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

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)

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

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.

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.

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

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

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

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.)

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

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

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.

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.

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

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.

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

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

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

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

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)

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

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)

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.

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.

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

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

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)

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.

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

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

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

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

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)

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

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

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

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

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.

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...

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)

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)

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

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.)

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

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)

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

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.

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.)

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

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.

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)

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

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

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.

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.

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

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

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)

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.

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

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

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

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.

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.)

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)

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

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.

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

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

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.

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)

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.

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

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.

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

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

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

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)

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)

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

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

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

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

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

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.

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.

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.

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

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

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

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

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

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.

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

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

Radiation transport analyses for IFMIF design by the Attila software using a Monte-Carlo source model

International Nuclear Information System (INIS)

Arter, W.; Loughlin, M.J.

2009-01-01

Accurate calculation of the neutron transport through the shielding of the IFMIF test cell, defined by CAD, is a difficult task for several reasons. The ability of the powerful deterministic radiation transport code Attila, to do this rapidly and reliably has been studied. Three models of increasing geometrical complexity were produced from the CAD using the CADfix software. A fourth model was produced to represent transport within the cell. The work also involved the conversion of the Vitenea-IEF database for high energy neutrons into a format usable by Attila, and the conversion of a particle source specified in MCNP wssaformat to a form usable by Attila. The final model encompassed the entire test cell environment, with only minor modifications. On a state-of-the-art PC, Attila took approximately 3 h to perform the calculations, as a consequence of a careful mesh 'layering'. The results strongly suggest that Attila will be a valuable tool for modelling radiation transport in IFMIF, and for similar problems

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

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.

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

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

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)

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)

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

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

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

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

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

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.

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

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.)

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.

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

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)

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

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

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

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

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

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.

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.

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.

Development of radiation dose assessment system for radiation accident (RADARAC)

International Nuclear Information System (INIS)

Takahashi, Fumiaki; Shigemori, Yuji; Seki, Akiyuki

2009-07-01

The possibility of radiation accident is very rare, but cannot be regarded as zero. Medical treatments are quite essential for a heavily exposed person in an occurrence of a radiation accident. Radiation dose distribution in a human body is useful information to carry out effectively the medical treatments. A radiation transport calculation utilizing the Monte Carlo method has an advantageous in the analysis of radiation dose inside of the body, which cannot be measured. An input file, which describes models for the accident condition and quantities of interest, should be prepared to execute the radiation transport calculation. Since the accident situation, however, cannot be prospected, many complicated procedures are needed to make effectively the input file soon after the occurrence of the accident. In addition, the calculated doses are to be given in output files, which usually include much information concerning the radiation transport calculation. Thus, Radiation Dose Assessment system for Radiation Accident (RADARAC) was developed to derive effectively radiation dose by using the MCNPX or MCNP code. RADARAC mainly consists of two parts. One part is RADARAC - INPUT, which involves three programs. A user can interactively set up necessary resources to make input files for the codes, with graphical user interfaces in a personnel computer. The input file includes information concerning the geometric structure of the radiation source and the exposed person, emission of radiations during the accident, physical quantities of interest and so on. The other part is RADARAC - DOSE, which has one program. The results of radiation doses can be effectively indicated with numerical tables, graphs and color figures visibly depicting dose distribution by using this program. These results are obtained from the outputs of the radiation transport calculations. It is confirmed that the system can effectively make input files with a few thousand lines and indicate more than 20

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.

