuclear log assessment - an overview

    International Nuclear Information System (INIS)

    Badruzzaman, A.

    1996-01-01

    Methods of radiation transport have been applied to well-logging problems with nuclear sources since the early 1960s. Nuclear sondes are used in identifying rock compositions and fluid properties in reservoirs to predict the porosity and oil saturation. Early computational effort in nuclear logging used diffusion techniques. As computers became more powerful, deterministic transport methods and, finally, Monte Carlo methods were applied to solve these problems in three dimensions. Recently, the application has been extended to problems with a new generation of devices, including spectroscopic sondes that measure such quantities as the carbon/oxygen ratio to predict oil saturation and logging-while-drilling (LWD) sondes that take neutron and gamma measurements as they rotate in the borehole. These measurements present conditions that will be difficult to calibrate in the laboratory

  16. 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)

  17. THE CALCULATION OF THE ENERGY RECOVERY ELECTRIFIED URBAN TRANSPORT DURING THE INSTALLATION DRIVE FOR TRACTION SUBSTATION

    Directory of Open Access Journals (Sweden)

    A. A. Sulim

    2014-01-01

    Full Text Available At present a great attention is paid to increasing of energy efficiency at operated electrified urban transport. Perspective direction for increasing energy efficiency at that type of transport is the application of regenerative braking. For additional increasing of energy efficiency there were suggested the use of capacitive drive on tires of traction substation. One of the main task is the analysis of energy recovery application  with drive and without it.These analysis demonstrated that the calculation algorithms don’t allow in the full volume to carry out calculations of amount and cost of energy recovery without drive and with it. That is why we see the current interest to this topic. The purpose of work is to create methods of algorithms calculation for definite amount and cost of consumed, redundant and recovery energy of electrified urban transport due to definite regime of motion on wayside. There is algorithm developed, which allow to calculate amount and cost of consumed, redundant and recovery energy of electrified urban transport on wayside during the installation capacitive drive at traction substation. On the basis of developed algorithm for the definite regime of wagon motion of subway there were fulfilled the example of energy recovery amount and its cost calculation, among them with limited energy intensity drive, when there are 4 trains on wayside simultaneously.

  18. Improved method for calculating neoclassical transport coefficients in the banana regime

    Energy Technology Data Exchange (ETDEWEB)

    Taguchi, M., E-mail: taguchi.masayoshi@nihon-u.ac.jp [College of Industrial Technology, Nihon University, Narashino 275-8576 (Japan)

    2014-05-15

    The conventional neoclassical moment method in the banana regime is improved by increasing the accuracy of approximation to the linearized Fokker-Planck collision operator. This improved method is formulated for a multiple ion plasma in general tokamak equilibria. The explicit computation in a model magnetic field shows that the neoclassical transport coefficients can be accurately calculated in the full range of aspect ratio by the improved method. The some neoclassical transport coefficients for the intermediate aspect ratio are found to appreciably deviate from those obtained by the conventional moment method. The differences between the transport coefficients with these two methods are up to about 20%.

  19. AUTOMATION OF CALCULATION ALGORITHMS FOR EFFICIENCY ESTIMATION OF TRANSPORT INFRASTRUCTURE DEVELOPMENT

    Directory of Open Access Journals (Sweden)

    Sergey Kharitonov

    2015-06-01

    Full Text Available Optimum transport infrastructure usage is an important aspect of the development of the national economy of the Russian Federation. Thus, development of instruments for assessing the efficiency of infrastructure is impossible without constant monitoring of a number of significant indicators. This work is devoted to the selection of indicators and the method of their calculation in relation to the transport subsystem as airport infrastructure. The work also reflects aspects of the evaluation of the possibilities of algorithmic computational mechanisms to improve the tools of public administration transport subsystems.

  20. Monte Carlo calculation of scattered radiation from applicators in low energy clinical electron beams

    International Nuclear Information System (INIS)

    Jabbari, N.; Hashemi-Malayeri, B.; Farajollahi, A. R.; Kazemnejad, A.

    2007-01-01

    In radiotherapy with electron beams, scattered radiation from an electron applicator influences the dose distribution in the patient. The contribution of this radiation to the patient dose is significant, even in modern accelerators. In most of radiotherapy treatment planning systems, this component is not explicitly included. In addition, the scattered radiation produced by applicators varies based on the applicator design as well as the field size and distance from the applicators. The aim of this study was to calculate the amount of scattered dose contribution from applicators. We also tried to provide an extensive set of calculated data that could be used as input or benchmark data for advanced treatment planning systems that use Monte Carlo algorithms for dose distribution calculations. Electron beams produced by a NEPTUN 10PC medical linac were modeled using the BEAMnrc system. Central axis depth dose curves of the electron beams were measured and calculated, with and without the applicators in place, for different field sizes and energies. The scattered radiation from the applicators was determined by subtracting the central axis depth dose curves obtained without the applicators from that with the applicator. The results of this study indicated that the scattered radiation from the electron applicators of the NEPTUN 10PC is significant and cannot be neglected in advanced treatment planning systems. Furthermore, our results showed that the scattered radiation depends on the field size and decreases almost linearly with depth. (author)

  1. Development of a database system for the calculation of indicators of environmental pressure caused by transport

    Energy Technology Data Exchange (ETDEWEB)

    Giannouli, Myrsini; Samaras, Zissis [Aristotle University of Thessaloniki, Laboratory of Applied Thermodynamics, Mechanical Engineering Department, GR 54124, Thessaloniki, P.O. Box 458 (Greece); Keller, Mario; De Haan, Peter [INFRAS, Muhlemattstrasse 45 CH-3007, Bern (Switzerland); Kallivoda, Manfred [psiA-Consult, Environmental Research and Engineering GmbH, Lastenstrasse 38/1, 1230 Wien (Austria); Sorenson, Spencer; Georgakaki, Aliki [DTU: Technical University of Denmark, Nils Koppels Alle, Building 403, DK 2800 Kgs. Lyngby (Denmark)

    2006-03-15

    The scope of this paper is to summarise a methodology developed for TRENDS (TRansport and ENvironment Database System-TRENDS). The main objective of TRENDS was the calculation of environmental pressure indicators caused by transport. The environmental pressures considered are associated with air emissions from the four main transport modes, i.e. road, rail, ships and air. In order to determine these indicators a system for calculating a range of environmental pressures due to transport was developed within a PC-based MS Access environment. Emphasis is given on the latest features incorporated in the model and their applications. One of the recently developed features of the software provides an option for simple scenario analysis including vehicle dynamics (such as turnover and evolution) for all EU15 member states. This feature is called the Transport Activity Balance module (TAB) and enables the production of collective results for all transport modes as well as a comparative assessment of air emissions produced by the various modes. Traffic activity and emission data obtained according to a basic (reference) scenario are displayed for the time period 1970-2020. In addition, a detailed assessment of the results produced by TRENDS was conducted by means of comparison with data found in the literature. Finally, vehicle emissions produced by the model for the EU15 member states were spatially disaggregated for the base year, 1995 and GIS maps were generated. Examples of these maps are displayed in this document, for the various modes of transport considered in the study. (author)

  2. Radiative transport and collisional transfer of excitation energy in Cs vapors mixed with Ar or He

    International Nuclear Information System (INIS)

    Vadla, Cedomil; Horvatic, Vlasta; Niemax, Kay

    2003-01-01

    This paper is a review (with a few original additions) on the radiative transport and collisional transfer of energy in laser-excited cesium vapors in the presence of argon or helium. Narrow-band excitation of lines with Lorentz, Doppler and Voigt profiles is studied in order to calculate effective rates for pumping of spectral lines with profiles comprising inhomogeneous broadening components. The radiative transport of excitation energy is considered, and a new, simple and robust, but accurate theoretical method for quantitative treatment of radiation trapping in relatively optically thin media is presented. Furthermore, comprehensive lists of experimental values for the excitation energy transfer cross-sections related to thermal collisions in Cs-Ar and Cs-He mixtures are given. Within the collected cross-section data sets, specific regularities with respect to the energy defect, as well as the temperature, are discerned. A particular emphasis is put on the radiative and collisional processes important for the optimization of resonance-fluorescence imaging atomic filters based on Cs-noble gas systems

  3. Discussion of and guidance on the optimization of radiation protection in the transport of radioactive material

    International Nuclear Information System (INIS)

    1986-05-01

    The document provides guidance on one of the components of the system of dose limitation as it applies to the transport of radioactive material, namely the optimization of radiation protection. It focuses on the following parts of the transport system: design, maintenance, preparation for transport, transport, storage-in-transit and handling and it considers occupational and public exposures. The application is intended mainly for those transport situations within the regulatory requirements where potential radiation exposures could be beneficially reduced

  4. Radiation protection programmes for the transport of radioactive material. Safety guide

    International Nuclear Information System (INIS)

    2007-01-01

    This Safety Guide provides guidance on meeting the requirements for the establishment of radiation protection programmes (RPPs) for the transport of radioactive material, to optimize radiation protection in order to meet the requirements for radiation protection that underlie the Regulations for the Safe Transport of Radioactive Material. This Guide covers general aspects of meeting the requirements for radiation protection, but does not cover criticality safety or other possible hazardous properties of radioactive material. The annexes of this Guide include examples of RPPs, relevant excerpts from the Transport Regulations, examples of total dose per transport index handled, a checklist for road transport, specific segregation distances and emergency instructions for vehicle operators

  5. 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

  6. 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

  7. 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.

  8. Line photon transport in a non-homogeneous plasma using radiative coupling coefficients

    International Nuclear Information System (INIS)

    Florido, R.; Gil, J.M.; Rodriguez, R.; Rubiano, J.G.; Martel, P.; Florido, R.; Gil, J.M.; Rodriguez, R.; Rubiano, J.G.; Martel, P.; Minguez, E.

    2006-01-01

    We present a steady-state collisional-radiative model for the calculation of level populations in non-homogeneous plasmas with planar geometry. The line photon transport is taken into account following an angle- and frequency-averaged escape probability model. Several models where the same approach has been used can be found in the literature, but the main difference between our model and those ones is that the details of geometry are exactly treated in the definition of coupling coefficients and a local profile is taken into account in each plasma cell. (authors)

  9. Electron transport in nanometer GaAs structure under radiation exposure

    CERN Document Server

    Demarina, N V

    2002-01-01

    One investigates into effect of neutron and proton irradiation on electron transport in nanometer GaAs structures. Mathematical model takes account of radiation defects via introduction of additional mechanisms od scattering of carriers at point defects and disordered regions. To investigate experimentally into volt-ampere and volt-farad characteristics one used a structure based on a field-effect transistor with the Schottky gate and a built-in channel. Calculation results of electron mobility, drift rate of electrons, time of energy relaxation and electron pulse are compared with the experimental data

  10. Changing transport processes in the stratosphere by radiative heating of sulfate aerosols

    Directory of Open Access Journals (Sweden)

    U. Niemeier

    2017-12-01

    Full Text Available The injection of sulfur dioxide (SO2 into the stratosphere to form an artificial stratospheric aerosol layer is discussed as an option for solar radiation management. Sulfate aerosol scatters solar radiation and absorbs infrared radiation, which warms the stratospheric sulfur layer. Simulations with the general circulation model ECHAM5-HAM, including aerosol microphysics, show consequences of this warming, including changes of the quasi-biennial oscillation (QBO in the tropics. The QBO slows down after an injection of 4 Tg(S yr−1 and completely shuts down after an injection of 8 Tg(S yr−1. Transport of species in the tropics and sub-tropics depends on the phase of the QBO. Consequently, the heated aerosol layer not only impacts the oscillation of the QBO but also the meridional transport of the sulfate aerosols. The stronger the injection, the stronger the heating and the simulated impact on the QBO and equatorial wind systems. With increasing injection rate the velocity of the equatorial jet streams increases, and the less sulfate is transported out of the tropics. This reduces the global distribution of sulfate and decreases the radiative forcing efficiency of the aerosol layer by 10 to 14 % compared to simulations with low vertical resolution and without generated QBO. Increasing the height of the injection increases the radiative forcing only for injection rates below 10 Tg(S yr−1 (8–18 %, a much smaller value than the 50 % calculated previously. Stronger injection rates at higher levels even result in smaller forcing than the injections at lower levels.

  11. 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.

  12. Comparison of radiation measurements and calculations of reactor surroundings for skyshine analysis

    Energy Technology Data Exchange (ETDEWEB)

    Tsubosaka, A.; Nomura, Y. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kawabe, T. [Japan Research Institute, Limited, Osaka (Japan); Zharkov, V.P.; Kartashev, I.A.; Netecha, M.E.; Orlov, Y.V. [Research and Development Institute of Power Engineering, Moscow (Russian Federation)

    2000-03-01

    ISTC Project 'Experimental Studies of Radiation Scattering in the Atmosphere' were conducted using the IVG-1M and RA reactors by RDIPE in collaboration with IAE NNC RK and JAERI during 1996-1998. The radial distributions of fast neutron flux, thermal neutron flux and gamma radiation dose rate were measured above these two reactors at three heights. Neutron spectra above these two reactors and thermal and fast neutron fluxes over the hollow pipe height in the IVG-1M reactor were also measured in order to determine the radiation characteristics for skyshine analysis. For verifying the computer codes the calculations of reactor surroundings were performed using MCNP and DORT/DOT-3.5. The comparisons between the measurements and the calculations show that MCNP and DORT/DOT-3.5 codes can be widely applied to the shielding problems by selecting properly the calculation conditions. (author)

  13. ETRAN, Electron Transport and Gamma Transport with Secondary Radiation in Slab by Monte-Carlo

    International Nuclear Information System (INIS)

    1992-01-01

    A - Nature of physical problem solved: ETRAN computes the transport of electrons and photons through plane-parallel slab targets that have a finite thickness in one dimension and are unbound in the other two-dimensions. The incident radiation can consist of a beam of either electrons or photons with specified spectral and directional distribution. Options are available by which all orders of the electron-photon cascade can be included in the calculation. Thus electrons are allowed to give rise to secondary knock-on electrons, continuous Bremsstrahlung and characteristic x-rays; and photons are allowed to produce photo-electrons, Compton electrons, and electron- positron pairs. Annihilation quanta, fluorescence radiation, and Auger electrons are also taken into account. If desired, the Monte- Carlo histories of all generations of secondary radiations are followed. The information produced by ETRAN includes the following items: 1) reflection and transmission of electrons or photons, differential in energy and direction; 2) the production of continuous Bremsstrahlung and characteristic x-rays by electrons and the emergence of such radiations from the target (differential in photon energy and direction); 3) the spectrum of the amounts of energy left behind in a thick target by an incident electron beam; 4) the deposition of energy and charge by an electron beam as function of the depth in the target; 5) the flux of electrons, differential in energy, as function of the depth in the target. B - Method of solution: A programme called DATAPAC-4 takes data for a particular material from a library tape and further processes them. The function of DATAPAC-4 is to produce single-scattering and multiple-scattering data in the form of tabular arrays (again stored on magnetic tape) which facilitate the rapid sampling of electron and photon Monte Carlo histories in ETRAN. The photon component of the electron-photon cascade is calculated by conventional random sampling that imitates

  14. Calculation of radiation and pair production probabilities at arbitrary incidence angles to crystal planes

    International Nuclear Information System (INIS)

    Tikhonirov, V.V.

    1993-01-01

    The results of calculations of the intensity and polarization of radiation from channeled and unchanneled e +- are presented. The Fourier transformation (FT) is used to calculate numerous matrix elements. The calculations for channeled e + showed fast approach of spectral intensity to its value calculated in the approximation of self-consistent field (ASCF) with growing photon energy. In the case of 150 GeV unchanneled e - in Ge at T=293 K the ASCF gives a significantly higher value as compared to the FT. 4 refs., 3 figs

  15. A method for calculating effective lifetime risk of radiation-induced cancer from screening mammography

    International Nuclear Information System (INIS)

    Ali, R.M.; England, A.; McEntee, M.F.; Hogg, P.

    2015-01-01

    Purpose: To propose a method for evaluating the effective lifetime risk of radiation-induced cancer from screening mammography and to present initial data for the UK National Breast Screening Programme. Material and methods: The imaging was undertaken using a Hologic Selenia full field digital mammographic unit. The proposed method utilises an ATOM phantom containing thermoluminescent dosimeters and a perspex-polyethylene breast phantom to measure organ doses during a standard four view screening mammogram. Effective dose was calculated and effective risk was modelled for a range of client ages. The total lifetime effective risk was then calculated for the UK national screening programme. Calculation of effective risk includes the radiation dose to examined and contralateral breasts in addition to other body organs; this is an advantage over the mean glandular dose. Results: The contralateral breast, thyroid, thymus, brain, lung, salivary glands, and bone marrow all receive more than 1 μGy radiation dose during screening mammography. A major difference exists for total effective lifetime risk of radiation-induced cancer between clients with average and high breast cancer risk. Differences are attributed to the commencement age of screening and time interval between screens. Conclusion: This study proposes a method to evaluate effective lifetime risk of radiation-induced cancer from screening mammography in order to compare different mammography screening programmes. - Highlights: • We proposed a method for the calculation of radiation-induced cancer from screening mammography. • We measured the radiation absorbed dose of different organs during screening mammography. • There are major differences between mammography screening programme categories with regard to radiation effective risk.

  16. On calculating phase shifts and performing fits to scattering cross sections or transport properties

    International Nuclear Information System (INIS)

    Hepburn, J.W.; Roy, R.J. Le

    1978-01-01

    Improved methods of calculating quantum mechanical phase shifts and for performing least-squares fits to scattering cross sections or transport properties, are described. Their use in a five-parameter fit to experimental differential cross sections reduces the computer time by a factor of 4-7. (Auth.)

  17. Transport calculation of neutron flux distribution in reflector of PW reactor

    International Nuclear Information System (INIS)

    Remec, I.

    1982-01-01

    Two-dimensional transport calculation of the neutron flux and spectrum in the equatorial plain of PW reactor, using computer program DOT 3, is presented. Results show significant differences between neutron fields in which test samples and reactor vessel are exposed. (author)

  18. Neutron transport calculation for Activation Evaluation for Decommissioning of PET cyclotron Facility

    Science.gov (United States)

    Nobuhara, Fumiyoshi; Kuroyanagi, Makoto; Masumoto, Kazuyoshi; Nakamura, Hajime; Toyoda, Akihiro; Takahashi, Katsuhiko

    2017-09-01

    In order to evaluate the state of activation in a cyclotron facility used for the radioisotope production of PET diagnostics, we measured the neutron flux by using gold foils and TLDs. Then, the spatial distribution of neutrons and induced activity inside the cyclotron vault were simulated with the Monte Calro calculation code for neutron transport and DCHAIN-SP for activation calculation. The calculated results are in good agreement with measured values within factor 3. Therefore, the adaption of the advanced evaluation procedure for activation level is proved to be important for the planning of decommissioning of these facilities.

  19. Development of approximate shielding calculation method for high energy cosmic radiation on LEO satellites

    International Nuclear Information System (INIS)

    Sin, M. W.; Kim, M. H.