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)

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

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

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

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

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

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

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

Energy Technology Data Exchange (ETDEWEB)

Kostin, Mikhail [Michigan State Univ., East Lansing, MI (United States); Mokhov, Nikolai [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Niita, Koji [Research Organization for Information Science and Technology, Ibaraki-ken (Japan)

2013-09-25

A parallel computing framework has been developed to use with general-purpose radiation transport codes. The framework was implemented as a C++ module that uses MPI for message passing. It is intended to be used with older radiation transport codes implemented in Fortran77, Fortran 90 or C. The module is significantly independent of radiation transport codes it can be used with, and is connected to the codes by means of a number of interface functions. The framework was developed and tested in conjunction with the MARS15 code. It is possible to use it with other codes such as PHITS, FLUKA and MCNP after certain adjustments. Besides the parallel computing functionality, the framework offers a checkpoint facility that allows restarting calculations with a saved checkpoint file. The checkpoint facility can be used in single process calculations as well as in the parallel regime. The framework corrects some of the known problems with the scheduling and load balancing found in the original implementations of the parallel computing functionality in MARS15 and PHITS. The framework can be used efficiently on homogeneous systems and networks of workstations, where the interference from the other users is possible.

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

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)

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

A hybrid transport-diffusion method for Monte Carlo radiative-transfer simulations

International Nuclear Information System (INIS)

Densmore, Jeffery D.; Urbatsch, Todd J.; Evans, Thomas M.; Buksas, Michael W.

2007-01-01

Discrete Diffusion Monte Carlo (DDMC) is a technique for increasing the efficiency of Monte Carlo particle-transport simulations in diffusive media. If standard Monte Carlo is used in such media, particle histories will consist of many small steps, resulting in a computationally expensive calculation. In DDMC, particles take discrete steps between spatial cells according to a discretized diffusion equation. Each discrete step replaces many small Monte Carlo steps, thus increasing the efficiency of the simulation. In addition, given that DDMC is based on a diffusion equation, it should produce accurate solutions if used judiciously. In practice, DDMC is combined with standard Monte Carlo to form a hybrid transport-diffusion method that can accurately simulate problems with both diffusive and non-diffusive regions. In this paper, we extend previously developed DDMC techniques in several ways that improve the accuracy and utility of DDMC for nonlinear, time-dependent, radiative-transfer calculations. The use of DDMC in these types of problems is advantageous since, due to the underlying linearizations, optically thick regions appear to be diffusive. First, we employ a diffusion equation that is discretized in space but is continuous in time. Not only is this methodology theoretically more accurate than temporally discretized DDMC techniques, but it also has the benefit that a particle's time is always known. Thus, there is no ambiguity regarding what time to assign a particle that leaves an optically thick region (where DDMC is used) and begins transporting by standard Monte Carlo in an optically thin region. Also, we treat the interface between optically thick and optically thin regions with an improved method, based on the asymptotic diffusion-limit boundary condition, that can produce accurate results regardless of the angular distribution of the incident Monte Carlo particles. Finally, we develop a technique for estimating radiation momentum deposition during the

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

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)

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.

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)

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.

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

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

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.

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)

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

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

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

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

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)

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.

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

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.

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.

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

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.

Current status of radiation transport tools for proliferation and terrorism prevention

International Nuclear Information System (INIS)

Sale, K.E.

2004-01-01

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

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

Methodology of shielding calculation for nuclear reactors

International Nuclear Information System (INIS)

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

1982-01-01

A methodology of calculation that coupling a serie of computer codes in a net that make the possibility to calculate the radiation, neutron and gamma transport, is described, for deep penetration problems, typical of nuclear reactor shielding. This net of calculation begining with the generation of constant multigroups, for neutrons and gamma, by the AMPX system, coupled to ENDF/B-IV data library, the transport calculation of these radiations by ANISN, DOT 3.5 and Morse computer codes, up to the calculation of absorbed doses and/or equivalents buy SPACETRAN code. As examples of the calculation method, results from benchmark n 0 6 of Shielding Benchmark Problems - ORNL - RSIC - 25, namely Neutron and Secondary Gamma Ray fluence transmitted through a Slab of Borated Polyethylene, are presented. (Author) [pt

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

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)

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)

Radiation inactivation studies of renal brush border water and urea transport

International Nuclear Information System (INIS)

Verkman, A.S.; Dix, J.A.; Seifter, J.L.; Skorecki, K.L.; Jung, C.Y.; Ausiello, D.A.