    2002-01-01

    To calculate total dose effect on semi-conductor devices in satellite for a period of space mission effectively, two approximate calculation models for a comic radiation shielding were proposed. They are a sectoring method and a chord-length distribution method. When an approximate method was applied in this study, complex structure of satellite was described into multiple 1-dimensional slabs, structural materials were converted to reference material(aluminum), and the pre-calculated dose-depth conversion function was introduced to simplify the calculation process. Verification calculation was performed for orbit location and structure geometry of KITSAT-1 and compared with detailed 3-dimensional calculation results and experimental values. The calculation results from approximate method were estimated conservatively with acceptable error. However, results for satellite mission simulation were underestimated in total dose rate compared with experimental values

  20. Development of approximate shielding calculation method for high energy cosmic radiation on LEO satellites

    Energy Technology Data Exchange (ETDEWEB)

    Sin, M. W.; Kim, M. H. [Kyunghee Univ., Yongin (Korea, Republic of)

    2002-10-01

    To calculate total dose effect on semi-conductor devices in satellite for a period of space mission effectively, two approximate calculation models for a comic radiation shielding were proposed. They are a sectoring method and a chord-length distribution method. When an approximate method was applied in this study, complex structure of satellite was described into multiple 1-dimensional slabs, structural materials were converted to reference material(aluminum), and the pre-calculated dose-depth conversion function was introduced to simplify the calculation process. Verification calculation was performed for orbit location and structure geometry of KITSAT-1 and compared with detailed 3-dimensional calculation results and experimental values. The calculation results from approximate method were estimated conservatively with acceptable error. However, results for satellite mission simulation were underestimated in total dose rate compared with experimental values.

  1. F--Ray: A new algorithm for efficient transport of ionizing radiation

    Science.gov (United States)

    Mao, Yi; Zhang, J.; Wandelt, B. D.; Shapiro, P. R.; Iliev, I. T.

    2014-04-01

    We present a new algorithm for the 3D transport of ionizing radiation, called F2-Ray (Fast Fourier Ray-tracing method). The transfer of ionizing radiation with long mean free path in diffuse intergalactic gas poses a special challenge to standard numerical methods which transport the radiation in position space. Standard methods usually trace each individual ray until it is fully absorbed by the intervening gas. If the mean free path is long, the computational cost and memory load are likely to be prohibitive. We have developed an algorithm that overcomes these limitations and is, therefore, significantly more efficient. The method calculates the transfer of radiation collectively, using the Fast Fourier Transform to convert radiation between position and Fourier spaces, so the computational cost will not increase with the number of ionizing sources. The method also automatically combines parallel rays with the same frequency at the same grid cell, thereby minimizing the memory requirement. The method is explicitly photon-conserving, i.e. the depletion of ionizing photons is guaranteed to equal the photoionizations they caused, and explicitly obeys the periodic boundary condition, i.e. the escape of ionizing photons from one side of a simulation volume is guaranteed to be compensated by emitting the same amount of photons into the volume through the opposite side. Together, these features make it possible to numerically simulate the transfer of ionizing photons more efficiently than previous methods. Since ionizing radiation such as the X-ray is responsible for heating the intergalactic gas when first stars and quasars form at high redshifts, our method can be applied to simulate thermal distribution, in addition to cosmic reionization, in three-dimensional inhomogeneous cosmological density field.

  2. Microwave emulations and tight-binding calculations of transport in polyacetylene

    International Nuclear Information System (INIS)

    Stegmann, Thomas; Franco-Villafañe, John A.; Ortiz, Yenni P.; Kuhl, Ulrich; Mortessagne, Fabrice; Seligman, Thomas H.

    2017-01-01

    A novel approach to investigate the electron transport of cis- and trans-polyacetylene chains in the single-electron approximation is presented by using microwave emulation measurements and tight-binding calculations. In the emulation we take into account the different electronic couplings due to the double bonds leading to coupled dimer chains. The relative coupling constants are adjusted by DFT calculations. For sufficiently long chains a transport band gap is observed if the double bonds are present, whereas for identical couplings no band gap opens. The band gap can be observed also in relatively short chains, if additional edge atoms are absent, which cause strong resonance peaks within the band gap. The experimental results are in agreement with our tight-binding calculations using the nonequilibrium Green's function method. The tight-binding calculations show that it is crucial to include third nearest neighbor couplings to obtain the gap in the cis-polyacetylene. - Highlights: • Electronic transport in individual polyacetylene chains is studied. • Microwave emulation experiments and tight-binding calculations agree well. • In long chains a band-gap opens due the dimerization of the chain. • In short chains edge atoms cause strong resonance peaks in the center of the band-gap.

  3. Microwave emulations and tight-binding calculations of transport in polyacetylene

    Energy Technology Data Exchange (ETDEWEB)

    Stegmann, Thomas, E-mail: stegmann@icf.unam.mx [Instituto de Ciencias Físicas, Universidad Nacional Autónoma de México, Avenida Universidad s/n, 62210 Cuernavaca (Mexico); Franco-Villafañe, John A., E-mail: jofravil@fis.unam.mx [Instituto de Física, Benemérita Universidad Autónoma de Puebla, Apartado Postal J-48, 72570 Puebla (Mexico); Instituto de Ciencias Físicas, Universidad Nacional Autónoma de México, Avenida Universidad s/n, 62210 Cuernavaca (Mexico); Ortiz, Yenni P. [Instituto de Ciencias Físicas, Universidad Nacional Autónoma de México, Avenida Universidad s/n, 62210 Cuernavaca (Mexico); Kuhl, Ulrich [Université de Nice – Sophia Antipolis, Laboratoire de la Physique de la Matière Condensée, CNRS, Parc Valrose, 06108 Nice (France); Mortessagne, Fabrice, E-mail: fabrice.mortessagne@unice.fr [Université de Nice – Sophia Antipolis, Laboratoire de la Physique de la Matière Condensée, CNRS, Parc Valrose, 06108 Nice (France); Seligman, Thomas H. [Instituto de Ciencias Físicas, Universidad Nacional Autónoma de México, Avenida Universidad s/n, 62210 Cuernavaca (Mexico); Centro Internacional de Ciencias, 62210 Cuernavaca (Mexico)

    2017-01-05

    A novel approach to investigate the electron transport of cis- and trans-polyacetylene chains in the single-electron approximation is presented by using microwave emulation measurements and tight-binding calculations. In the emulation we take into account the different electronic couplings due to the double bonds leading to coupled dimer chains. The relative coupling constants are adjusted by DFT calculations. For sufficiently long chains a transport band gap is observed if the double bonds are present, whereas for identical couplings no band gap opens. The band gap can be observed also in relatively short chains, if additional edge atoms are absent, which cause strong resonance peaks within the band gap. The experimental results are in agreement with our tight-binding calculations using the nonequilibrium Green's function method. The tight-binding calculations show that it is crucial to include third nearest neighbor couplings to obtain the gap in the cis-polyacetylene. - Highlights: • Electronic transport in individual polyacetylene chains is studied. • Microwave emulation experiments and tight-binding calculations agree well. • In long chains a band-gap opens due the dimerization of the chain. • In short chains edge atoms cause strong resonance peaks in the center of the band-gap.

  4. Light scattering by multiple spheres: comparison between Maxwell theory and radiative-transfer-theory calculations.

    Science.gov (United States)

    Voit, Florian; Schäfer, Jan; Kienle, Alwin

    2009-09-01

    We present a methodology to compare results of classical radiative transfer theory against exact solutions of Maxwell theory for a high number of spheres. We calculated light propagation in a cubic scattering region (20 x 20 x 20 microm(3)) consisting of different concentrations of polystyrene spheres in water (diameter 2 microm) by an analytical solution of Maxwell theory and by a numerical solution of radiative transfer theory. The relative deviation of differential as well as total scattering cross sections obtained by both approaches was evaluated for each sphere concentration. For the considered case, we found that deviations due to radiative transfer theory remain small, even for concentrations up to ca. 20 vol. %.

  5. Optimized shielding calculation to the transport of 131I employed in nuclear medicine

    International Nuclear Information System (INIS)

    Sahyun, A.; Sordi, G.M.; Rodrigues, D.; Sanches, M.P.; Romero F, C.R.

    1996-01-01

    The objective of this paper is to present the basis for shielding calculation used in different situations that could occur during the transport of 131 I utilized in nuclear medicine for diagnostic and therapeutic purposes. The aim of these calculation is to optimize the shielding in order to satisfy the transport of radioactive material. These calculations were proposed for estimated activities around 1,85 GBq (50mCi), 3,7 GBq(100mCi) and 7,4 GBq(200mCi), considering the driver of the cargo company and his assistant as the critical group and the general people considered as effect of collective dose. The population density considered in the models is the one related to Sao Paulo city, because the transport is done by the highway across the city and the radioactive material is distributed from west to north and south, where the airports are located. This area ranges a perimeter of 40 km. For the collective dose calculation, it was considered a population dose of less than 1/100 of the annual limit dose for the public. Our main concern is related to the large volume of radioactive material that is transported per week, specially because 1/3 of this material has activities around 3,7 GBq (100mCi). During the calculations, we have figured out that the activities at the moment of transport are nearly 40% greater than the one related to the calibration date. As for the discrepancy of official alpha value of US$10000/man-Sv and the real value for our country of US$3000/man-Sv,a comparative study was performed. (authors). 3 refs., 2 figs., 2 tabs

  6. Examinations on Applications of Manual Calculation Programs on Lung Cancer Radiation Therapy Using Analytical Anisotropic Algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Min; Kim, Dae Sup; Hong, Dong Ki; Back, Geum Mun; Kwak, Jung Won [Dept. of Radiation Oncology, , Seoul (Korea, Republic of)

    2012-03-15

    There was a problem with using MU verification programs for the reasons that there were errors of MU when using MU verification programs based on Pencil Beam Convolution (PBC) Algorithm with radiation treatment plans around lung using Analytical Anisotropic Algorithm (AAA). On this study, we studied the methods that can verify the calculated treatment plans using AAA. Using Eclipse treatment planning system (Version 8.9, Varian, USA), for each 57 fields of 7 cases of Lung Stereotactic Body Radiation Therapy (SBRT), we have calculated using PBC and AAA with dose calculation algorithm. By developing MU of established plans, we compared and analyzed with MU of manual calculation programs. We have analyzed relationship between errors and 4 variables such as field size, lung path distance of radiation, Tumor path distance of radiation, effective depth that can affect on errors created from PBC algorithm and AAA using commonly used programs. Errors of PBC algorithm have showned 0.2{+-}1.0% and errors of AAA have showned 3.5{+-}2.8%. Moreover, as a result of analyzing 4 variables that can affect on errors, relationship in errors between lung path distance and MU, connection coefficient 0.648 (P=0.000) has been increased and we could calculate MU correction factor that is A.E=L.P 0.00903+0.02048 and as a result of replying for manual calculation program, errors of 3.5{+-}2.8% before the application has been decreased within 0.4{+-}2.0%. On this study, we have learned that errors from manual calculation program have been increased as lung path distance of radiation increases and we could verified MU of AAA with a simple method that is called MU correction factor.

  7. Examinations on Applications of Manual Calculation Programs on Lung Cancer Radiation Therapy Using Analytical Anisotropic Algorithm

    International Nuclear Information System (INIS)

    Kim, Jung Min; Kim, Dae Sup; Hong, Dong Ki; Back, Geum Mun; Kwak, Jung Won

    2012-01-01

    There was a problem with using MU verification programs for the reasons that there were errors of MU when using MU verification programs based on Pencil Beam Convolution (PBC) Algorithm with radiation treatment plans around lung using Analytical Anisotropic Algorithm (AAA). On this study, we studied the methods that can verify the calculated treatment plans using AAA. Using Eclipse treatment planning system (Version 8.9, Varian, USA), for each 57 fields of 7 cases of Lung Stereotactic Body Radiation Therapy (SBRT), we have calculated using PBC and AAA with dose calculation algorithm. By developing MU of established plans, we compared and analyzed with MU of manual calculation programs. We have analyzed relationship between errors and 4 variables such as field size, lung path distance of radiation, Tumor path distance of radiation, effective depth that can affect on errors created from PBC algorithm and AAA using commonly used programs. Errors of PBC algorithm have showned 0.2±1.0% and errors of AAA have showned 3.5±2.8%. Moreover, as a result of analyzing 4 variables that can affect on errors, relationship in errors between lung path distance and MU, connection coefficient 0.648 (P=0.000) has been increased and we could calculate MU correction factor that is A.E=L.P 0.00903+0.02048 and as a result of replying for manual calculation program, errors of 3.5±2.8% before the application has been decreased within 0.4±2.0%. On this study, we have learned that errors from manual calculation program have been increased as lung path distance of radiation increases and we could verified MU of AAA with a simple method that is called MU correction factor.

  8. Program for calculating multi-component high-intense ion beam transport

    International Nuclear Information System (INIS)

    Kazarinov, N.Yu.; Prejzendorf, V.A.

    1985-01-01

    The CANAL program for calculating transport of high-intense beams containing ions with different charges in a channel consisting of dipole magnets and quadrupole lenses is described. The equations determined by the method of distribution function momenta and describing coordinate variations of the local mass centres and r.m.s. transverse sizes of beams with different charges form the basis of the calculation. The program is adapted for the CDC-6500 and SM-4 computers. The program functioning is organized in the interactive mode permitting to vary the parameters of any channel element and quickly choose the optimum version in the course of calculation. The calculation time for the CDC-6500 computer for the 30-40 m channel at the integration step of 1 cm is about 1 min. The program is used for calculating the channel for the uranium ion beam injection from the collective accelerator into the heavy-ion synchrotron

  9. Calculation of Smith-Purcell radiation from a volume strip grating

    International Nuclear Information System (INIS)

    Kube, G.

    2005-01-01

    Smith-Purcell radiation is generated by a charged particle beam passing close to the surface of a diffraction grating. Experimental investigations show a strong dependency of the emitted radiation intensity on the form of the grating profile. This influence is expressed by the radiation factor which is a measure of the grating efficiency, in close analogy to reflection coefficients of optical grating theories. The radiation factor depends on beam energy and observation geometry. Up to now calculations for radiation factors exist for lamellar, sinusoidal and echelette-type grating profiles. In this paper, calculations of Smith-Purcell radiation factors for volume strip gratings which are separated by vacuum gaps are presented. They are based on the modal expansion method and restricted to perfectly conducting grating surfaces and to electron trajectories perpendicular to the grating grooves. An infinite system of coupled linear algebraic equations for the scattered and the transmitted wave amplitudes is derived by imposing the continuity condition at the open end of the grooves, and by the boundary conditions at the remaining part of the interface. Numerical results are presented and discussed in view of using Smith-Purcell radiation for particle beam diagnostic purposes

  10. Energy distributions and radiation transport in uranium plasmas

    International Nuclear Information System (INIS)

    Miley, G.; Bathke, C.; Maceda, E.; Choi, C.

    1976-01-01

    Electron energy distribution functions have been calculated in a 235 U-plasma at 1 atmosphere for various plasma temperatures (5000 to 8000 0 K) and neutron fluxes (2 x 10 12 to 2 x 10 16 neutrons/(cm 2 -sec)). Two sources of energetic electrons are included; namely fission-fragment and electron-impact ionization, resulting in a high-energy tail superimposed on the thermalized electron distribution. Consequential derivations from equilibrium collision rates are of interest relative to direct pumping of lasers and radiation emission. Results suggest that non-equilibrium excitation can best be achieved with an additive gas such as helium or in lower temperature plasmas requiring UF 6 . An approximate analytic model, based on continuous electron slowing, has been used for survey calculations. Where more accuracy is required, a Monte Carlo technique is used which combines an analytic representation of Coulombic collisions with a random-walk treatment of inelastic collisions

  11. Observation and calculation of the solar radiation on the Tibetan Plateau

    International Nuclear Information System (INIS)

    Liu Jiandong; Liu Jingmiao; Linderholm, Hans W.; Chen Deliang; Yu Qiang; Wu Dingrong; Haginoya, Shigenori

    2012-01-01

    Highlights: ► Five years of continuous solar-radiation observations on the Tibetan Plateau were analyzed. ► Eight solar models were calibrated and validated in this highland region. ► A strategy for calculating solar radiation on the Tibetan Plateau was presented. - Abstract: Distribution of solar radiation is vital to locate the most suitable regions for harvesting solar energy, but solar radiation is only observed at few stations due to high costs and difficult maintenance. From 2001 to 2005, a set of pyranometer instruments were set up in Gaize, on the Tibetan Plateau, to test the hypothesis of high solar-radiation levels in this region, and find a suitable method for estimating the radiation. Over the 5-year observation period, the average daily radiation was 21 MJ m −2 day −1 with maximum daily values of 27 MJ m −2 day −1 occurring in June and minimum values of 14 MJ m −2 day −1 in December, which is much higher than those measured in other regions at similar latitudes. The observational data were used to validate a set of radiation models: five sunshine based and three temperature based. The results showed that of the five sunshine-based models, a newly developed “comprehensive” model performed the best, but that the “vapor revised Angstrom model” was recommended to use for its simplicity and easy operation. The temperature-based models performed worse than the sunshine-based ones, where the Wu model is to be preferred if a temperature-based model is the only option. Moreover, it was shown that when estimating the solar radiation based on time-dependent coefficients, consideration of the seasonal variation of the coefficients has little predictive value and is thus unnecessary. Based on the results of this study, a strategy for the calculation of solar radiation on the Tibetan Plateau was made for potential users.

  12. Calculation of radiation exposures from patients to whom radioactive materials have been administered

    International Nuclear Information System (INIS)

    McCormack, J.; Shearer, J.

    1998-01-01

    Spreadsheet templates have been developed by the authors to calculate radiation exposures to others from patients to whom radioactive materials have been administered (or, indeed, from any source of radiation exposure) to be readily calculated. The time during which contact should be avoided, along with the residual activity at resumption of contact is also calculated using an iterative technique. These spreadsheets allow a great deal of flexibility in the specification of clearance rates and close contact patterns for individual patients. Estimates of doses, restriction times and residual activities for 131 l thyrotoxic therapy, for various contact patterns and group of patients, were calculated. The spreadsheets are implemented using Microsoft EXCEL for both PC and Macintosh computers, and are readily available from the authors

  13. A Deterministic Electron, Photon, Proton and Heavy Ion Radiation Transport Suite for the Study of the Jovian System

    Science.gov (United States)

    Norman, Ryan B.; Badavi, Francis F.; Blattnig, Steve R.; Atwell, William

    2011-01-01

    A deterministic suite of radiation transport codes, developed at NASA Langley Research Center (LaRC), which describe the transport of electrons, photons, protons, and heavy ions in condensed media is used to simulate exposures from spectral distributions typical of electrons, protons and carbon-oxygen-sulfur (C-O-S) trapped heavy ions in the Jovian radiation environment. The particle transport suite consists of a coupled electron and photon deterministic transport algorithm (CEPTRN) and a coupled light particle and heavy ion deterministic transport algorithm (HZETRN). The primary purpose for the development of the transport suite is to provide a means for the spacecraft design community to rapidly perform numerous repetitive calculations essential for electron, proton and heavy ion radiation exposure assessments in complex space structures. In this paper, the radiation environment of the Galilean satellite Europa is used as a representative boundary condition to show the capabilities of the transport suite. While the transport suite can directly access the output electron spectra of the Jovian environment as generated by the Jet Propulsion Laboratory (JPL) Galileo Interim Radiation Electron (GIRE) model of 2003; for the sake of relevance to the upcoming Europa Jupiter System Mission (EJSM), the 105 days at Europa mission fluence energy spectra provided by JPL is used to produce the corresponding dose-depth curve in silicon behind an aluminum shield of 100 mils ( 0.7 g/sq cm). The transport suite can also accept ray-traced thickness files from a computer-aided design (CAD) package and calculate the total ionizing dose (TID) at a specific target point. In that regard, using a low-fidelity CAD model of the Galileo probe, the transport suite was verified by comparing with Monte Carlo (MC) simulations for orbits JOI--J35 of the Galileo extended mission (1996-2001). For the upcoming EJSM mission with a potential launch date of 2020, the transport suite is used to compute

  14. Evaluation of secondary exposure doses to transportation and medical personnel in the radiation emergency medicine

    International Nuclear Information System (INIS)

    Kato, Hideki; Suzuki, Shoichi; Koga, Sukehiko; Mukoyama, Takashi; Tomatsu, Hirotaka; Suzuki, Yusuke

    2009-01-01

    Radionuclide contamination is one of the accidents that might occur while carrying out a periodical inspection of nuclear power stations during normal reactor operation. When such an accident occurs, rescue and medical personnel, involved in transporting and treating affected workers run the risk of exposure to secondary radiation. In this study, the ambient dose equivalent rate at a certain distance from the surface of the human body contaminated with typical radioactive corrosion products was calculated. Further, the relationships among the adhesion area, contamination density, and secondary exposure dose were clarified. The secondary exposure dose and permissible working hours in a radiation emergency medicine were estimated by presenting these relationships in the form of a chart and by specifying the contamination levels. (author)

  15. The correlation schemes in calculations of the rate constants of some radiation chemical reactions

    International Nuclear Information System (INIS)

    Zagorets, P.A.; Shostenko, A.G.; Kim, V.