1985-01-01

Radiation inactivation was used to determine the nature and molecular weight of water and urea transport pathways in brush border membrane vesicles (BBMV) isolated from rabbit renal cortex. BBMV were frozen to -50 degrees C, irradiated with 1.5 MeV electrons, thawed, and assayed for transport or enzyme activity. The freezing process had no effect on enzyme or transport kinetics. BBMV alkaline phosphatase activity gave linear ln(activity) vs. radiation dose plots with a target size of 68 +/- 3 kDa, similar to previously reported values. Water and solute transport were measured using the stopped-flow light-scattering technique. The rates of acetamide and osmotic water transport did not depend on radiation dose (0-7 Mrad), suggesting that transport of these substances does not require a protein carrier. In contrast, urea and thiourea transport gave linear ln(activity) vs. dose curves with a target size of 125-150 kDa; 400 mM urea inhibited thiourea flux by -50% at 0 and 4.7 Mrad, showing that radiation does not affect inhibitor binding to surviving transporters. These studies suggest that BBMV urea transport requires a membrane protein, whereas osmotic water transport does not

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.)

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

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...

Radiation transport simulation of the Martian GCR surface flux and dose estimation using spherical geometry in PHITS compared to MSL-RAD measurements.

Science.gov (United States)

Flores-McLaughlin, John

2017-08-01

Planetary bodies and spacecraft are predominantly exposed to isotropic radiation environments that are subject to transport and interaction in various material compositions and geometries. Specifically, the Martian surface radiation environment is composed of galactic cosmic radiation, secondary particles produced by their interaction with the Martian atmosphere, albedo particles from the Martian regolith and occasional solar particle events. Despite this complex physical environment with potentially significant locational and geometric dependencies, computational resources often limit radiation environment calculations to a one-dimensional or slab geometry specification. To better account for Martian geometry, spherical volumes with respective Martian material densities are adopted in this model. This physical description is modeled with the PHITS radiation transport code and compared to a portion of measurements from the Radiation Assessment Detector of the Mars Science Laboratory. Particle spectra measured between 15 November 2015 and 15 January 2016 and PHITS model results calculated for this time period are compared. Results indicate good agreement between simulated dose rates, proton, neutron and gamma spectra. This work was originally presented at the 1st Mars Space Radiation Modeling Workshop held in 2016 in Boulder, CO. Copyright © 2017. Published by Elsevier Ltd.

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

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.

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

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

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

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

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

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.

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

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.

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

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)

Parameters calculation of a shielding experiment and evaluation of calculation methodology

International Nuclear Information System (INIS)

Gavazza, S.; Otto, A.C.; Gomes, I.C.; Maiorino, J.R.

1986-01-01

In this text is carried out the evaluation of radiation transport methodology, comparying the calculated reactions and dose rates, for neutrons and gamma-rays, with the experimental measurements obtained on iron shield, irradiated in YAYOI reactor. Were employed the ENDF/B-IV and VITAMIN-C libraries and the AMPX-II modular system for generation of cross sections, collapsed by the ANISN code. The transport calculation were made by using the DOT 3.5 code, adjusting the spectrum of the iron shield boundary source to the reactions and dose rates, measured at the beginning of shield. The distributions calculated for neutrons and gamma-rays, on iron shield, presented coherence with the experimental measurements. (Author) [pt

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

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.

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)

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)

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...

Sensitivity of a two-dimensional chemistry-transport model to changes in parameterizations of radiative processes

International Nuclear Information System (INIS)

Grant, K.E.; Ellingson, R.G.; Wuebbles, D.J.