    1983-01-01

    The various correlation relationships of the evaluation of the rate constants of radiation chemical reactions of addition, abstraction and isomerization were considered. It was shown that neglection of the influence of solvent can result in errors in calculations of rate constants equalling two orders in magnitude. Several examples of isokinetic relationship are given. The methods of calculation of transmission coefficient of reaction addition have been discussed. (author)

  16. Survey of radiation protection programmes for transport; Etude des programmes de radioprotection pour les transports de matieres radioactives

    Energy Technology Data Exchange (ETDEWEB)

    Lizot, M.T.; Perrin, M.L.; Sert, G. [CEA Fontenay-aux-Roses, Inst. de Protection et de Surete Nucleaire, Dept. de Protection et de Surete Nucleaire, 92 (France); Lange, F.; Schwarz, G.; Feet, H.J.; Christ, R. [Gesellschaft fur Anlagen-und Reaktorsicherheit, GRS, mbH, Cologne (Germany); Shaw, K.B.; Hughes, J.S.; Gelder, R. [National Radiological Protection Board (NRPB), Oxon, OX (United Kingdom)

    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.)

  17. Risks of transport of radioactive materials on the road; some exploring calculations performed with the INTERTRAN-model

    International Nuclear Information System (INIS)

    1987-04-01

    Under the auspices of the IAEA a computercode, named INTERTRAN, has been developed in order to calculate the risks of the transport of radioactive materials. This code has to be tested nearer. For the Dutch situation a number of calculations has been performed of more or less realistic cases in which four transport streams have been investigated. Two transport routes are chosen. The risks thus obtained are compared quantitatively with the risks of LPG-transports. 4 refs.; 9 figs

  18. Calculation and interpretation of In-Situ measurements of initial radiations at Hiroshima and Nagasaki

    International Nuclear Information System (INIS)

    Loewe, W.E.

    1983-01-01

    Cobalt activation calculations will be reviewed, and similar comparisons of sulfur activation interior to electrical insulators on power transmission lines will be discussed. The relationship between neutron tissue kermas one to two kilometers from hypocenter and the particular activations of cobalt and sulfur are reviewed. At present, measured and calculated quantities agree within associated uncertainties, which are substantial. Additional work to shrink these uncertainties will be discussed. Particular cobalt activation topics will include: the sensitivity to thermal neutrons outside the pillar; calculated values using actual Nagasaki concrete composition; and calculational advances to improve modelling of the actual configuration. Particular sulfur activation topics will include: absolute comparisons of measured and calculated ratios of dpm/gm of 32 P at all measured ranges, based on approximate experimental values for insulator attentuation and source radiations; the relationship between sulfur activation within a kilometer of hypocenter and kermas at two kilometers; and calculational advances to improve modelling of the actual configuration

  19. 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...

  20. Search for a transport method for the calculation of the PWR control and safety clusters

    International Nuclear Information System (INIS)

    Bruna, G.B.; Van Frank, C.; Vergain, M.L.; Chauvin, J.P.; Palmiotti, G.; Nobile, M.

    1990-01-01

    The project studies of power reactors rely mainly on diffusion calculations, but transport ones are often needed for assessing fine effects, intimately linked to geometry and spectrum heterogeneities. Accurate transport computations are necessary, in particular, for shielded cross section generation, and when homogenization and dishomogenization processes are involved. The transport codes, generally, offer the user a variety of computational options, related to different approximation levels. In every case, it is obviously desirable to be able to choose the reliable degree of approximation to be accepted in any particular computational circumstance of the project. The search for such adapted procedures is to be made on the basis of critical experiments. In our studies, this task was made possible by the availability of suitable results of the CAMELEON critical experiment, carried on in the EOLE facility at CEA's Center of Cadarache. In this paper, we summarize some of the work in progress at FRAMATOME on the definition of an assembly based transport calculation scheme to be used for PWR control and safety cluster computations. Two main items, devoted to the search of the optimum computational procedures, are presented here: - a parametrical study on computational options, made in an infinite medium assembly geometry, - a series of comparisons between calculated and experimental values of pin power distribution

  1. International report to validate criticality safety calculations for fissile material transport

    International Nuclear Information System (INIS)

    Whitesides, G.E.

    1984-01-01

    During the past three years a Working Group established by the Organization for Economic Co-operation and Development's Nuclear Energy Agency (OECD-NEA) in Paris, France, has been studying the validity and applicability of a variety of criticality safety computer programs and their associated nuclear data for the computation of the neutron multiplication factor, k/sub eff/, for various transport packages used in the fuel cycle. The principal objective of this work has been to provide an internationally acceptable basis for the licensing authorities in a country to honor licensing approvals granted by other participating countries. Eleven countries participated in the initial study which consisted of examining criticality safety calculations for packages designed for spent light water reactor fuel transport. This paper presents a summary of this study which has been completed and reported in an OECD-NEA Report No. CSNI-71. The basic goal of this study was to outline a satisfactory validation procedure for this particular application. First, a set of actual critical experiments were chosen which contained the various material and geometric properties present in typical LWR transport containers. Secondly, calculations were made by each of the methods in order to determine how accurately each method reproduced the experimental values. This successful effort in developing a benchmark procedure for validating criticality calculations for spent LWR transport packages along with the successful intercomparison of a number of methods should provide increased confidence by licensing authorities in the use of these methods for this area of application. 4 references, 2 figures

  2. Calculation of the secondary gamma radiation by the Monte Carlo method at displaced sampling from distributed sources

    International Nuclear Information System (INIS)

    Petrov, Eh.E.; Fadeev, I.A.

    1979-01-01

    A possibility to use displaced sampling from a bulk gamma source in calculating the secondary gamma fields by the Monte Carlo method is discussed. The algorithm proposed is based on the concept of conjugate functions alongside the dispersion minimization technique. For the sake of simplicity a plane source is considered. The algorithm has been put into practice on the M-220 computer. The differential gamma current and flux spectra in 21cm-thick lead have been calculated. The source of secondary gamma-quanta was assumed to be a distributed, constant and isotropic one emitting 4 MeV gamma quanta with the rate of 10 9 quanta/cm 3 xs. The calculations have demonstrated that the last 7 cm of lead are responsible for the whole gamma spectral pattern. The spectra practically coincide with the ones calculated by the ROZ computer code. Thus the algorithm proposed can be offectively used in the calculations of secondary gamma radiation transport and reduces the computation time by 2-4 times

  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. Optical photon transport in powdered-phosphor scintillators. Part II. Calculation of single-scattering transport parameters

    Energy Technology Data Exchange (ETDEWEB)

    Poludniowski, Gavin G. [Joint Department of Physics, Division of Radiotherapy and Imaging, Institute of Cancer Research and Royal Marsden NHS Foundation Trust, Downs Road, Sutton, Surrey SM2 5PT, United Kingdom and Centre for Vision Speech and Signal Processing (CVSSP), Faculty of Engineering and Physical Sciences, University of Surrey, Guildford, Surrey GU2 7XH (United Kingdom); Evans, Philip M. [Centre for Vision Speech and Signal Processing (CVSSP), Faculty of Engineering and Physical Sciences, University of Surrey, Guildford, Surrey GU2 7XH (United Kingdom)

    2013-04-15

    Purpose: Monte Carlo methods based on the Boltzmann transport equation (BTE) have previously been used to model light transport in powdered-phosphor scintillator screens. Physically motivated guesses or, alternatively, the complexities of Mie theory have been used by some authors to provide the necessary inputs of transport parameters. The purpose of Part II of this work is to: (i) validate predictions of modulation transform function (MTF) using the BTE and calculated values of transport parameters, against experimental data published for two Gd{sub 2}O{sub 2}S:Tb screens; (ii) investigate the impact of size-distribution and emission spectrum on Mie predictions of transport parameters; (iii) suggest simpler and novel geometrical optics-based models for these parameters and compare to the predictions of Mie theory. A computer code package called phsphr is made available that allows the MTF predictions for the screens modeled to be reproduced and novel screens to be simulated. Methods: The transport parameters of interest are the scattering efficiency (Q{sub sct}), absorption efficiency (Q{sub abs}), and the scatter anisotropy (g). Calculations of these parameters are made using the analytic method of Mie theory, for spherical grains of radii 0.1-5.0 {mu}m. The sensitivity of the transport parameters to emission wavelength is investigated using an emission spectrum representative of that of Gd{sub 2}O{sub 2}S:Tb. The impact of a grain-size distribution in the screen on the parameters is investigated using a Gaussian size-distribution ({sigma}= 1%, 5%, or 10% of mean radius). Two simple and novel alternative models to Mie theory are suggested: a geometrical optics and diffraction model (GODM) and an extension of this (GODM+). Comparisons to measured MTF are made for two commercial screens: Lanex Fast Back and Lanex Fast Front (Eastman Kodak Company, Inc.). Results: The Mie theory predictions of transport parameters were shown to be highly sensitive to both grain size

  5. X-radiation effect on water transport in ascite cells of Ehrlich carcinoma

    International Nuclear Information System (INIS)

    Barnov, V.A.; Ajvazishvili, M.A.; Kartvelishvili, I.I.; Tushishvili, D.I.