1988-08-01

Radiative processes strongly effect equilibrium trace gas concentrations both directly, through photolysis reactions, and indirectly through temperature and transport processes. As part of our continuing radiative submodel development and validation, we have used the LLNL 2-D chemical-radiative-transport (CRT) model to investigate the net sensitivity of equilibrium ozone concentrations to several changes in radiative forcing. Doubling CO 2 from 300 ppmv to 600 ppmv resulted in a temperature decrease of 5 K to 8 K in the middle stratosphere along with an 8% to 16% increase in ozone in the same region. Replacing our usual shortwave scattering algorithms with a simplified Rayleigh algorithm led to a 1% to 2% increase in ozone in the lower stratosphere. Finally, modifying our normal CO 2 cooling rates by corrections derived from line-by-line calculations resulted in several regions of heating and cooling. We observed temperature changes on the order of 1 K to 1.5 K with corresponding changes of 0.5% to 1.5% in O 3 . Our results for doubled CO 2 compare favorably with those by other authors. Results for our two perturbation scenarios stress the need for accurately modeling radiative processes while confirming the general validity of current 2-D CRT models. 15 refs., 5 figs

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...

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.

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

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.

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)

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

C5 Benchmark Problem with Discrete Ordinate Radiation Transport Code DENOVO

Energy Technology Data Exchange (ETDEWEB)

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

2011-01-01

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

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)

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.

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)

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

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

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.

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

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.

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

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)

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)

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.

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.

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

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

Radiation Detection Computational Benchmark Scenarios

Energy Technology Data Exchange (ETDEWEB)

Shaver, Mark W.; Casella, Andrew M.; Wittman, Richard S.; McDonald, Ben S.

2013-09-24

Modeling forms an important component of radiation detection development, allowing for testing of new detector designs, evaluation of existing equipment against a wide variety of potential threat sources, and assessing operation performance of radiation detection systems. This can, however, result in large and complex scenarios which are time consuming to model. A variety of approaches to radiation transport modeling exist with complementary strengths and weaknesses for different problems. This variety of approaches, and the development of promising new tools (such as ORNL’s ADVANTG) which combine benefits of multiple approaches, illustrates the need for a means of evaluating or comparing different techniques for radiation detection problems. This report presents a set of 9 benchmark problems for comparing different types of radiation transport calculations, identifying appropriate tools for classes of problems, and testing and guiding the development of new methods. The benchmarks were drawn primarily from existing or previous calculations with a preference for scenarios which include experimental data, or otherwise have results with a high level of confidence, are non-sensitive, and represent problem sets of interest to NA-22. From a technical perspective, the benchmarks were chosen to span a range of difficulty and to include gamma transport, neutron transport, or both and represent different important physical processes and a range of sensitivity to angular or energy fidelity. Following benchmark identification, existing information about geometry, measurements, and previous calculations were assembled. Monte Carlo results (MCNP decks) were reviewed or created and re-run in order to attain accurate computational times and to verify agreement with experimental data, when present. Benchmark information was then conveyed to ORNL in order to guide testing and development of hybrid calculations. The results of those ADVANTG calculations were then sent to PNNL for

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.

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

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

Experimental validation of GADRAS's coupled neutron-photon inverse radiation transport solver

International Nuclear Information System (INIS)

Mattingly, John K.; Mitchell, Dean James; Harding, Lee T.

2010-01-01

Sandia National Laboratories has developed an inverse radiation transport solver that applies nonlinear regression to coupled neutron-photon deterministic transport models. The inverse solver uses nonlinear regression to fit a radiation transport model to gamma spectrometry and neutron multiplicity counting measurements. The subject of this paper is the experimental validation of that solver. This paper describes a series of experiments conducted with a 4.5 kg sphere of α-phase, weapons-grade plutonium. The source was measured bare and reflected by high-density polyethylene (HDPE) spherical shells with total thicknesses between 1.27 and 15.24 cm. Neutron and photon emissions from the source were measured using three instruments: a gross neutron counter, a portable neutron multiplicity counter, and a high-resolution gamma spectrometer. These measurements were used as input to the inverse radiation transport solver to evaluate the solver's ability to correctly infer the configuration of the source from its measured radiation signatures.

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

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

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

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

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.

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.

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...

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.

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.

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

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

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

International Nuclear Information System (INIS)

Yamano, Naoki; Minami, Kazuyoshi; Koyama, Kinji; Naito, Yoshitaka.

1989-03-01

A modular code system RADHEAT-V4 has been developed for performing precisely neutron and photon transport analyses, and shielding safety evaluations. The system consists of the functional modules for producing coupled multi-group neutron and photon cross section sets, for analyzing the neutron and photon transport, and for calculating the atom displacement and the energy deposition due to radiations in nuclear reactor or shielding material. A precise method named Direct Angular Representation (DAR) has been developed for eliminating an error associated with the method of the finite Legendre expansion in evaluating angular distributions of cross sections and radiation fluxes. The DAR method implemented in the code system has been described in detail. To evaluate the accuracy and applicability of the code system, some test calculations on strong anisotropy problems have been performed. From the results, it has been concluded that RADHEAT-V4 is successfully applicable to evaluating shielding problems accurately for fission and fusion reactors and radiation sources. The method employed in the code system is very effective in eliminating negative values and oscillations of angular fluxes in a medium having an anisotropic source or strong streaming. Definitions of the input data required in various options of the code system and the sample problems are also presented. (author)

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.)

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

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)

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

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

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.

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.

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)

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

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.)

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.

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)

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

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.

Calculations of radiation fields and monkey mid-head and mid-thorax responses in AFRRI-TRIGA reactor facility experiments

International Nuclear Information System (INIS)

Johnson, J.O.; Emmett, M.B.; Pace, J.V. III.

1983-07-01

A computational study was performed to characterize the radiation exposure fields and the mid-head and mid-thorax response functions for monkeys irradiated in the Armed Forces Radiobiological Research Institute (AFRRI) reactor exposure facilities. Discrete ordinates radiation transport calculations were performed in one-dimensional spherical geometry to obtain the energy spectra of the neutrons and gamma rays entering the room through various spectrum modifiers and reaching the irradiation position. Adjoint calculations performed in two-dimensional cylindrical geometry yielded the mid-head and mid-thorax response functions, which were then folded with flux spectra to obtain the monkey mid-head and mid-thorax doses (kerma rates) received at the irradiation position. The results of the study are presented both as graphs and as tables. The resulting spectral shapes compared favorably with previous work; however, the magnitudes of the fluxes did not. The differences in the magnitudes may be due to the normalization factor used

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.

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.)

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

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

Radiological emergency: road map for radiation accident victim transport; Emergência radiológica: roadmap para o transporte de radioacidentado

Energy Technology Data Exchange (ETDEWEB)

Costa, V.S.G.; Alcantara, Y.P. [Faculdade Casa Branca, SP (Brazil); Lima, C.M.A. [MAXIM Cursos, Rio de Janeiro, RJ (Brazil); Silva, F. C. A. da, E-mail: franciscodasilva13uk@gmail.com [Instituto de Radioproteção e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

2017-07-01

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.

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

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)

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

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

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

Nuclear medicine radiation dosimetry

CERN Document Server

McParland, Brian J

2010-01-01

Complexities of the requirements for accurate radiation dosimetry evaluation in both diagnostic and therapeutic nuclear medicine (including PET) have grown over the past decade. This is due primarily to four factors: growing consideration of accurate patient-specific treatment planning for radionuclide therapy as a means of improving the therapeutic benefit, development of more realistic anthropomorphic phantoms and their use in estimating radiation transport and dosimetry in patients, design and use of advanced Monte Carlo algorithms in calculating the above-mentioned radiation transport and

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)

Cross sections needed for investigations into track phenomena and Monte-Carlo calculations

International Nuclear Information System (INIS)

Paretzke, H.G.