    1988-01-01

    Effect of local X radiation with doses 0.05 and 0.15 C/kg on water transport in ascitic cells of Erlich carcinoma is studied in rats. To study water transport through cell membranes, tritium mark was used. It is concluded that radiation effect on water transport in cells of Erlich carcinoma may be related to change in ionic permittivity of the membrane, because small changes in transmembrane ion transport affect immediately the osmotic motion of water. 5 refs

  6. 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

  7. The effect of gamma-ray transport on afterheat calculations for accident analysis

    International Nuclear Information System (INIS)

    Reyes, S.; Latkowski, J.F.; Sanz, J.

    2000-01-01

    Radioactive afterheat is an important source term for the release of radionuclides in fusion systems under accident conditions. Heat transfer calculations are used to determine time-temperature histories in regions of interest, but the true source term needs to be the effective afterheat, which considers the transport of penetrating gamma rays. Without consideration of photon transport, accident temperatures may be overestimated in others. The importance of this effect is demonstrated for a simple, one-dimensional problem. The significance of this effect depends strongly on the accident scenario being analyzed

  8. Time dependent AN neutron transport calculations in finite media using a numerical inverse Laplace transform technique

    International Nuclear Information System (INIS)

    Ganapol, B.D.; Sumini, M.

    1990-01-01

    The time dependent space second order discrete form of the monokinetic transport equation is given an analytical solution, within the Laplace transform domain. Th A n dynamic model is presented and the general resolution procedure is worked out. The solution in the time domain is then obtained through the application of a numerical transform inversion technique. The justification of the research relies in the need to produce reliable and physically meaningful transport benchmarks for dynamic calculations. The paper is concluded by a few results followed by some physical comments

  9. Intercomparison of radiative forcing calculations of stratospheric water vapour and contrails

    Energy Technology Data Exchange (ETDEWEB)

    Myhre, Gunnar [Dept. of Geosciences, Univ. of Oslo (Norway); Center for International Climate and Environmental Research-Oslo (CICERO), Oslo (Norway); Kvalevaag, Maria [Dept. of Geosciences, Univ. of Oslo (Norway); Raedel, Gaby; Cook, Jolene; Shine, Keith P. [Dept. of Meteorology, Univ. of Reading (United Kingdom); Clark, Hannah [CNRM/GAME Meteo France, Toulouse (France); Lab. d' Aerologie, Univ. de Toulouse (France); Karcher, Fernand [CNRM/GAME Meteo France, Toulouse (France); Markowicz, Krzysztof; Kardas, Aleksandra; Wolkenberg, Paulina [Inst. of Geophysics, Univ. of Warsaw (Poland); Balkanski, Yves [LSCE/IPSL, Lab. CEA-CNRS-UVSQ (France); Ponater, Michael [Deutsches Zentrum fuer Luft und Raumfahrt (DLR), Inst. fuer Physik der Atmosphaere, Oberpfaffenhofen (Germany); Forster, Piers; Rap, Alexandru [School of Earth and Environment, Univ. of Leeds (United Kingdom); Leon, Ruben Rodriguez de [Manchester Metropolitan Univ. (United Kingdom)

    2009-12-15

    Seven groups have participated in an intercomparison study of calculations of radiative forcing (RF) due to stratospheric water vapour (SWV) and contrails. a combination of detailed radiative transfer schemes and codes for global-scale calculations have been used, as well as a combination of idealized simulations and more realistic global-scale changes in stratospheric water vapour and contrails. Detailed line-by-line codes agree within about 15% for longwave (LW) and shortwave (SW) RF, except in one case where the difference is 30%. Since the LW and SW RF due to contrails and SWV changes are of opposite sign, the differences between the models seen in the individual LW and SW components can be either compensated or strengthened in the net RF. and thus in relative terms uncertainties are much larger for the net RF. Some of the models used for global-scale simulations of changes in SWV and contrails differ substantially in RF from the more detailed radiative transfer schemes. For the global-scale calculations we use a method of weighting the results to calculate a best estimate based on their performance compared to the more detailed radiative transfer schemes in the idealized simulations. (orig.)

  10. Calculation of radiation heat generation on a graphite reflector side of IAN-R1 Reactor

    International Nuclear Information System (INIS)

    Duque O, J.; Velez A, L.H.

    1987-01-01

    Calculation methods for radiation heat generation in nuclear reactor, based on the point kernel approach are revisited and applied to the graphite reflector of IAN-R1 reactor. A Fortran computer program was written for the determination of total heat generation in the reflector, taking 1155 point in it

  11. Model-Independent Calculation of Radiative Neutron Capture on Lithium-7

    NARCIS (Netherlands)

    Rupak, Gautam; Higa, Renato

    2011-01-01

    The radiative neutron capture on lithium-7 is calculated model independently using a low-energy halo effective field theory. The cross section is expressed in terms of scattering parameters directly related to the S-matrix elements. It depends on the poorly known p-wave effective range parameter

  12. Calculation of radiation loss of 1. 2 GeV-electrons in a thick silicon monocrystal

    Energy Technology Data Exchange (ETDEWEB)

    Keshtova, S.V.; Komarov, F.F.; Telegin, V.I.

    1988-10-01

    The angular distribution of radiation loss of different fractions of 1.2 GeV-electrons during axial channeling in a Si monocrystal of 1.6 mm thickness is discussed. The results of the numerical calculations are compared with the experimental data.

  13. 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

  14. 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

  15. Charge transport properties of CdMnTe radiation detectors

    Energy Technology Data Exchange (ETDEWEB)

    Kim K.; Rafiel, R.; Boardman, M.; Reinhard, I.; Sarbutt, A.; Watt, G.; Watt, C.; Uxa, S.; Prokopovich, D.A.; Belas, E.; Bolotnikov, A.E.; James, R.B.

    2012-04-11

    Growth, fabrication and characterization of indium-doped cadmium manganese telluride (CdMnTe)radiation detectors have been described. Alpha-particle spectroscopy measurements and time resolved current transient measurements have yielded an average charge collection efficiency approaching 100 %. Spatially resolved charge collection efficiency maps have been produced for a range of detector bias voltages. Inhomogeneities in the charge transport of the CdMnTe crystals have been associated with chains of tellurium inclusions within the detector bulk. Further, it has been shown that the role of tellurium inclusions in degrading chargecollection is reduced with increasing values of bias voltage. The electron transit time was determined from time of flight measurements. From the dependence of drift velocity on applied electric field the electron mobility was found to be n = (718 55) cm2/Vs at room temperature.

  16. Radiation skyshine calculation with MARS15 for the Mu2e Experiment at Fermilab

    International Nuclear Information System (INIS)

    Leveling, A.F.

    2015-01-01

    The Fermilab Antiproton source is to be re-purposed to provide an 8 kW proton beam to the Mu2e experiment by 1/3 integer, slow resonant extraction. Shielding provided by the existing facility must be supplemented with in-tunnel shielding to limit the radiation effective dose rate above the shield in the AP30 service building. In addition to the nominal radiation shield calculations, radiation skyshine calculations were required to ensure compliance with Fermilab Radiological Controls Manual. A complete model of the slow resonant extraction system including magnets, electrostatic septa, magnetic fields, tunnel enclosure with shield, and a nearby exit stairway are included in the model. The skyshine model extends above the beam enclosure surface to 10 km vertically and 5 km radially. (authors)

  17. 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

  18. Transport of radioactive wastes to the planned final waste repository Konrad: Radiation exposure resulting from normal transport and radiological risks from transport accidents

    International Nuclear Information System (INIS)

    Lange, F.; Fett, H.J.; Gruendler, D.; Schwarz, G.

    1993-01-01

    Radiation exposures of members of critical groups of the general population and of transport personnel resulting from normal transport of radioactive wastes to the planned final waste repository Konrad have been evaluated in detail. By applying probabilistic safety assessment techniques radiological risks from transport accidents have been analysed by quantifying potential radiation exposures and contaminations of the biosphere in connection with their expected frequencies of occurrence. The Konrad transport study concentrates on the local region of the waste repository, where all transports converge. (orig.) [de

  19. Evaluation and comparison of SN and Monte-Carlo charged particle transport calculations

    International Nuclear Information System (INIS)

    Hadad, K.

    2000-01-01

    A study was done to evaluate a 3-D S N charged particle transport code called SMARTEPANTS 1 and another 3-D Monte Carlo code called Integrated Tiger Series, ITS 2 . The evaluation study of SMARTEPANTS code was based on angular discretization and reflected boundary sensitivity whilst the evaluation of ITS was based on CPU time and variance reduction. The comparison of the two code was based on energy and charge deposition calculation in block of Gallium Arsenide with embedded gold cylinders. The result of evaluation tests shows that an S 8 calculation maintains both accuracy and speed and calculations with reflected boundaries geometry produces full symmetrical results. As expected for ITS evaluation, the CPU time and variance reduction are opposite to a point beyond which the history augmentation while increasing the CPU time do not result in variance reduction. The comparison test problem showed excellent agreement in total energy deposition calculations

  20. Thermal transport across metal silicide-silicon interfaces: First-principles calculations and Green's function transport simulations

    Science.gov (United States)

    Sadasivam, Sridhar; Ye, Ning; Feser, Joseph P.; Charles, James; Miao, Kai; Kubis, Tillmann; Fisher, Timothy S.

    2017-02-01

    Heat transfer across metal-semiconductor interfaces involves multiple fundamental transport mechanisms such as elastic and inelastic phonon scattering, and electron-phonon coupling within the metal and across the interface. The relative contributions of these different transport mechanisms to the interface conductance remains unclear in the current literature. In this work, we use a combination of first-principles calculations under the density functional theory framework and heat transport simulations using the atomistic Green's function (AGF) method to quantitatively predict the contribution of the different scattering mechanisms to the thermal interface conductance of epitaxial CoSi2-Si interfaces. An important development in the present work is the direct computation of interfacial bonding from density functional perturbation theory (DFPT) and hence the avoidance of commonly used "mixing rules" to obtain the cross-interface force constants from bulk material force constants. Another important algorithmic development is the integration of the recursive Green's function (RGF) method with Büttiker probe scattering that enables computationally efficient simulations of inelastic phonon scattering and its contribution to the thermal interface conductance. First-principles calculations of electron-phonon coupling reveal that cross-interface energy transfer between metal electrons and atomic vibrations in the semiconductor is mediated by delocalized acoustic phonon modes that extend on both sides of the interface, and phonon modes that are localized inside the semiconductor region of the interface exhibit negligible coupling with electrons in the metal. We also provide a direct comparison between simulation predictions and experimental measurements of thermal interface conductance of epitaxial CoSi2-Si interfaces using the time-domain thermoreflectance technique. Importantly, the experimental results, performed across a wide temperature range, only agree well with

  1. Numerical calculation of acoustic radiation from band-vibrating structures via FEM/FAQP method

    Directory of Open Access Journals (Sweden)

    GAO Honglin

    2017-08-01

    Full Text Available The Finite Element Method (FEM combined with the Frequency Averaged Quadratic Pressure method (FAQP are used to calculate the acoustic radiation of structures excited in the frequency band. The surface particle velocity of stiffened cylindrical shells under frequency band excitation is calculated using finite element software, the normal vibration velocity is converted from the surface particle velocity to calculate the average energy source (frequency averaged across intensity, frequency averaged across pressure and frequency averaged across velocity, and the FAQP method is used to calculate the average sound pressure level within the bandwidth. The average sound pressure levels are then compared with the bandwidth using finite element and boundary element software, and the results show that FEM combined with FAQP is more suitable for high frequencies and can be used to calculate the average sound pressure level in the 1/3 octave band with good stability, presenting an alternative to applying frequency-by-frequency calculation and the average frequency process. The FEM/FAQP method can be used as a prediction method for calculating acoustic radiation while taking the randomness of vibration at medium and high frequencies into consideration.

  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. Continuous Energy, Multi-Dimensional Transport Calculations for Problem Dependent Resonance Self-Shielding

    International Nuclear Information System (INIS)

    Downar, T.

    2009-01-01

    The overall objective of the work here has been to eliminate the approximations used in current resonance treatments by developing continuous energy multi-dimensional transport calculations for problem dependent self-shielding calculations. The work here builds on the existing resonance treatment capabilities in the ORNL SCALE code system. The overall objective of the work here has been to eliminate the approximations used in current resonance treatments by developing continuous energy multidimensional transport calculations for problem dependent self-shielding calculations. The work here builds on the existing resonance treatment capabilities in the ORNL SCALE code system. Specifically, the methods here utilize the existing continuous energy SCALE5 module, CENTRM, and the multi-dimensional discrete ordinates solver, NEWT to develop a new code, CENTRM( ) NEWT. The work here addresses specific theoretical limitations in existing CENTRM resonance treatment, as well as investigates advanced numerical and parallel computing algorithms for CENTRM and NEWT in order to reduce the computational burden. The result of the work here will be a new computer code capable of performing problem dependent self-shielding analysis for both existing and proposed GENIV fuel designs. The objective of the work was to have an immediate impact on the safety analysis of existing reactors through improvements in the calculation of fuel temperature effects, as well as on the analysis of more sophisticated GENIV/NGNP systems through improvements in the depletion/transmutation of actinides for Advanced Fuel Cycle Initiatives.

  4. 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)

  5. Calculational results for radiation embrittlement of WWER pressure vessel at the Kozloduy NPP

    Energy Technology Data Exchange (ETDEWEB)

    Apostolov, T; Ilieva, K; Petrova, T [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika

    1996-12-31

    Determination of radiation impact on metal state in the case of WWER-440/230 is made only by calculation methods since a special sample-witness (SW) incorporation had not been implemented. In WWER-1000 reactors such SW are foreseen but their spots are high above the active core. This is why in both reactors the appliance of a calculational procedure for radiation embrittlement determination is compulsory. The authors propose such a procedure accounting for the change in critical temperature of neutron brittleness by the neutron fluence. The neutron fluence and the shift of critical embrittlement temperature have been calculated for the maximum overloaded location and for the weld metal of the Kozloduy-5 and Kozloduy-6 reactors (WWER-1000). The shift of critical temperature in weld 4 of the Units 1-4 (WWER-440) is plotted versus work cycles and compared to experimental values. 4 figs., 5 tabs.

  6. Generalized Bloch Theorem for Complex Periodic Potentials - A Powerful Application to Quantum Transport Calculations

    International Nuclear Information System (INIS)

    Zhang, Xiaoguang; Varga, Kalman; Pantelides, Sokrates T

    2007-01-01

    Band-theoretic methods with periodically repeated supercells have been a powerful approach for ground-state electronic structure calculations, but have not so far been adapted for quantum transport problems with open boundary conditions. Here we introduce a generalized Bloch theorem for complex periodic potentials and use a transfer-matrix formulation to cast the transmission probability in a scattering problem with open boundary conditions in terms of the complex wave vectors of a periodic system with absorbing layers, allowing a band technique for quantum transport calculations. The accuracy and utility of the method is demonstrated by the model problems of the transmission of an electron over a square barrier and the scattering of a phonon in an inhomogeneous nanowire. Application to the resistance of a twin boundary in nanocrystalline copper yields excellent agreement with recent experimental data

  7. Calculations of Neutron Flux Distributions by Means of Integral Transport Methods

    Energy Technology Data Exchange (ETDEWEB)

    Carlvik, I

    1967-05-15

    Flux distributions have been calculated mainly in one energy group, for a number of systems representing geometries interesting for reactor calculations. Integral transport methods of two kinds were utilised, collision probabilities (CP) and the discrete method (DIT). The geometries considered comprise the three one-dimensional geometries, planes, sphericals and annular, and further a square cell with a circular fuel rod and a rod cluster cell with a circular outer boundary. For the annular cells both methods (CP and DIT) were used and the results were compared. The purpose of the work is twofold, firstly to demonstrate the versatility and efficacy of integral transport methods and secondly to serve as a guide for anybody who wants to use the methods.

  8. Calculation of radiation exposures from patients to whom radioactive materials have been administered

    Science.gov (United States)

    Cormack, John; Shearer, Jane

    1998-03-01

    Spreadsheet templates which calculate cumulative exposures to other persons from patients to whom radioactive materials have been administered have been developed by the authors. Calculations can be based on any specified single-, bi- or tri-exponential whole-body clearance rate and a diurnal (or any other periodic) contact pattern. The time (post-administration) during which close contact should be avoided in order to constrain the radiation exposure and exposure rates to selected limits is also calculated using an iterative technique (Newton's method), and the residual activity at the time when contact can resume is also calculated. These templates find particular application in the calculation of exposures to persons who are in contact with patients who have received for therapeutic purposes. The effect of changing dose limits, contact patterns and using individually derived clearance rates may be readily modelled.

  9. Measurement of secondary cosmic radiation and calculation of associated dose conversion coefficients for humans

    International Nuclear Information System (INIS)

    Simmer, Gregor

    2012-01-01

    Due to secondary cosmic radiation (SCR), pilots and flight attendants receive elevated effective doses at flight altitudes. For this reason, since 2003 aircrew members are considered as occupationally exposed, in Germany. This work deals with the calculation of dose conversion coefficients (DCC) for protons, neutrons, electrons, positrons, photons and myons, which are crucial for estimation of effective dose from SCR. For the first time, calculations were performed combining Geant4 - a Monte Carlo code developed at CERN - with the voxel phantoms for the reference female and male published in 2008 by ICRP and ICRU. Furthermore, measurements of neutron fluence spectra - which contribute the major part to the effective dose of SCR - were carried out at the Environmental Research Station Schneefernerhaus (UFS) located at 2650 m above sea level nearby the Zugspitze mountain, Germany. These measured neutron spectra, and additionally available calculated spectra, were then folded with the DCC calculated in this work, and effective dose rates for different heights were calculated.

  10. Calculation and mapping of direct and diffuse solar radiation in Costa Rica

    International Nuclear Information System (INIS)

    Wright, Jaime

    2008-01-01

    Knowledge of direct and diffuse solar radiation has been of vital importance in assessing the energy potential of Costa Rica. The work is focused on the calculation and plotting of contour maps of the direct and diffuse solar radiation, based in sixty-two radiometric stations scattered throughout the country. In tracing these contours have been used experimental and predicted values of direct and diffuse radiation. Additionally, direct and diffuse solar radiation is compared during the dry season and the rainy season in the six climatic regions of the country: Valle Central, North Pacific, Central Pacific, South Pacific, North Zone and Caribbean Region. Daily average levels of radiation observed directly have been from 6.1 and 10.1 MJ/m 2 , with higher values in the northern sections of the Pacific Slope, west of Valle Central and the tops of the highest mountains. The lowest values have coincided with the North Zone and Caribbean Region. The highest values of diffuse radiation have coincided with the North Zone and South Pacific. An increase in direct solar radiation by 40% is observed in the month of the dry season. (author) [es

  11. A set of integrated environmental transport and diffusion models for calculating hazardous releases

    International Nuclear Information System (INIS)

    Pepper, D.W.