1983-01-01

Investigations into basic radiation action mechanisms as well as into applied radiation transport problems (e.g. electron microscopy) greatly benefit from detailed computer simulations of charged particle track structures in matter. The first and in fact most important and most difficult step in any such calculation is the derivation of reliable cross sections for the most relevant interaction processes in the material(s) under consideration. The second step in radiation transport calculations is the testing of results or intermediate results for quantitative or qualitative consistency with other experimental or theoretical information (e.g. yields, backscatter factors). This paper discusses the types of the most important collision cross sections for studies on track phenomena by detailed Monte-Carlo calculations, the necessary accuracy of such data and various means of consistency checks of calculated results. This will be done mainly with examples taken from radiation physics as applied to dosimetric and biological problems (i.e. to gaseous and condensed targets). 12 references, 8 figures

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.

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

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

Theoretical calculations of electron-impact and radiative processes in atoms

International Nuclear Information System (INIS)

Pindzola, M.S.

1975-01-01

Electron-impact and radiative processes in atoms are investigated with particular attention paid to the effects of electron correlations. Using the optical potential method, the cross section for the elastic scattering of electrons by the neutral argon atom is calculated from 0 to 300 eV. Corrections to the Hartree--Fock cross section are obtained from a many-particle perturbation expansion. The effects of electron correlations are found to be quite significant at low energy. The optical potential results are compared with a polarized orbital calculation, the Born approximation and experiment. The 2s and 2p excitation cross sections for electron scattering on hydrogen are calculated by two similar methods. The distorted wave method is applied and the effect of calculating the outgoing scattered electron in the potential of the initial or final state is investigated. The imaginary part of the optical potential is also calculated in lowest order by the use of many-body diagrams. The subshell photoionization cross sections in argon are calculated using the acceleration, length and velocity forms of the dipole operator. First order electron correlation corrections to the Hartree--Fock approximation are obtained through the use of many-body perturbation theory. Also investigated is the two photon ionization cross section for the neutral argon atom. A double perturbation expansion in the Coulomb correlations and the atom-radiation field interaction is made. Contributions from intermediate states are obtained by direct summation over Hartree--Fock bound and continuum single particle states. The effects of electron correlations and photon radiative corrections are investigated

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

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

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.

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

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.

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)

Calculating Quenching Weights

CERN Document Server

Salgado, C A; Salgado, Carlos A.; Wiedemann, Urs Achim

2003-01-01

We calculate the probability (``quenching weight'') that a hard parton radiates an additional energy fraction due to scattering in spatially extended QCD matter. This study is based on an exact treatment of finite in-medium path length, it includes the case of a dynamically expanding medium, and it extends to the angular dependence of the medium-induced gluon radiation pattern. All calculations are done in the multiple soft scattering approximation (Baier-Dokshitzer-Mueller-Peign\\'e-Schiff--Zakharov ``BDMPS-Z''-formalism) and in the single hard scattering approximation (N=1 opacity approximation). By comparison, we establish a simple relation between transport coefficient, Debye screening mass and opacity, for which both approximations lead to comparable results. Together with this paper, a CPU-inexpensive numerical subroutine for calculating quenching weights is provided electronically. To illustrate its applications, we discuss the suppression of hadronic transverse momentum spectra in nucleus-nucleus colli...

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

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.)

Radiation risk assessment for the transport of radioisotopes using KRI-BGM B(U) type container

International Nuclear Information System (INIS)