    1996-01-01

    A set of numerical transport and dispersion models is incorporated within a graphical interface shell to predict hazardous material released into the environment. The visual shell (EnviroView) consists of an object-oriented knowledge base, which is used for inventory control, site mapping and orientation, and monitoring of materials. Graphical displays of detailed sites, building locations, floor plans, and three-dimensional views within a room are available to the user using a point and click interface. In the event of a release to the environment, the user can choose from a selection of analytical, finite element, finite volume, and boundary element methods, which calculate atmospheric transport, groundwater transport, and dispersion within a building interior. The program runs on 486 personal computers under WINDOWS

  12. Gyrokinetic Calculations of Microinstabilities and Transport During RF H-Modes on Alcator C-Mod

    International Nuclear Information System (INIS)

    Redi, M.H.; Fiore, C.; Bonoli, P.; Bourdelle, C.; Budny, R.; Dorland, W.D.; Ernst, D.; Hammett, G.; Mikkelsen, D.; Rice, J.; Wukitch, S.

    2002-01-01

    Physics understanding for the experimental improvement of particle and energy confinement is being advanced through massively parallel calculations of microturbulence for simulated plasma conditions. The ultimate goal, an experimentally validated, global, non-local, fully nonlinear calculation of plasma microturbulence is still not within reach, but extraordinary progress has been achieved in understanding microturbulence, driving forces and the plasma response in recent years. In this paper we discuss gyrokinetic simulations of plasma turbulence being carried out to examine a reproducible, H-mode, RF heated experiment on the Alcator CMOD tokamak3, which exhibits an internal transport barrier (ITB). This off axis RF case represents the early phase of a very interesting dual frequency RF experiment, which shows density control with central RF heating later in the discharge. The ITB exhibits steep, spontaneous density peaking: a reduction in particle transport occurring without a central particle source. Since the central temperature is maintained while the central density is increasing, this also suggests a thermal transport barrier exists. TRANSP analysis shows that ceff drops inside the ITB. Sawtooth heat pulse analysis also shows a localized thermal transport barrier. For this ICRF EDA H-mode, the minority resonance is at r/a * 0.5 on the high field side. There is a normal shear profile, with q monotonic

  13. Ab initio calculation of transport properties between PbSe quantum dots facets with iodide ligands

    Science.gov (United States)

    Wang, B.; Patterson, R.; Chen, W.; Zhang, Z.; Yang, J.; Huang, S.; Shrestha, S.; Conibeer, G.

    2018-01-01

    The transport properties between Lead Selenide (PbSe) quantum dots decorated with iodide ligands has been studied using density functional theory (DFT). Quantum conductance at each selected energy levels has been calculated along with total density of states and projected density of states. The DFT calculation is carried on using a grid-based planar augmented wave (GPAW) code incorporated with the linear combination of atomic orbital (LCAO) mode and Perdew Burke Ernzerhof (PBE) exchange-correlation functional. Three iodide ligand attached low index facets including (001), (011), (111) are investigated in this work. P-orbital of iodide ligand majorly contributes to density of state (DOS) at near top valence band resulting a significant quantum conductance, whereas DOS of Pb p-orbital shows minor influence. Various values of quantum conductance observed along different planes are possibly reasoned from a combined effect electrical field over topmost surface and total distance between adjacent facets. Ligands attached to (001) and (011) planes possess similar bond length whereas it is significantly shortened in (111) plane, whereas transport between (011) has an overall low value due to newly formed electric field. On the other hand, (111) plane with a net surface dipole perpendicular to surface layers leading to stronger electron coupling suggests an apparent increase of transport probability. Apart from previously mentioned, the maximum transport energy levels located several eVs (1 2 eVs) from the edge of valence band top.

  14. 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

  15. Analytic calculation of radiative-recoil corrections to muonium hyperfine splitting: Electron-line contribution

    International Nuclear Information System (INIS)

    Eides, M.I.; Karshenboim, S.G.; Shelyuto, V.A.

    1991-01-01

    The detailed account of analytic calculation of radiative-recoil correction to muonium hyperfine splitting, induced by electron-line radiative insertions, is presented. The consideration is performed in the framework of the effective two-particle formalism. A good deal of attention is paid to the problem of the divergence cancellation and the selection of graphs, relevant to radiative-recoil corrections. The analysis is greatly facilitated by use of the Fried-Yennie gauge for radiative photons. The obtained set of graphs turns out to be gauge-invariant and actual calculations are performed in the Feynman gauge. The main technical tricks, with the help of which we have effectively utilized the existence in the problem of the small parameter-mass ratio and managed to perform all calculations in the analytic form are described. The main intermediate results, as well as the final answer, δE rr = (α(Ζα)/π 2 )(m/M)E F (6ζ(3) + 3π 2 In 2 + π 2 /2 + 17/8), are also presented

  16. Energy imparted to water slabs by photons in the energy range 5-300 keV. Calculations using a Monte Carlo photon transport model

    International Nuclear Information System (INIS)

    Persliden, J.; Carlsson, G.A.

    1984-01-01

    In diagnostic examinations of the trunk and head, the energy imparted to the patient is related to the radiation risk. In this work, the energy imparted to laterally infinite, 10-300 mm thick water slabs by 5-300 keV photons is calculated using a Monte Carlo photon transport model. The energy imparted is also derived for energy spectra of primary photons relevant to diagnostic radiology. In addition to values of energy imparted, values of backscattered and transmitted energies, quantities primarily obtained in the transport calculations, are reported. Assumptions about coherent scattering are shown to be important for values of backscattered and transmitted energies but unimportant with respect to values of energy imparted. Comparisons are made with other Monte Carlo results from the literature. Discrepancies of 10-20% in some calculated quantities can be traced back to the use of different tabulations of interaction cross-sections by various authors. (author)

  17. Radiation safety assessment and development of environmental radiation monitoring technology; standardization of input parameters for the calculation of annual dose from routine releases from commercial reactor effluents

    Energy Technology Data Exchange (ETDEWEB)

    Rhee, I. H.; Cho, D.; Youn, S. H.; Kim, H. S.; Lee, S. J.; Ahn, H. K. [Soonchunhyang University, Ahsan (Korea)

    2002-04-01

    This research is to develop a standard methodology for determining the input parameters that impose a substantial impact on radiation doses of residential individuals in the vicinity of four nuclear power plants in Korea. We have selected critical nuclei, pathways and organs related to the human exposure via simulated estimation with K-DOSE 60 based on the updated ICRP-60 and sensitivity analyses. From the results, we found that 1) the critical nuclides were found to be {sup 3}H, {sup 133}Xe, {sup 60}Co for Kori plants and {sup 14}C, {sup 41}Ar for Wolsong plants. The most critical pathway was 'vegetable intake' for adults and 'milk intake' for infants. However, there was no preference in the effective organs, and 2) sensitivity analyses showed that the chemical composition in a nuclide much more influenced upon the radiation dose than any other input parameters such as food intake, radiation discharge, and transfer/concentration coefficients by more than 102 factor. The effect of transfer/concentration coefficients on the radiation dose was negligible. All input parameters showed highly estimated correlation with the radiation dose, approximated to 1.0, except for food intake in Wolsong power plant (partial correlation coefficient (PCC)=0.877). Consequently, we suggest that a prediction model or scenarios for food intake reflecting the current living trend and a formal publications including details of chemical components in the critical nuclei from each plant are needed. Also, standardized domestic values of the parameters used in the calculation must replace the values of the existed or default-set imported factors via properly designed experiments and/or modelling such as transport of liquid discharge in waters nearby the plants, exposure tests on corps and plants so on. 4 figs., 576 tabs. (Author)

  18. RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, Y.C. [Square Y Consultants, Orchard Park, NY (US); Chen, S.Y.; Biwer, B.M.; LePoire, D.J. [Argonne National Lab., IL (US)

    1995-11-01

    This report presents the technical details of RISKIND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, interactive program that can be run on an IBM or equivalent personal computer under the Windows{trademark} environment. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incident-free models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionuclide inventory and dose conversion factors. In addition, the flexibility of the models allows them to be used for assessing any accidental release involving radioactive materials. The RISKIND code allows for user-specified accident scenarios as well as receptor locations under various exposure conditions, thereby facilitating the estimation of radiological consequences and health risks for individuals. Median (50% probability) and typical worst-case (less than 5% probability of being exceeded) doses and health consequences from potential accidental releases can be calculated by constructing a cumulative dose/probability distribution curve for a complete matrix of site joint-wind-frequency data. These consequence results, together with the estimated probability of the entire spectrum of potential accidents, form a comprehensive, probabilistic risk assessment of a spent nuclear fuel transportation accident.

  19. RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

    International Nuclear Information System (INIS)

    Yuan, Y.C.; Chen, S.Y.; Biwer, B.M.; LePoire, D.J.

    1995-11-01

    This report presents the technical details of RISKIND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, interactive program that can be run on an IBM or equivalent personal computer under the Windows trademark environment. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incident-free models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionuclide inventory and dose conversion factors. In addition, the flexibility of the models allows them to be used for assessing any accidental release involving radioactive materials. The RISKIND code allows for user-specified accident scenarios as well as receptor locations under various exposure conditions, thereby facilitating the estimation of radiological consequences and health risks for individuals. Median (50% probability) and typical worst-case (less than 5% probability of being exceeded) doses and health consequences from potential accidental releases can be calculated by constructing a cumulative dose/probability distribution curve for a complete matrix of site joint-wind-frequency data. These consequence results, together with the estimated probability of the entire spectrum of potential accidents, form a comprehensive, probabilistic risk assessment of a spent nuclear fuel transportation accident

  20. Improvement of the efficiency of two-dimensional multigroup transport calculations assuming isotropic reflection with multilevel spatial discretisation

    International Nuclear Information System (INIS)

    Stankovski, Z.; Zmijarevic, I.

    1987-06-01

    This paper presents two approximations used in multigroup two-dimensional transport calculations in large, very homogeneous media: isotropic reflection together with recently proposed group-dependent spatial representations. These approximations are implemented as standard options in APOLLO 2 assembly transport code. Presented example calculations show that significant savings in computational costs are obtained while preserving the overall accuracy