Cho, Woon-Kap

2008-01-01

The radiation risks were estimated for the transportation of radioisotopes using KRI-BGM transport container. KRI-BGM container was specially designed for transportation of large amount of radioisotopes for industrial or medical applications. The container can carry maximum 370 TBq of solid Ir-192, 29.6 TBq of liquid Mo-99 and 37 TBq of liquid I-131 respectively. For the radiation risk assessment, it was assumed that maximum design activity of those radioisotopes was transported. Transportation route is from Daejeon where radioisotopes are produced to Seoul where radioisotopes are consumed. Transport distance is 200 km including highway and downtown area from Daejeon to Seoul. As the transportation conveyance, an ordinary cargo truck is used exclusively. Radiation risks were estimated for incident free and accident condition of transportation and RADTRAN 5.6 was used as the risk assessment tool. For the risk assessment of radioisotopes transportation, various parameters such as population density around transport route, weather condition, probability of specific accidents such as impact, fire, etc. were considered. From the results of this study, the exclusive transportation of radioisotopes using KRI-BGM transport container by truck showed low radiological risks with manageable safety and health consequences. This paper discusses the methods and results of the radiation risks assessment for the radioisotopes transportation by an ordinary truck and presents the expected radiation risks in person-Sv and latent cancer fatalities. (author)

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.

Integral and Lagrangian simulations of particle and radiation transport in plasma

International Nuclear Information System (INIS)

Christlieb, A J; Hitchon, W N G; Lawler, J E; Lister, G G

2009-01-01

Accurate integral and Lagrangian models of transport in plasmas, in which the models reflect the actual physical behaviour as closely as possible, are presented. These methods are applied to the behaviour of particles and photons in plasmas. First, to show how these types of models arise in a wide range of plasma physics applications, an application to radiation transport in a lighting discharge is given. The radiation transport is solved self-consistently with a model of the discharge to provide what are believed to be very accurate 1D simulations of fluorescent lamps. To extend these integral methods to higher dimensions is computationally very costly. The wide utility of 'treecodes' in solving massive integral problems in plasma physics is discussed, and illustrated in modelling vortex formation in a Penning trap, where a remarkably detailed simulation of vortex formation in the trap is obtained. Extension of treecode methods to other integral problems such as radiation transport is under consideration.

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)

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

Conceptual designs of radioactive canister transporters

International Nuclear Information System (INIS)

1978-02-01

This report covers conceptual designs of transporters for the vertical, horizontal, and inclined installation of canisters containing spent-fuel elements, high-level waste, cladding waste, and intermediate-level waste (low-level waste is not discussed). Included in the discussion are cask concepts; transporter vehicle designs; concepts for mechanisms for handling and manipulating casks, canisters, and concrete plugs; transporter and repository operating cycles; shielding calculations; operator radiation dosages; radiation-resistant materials; and criteria for future design efforts

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.

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

Parameters calculation of shielding experiment

International Nuclear Information System (INIS)

Gavazza, S.

1986-02-01

The radiation transport methodology comparing the calculated reactions and dose rates for neutrons and gama-rays, with experimental measurements obtained on iron shield, irradiated in the YAYOI reactor is evaluated. The ENDF/B-IV and VITAMIN-C libraries and the AMPX-II modular system, for cross sections generation collapsed by the ANISN code were used. The transport calculations were made using the DOT 3.5 code, adjusting the boundary iron shield source spectrum to the reactions and dose rates, measured at the beginning of shield. The neutron and gamma ray distributions calculated on the iron shield presented reasonable agreement with experimental measurements. An experimental arrangement using the IEA-R1 reactor to determine a shielding benchmark is proposed. (Author) [pt

VU-B radiation inhibits the photosynthetic electron transport chain in chlamydomonas reinhardtii

International Nuclear Information System (INIS)

Cai, W.; Li, X.; Chen, L.

2016-01-01

UV radiation of sunlight is one of harmful factors for earth organisms, especially for photoautotrophs because they require light for energy and biomass production. A number of works have already been done regarding the effects of UV-B radiation at biochemical and molecular level, which showed that UV-B radiation could inhibit photosynthesis activity and reduce photosynthetic electron transport. However quite limited information can accurately make out inhibition site of UV-B radiation on photosynthetic electron transport. In this study, this issue was investigated through measuring oxygen evolution activity, chlorophyll a fluorescence and gene expression in a model unicellular green alga Chlamydomonas reinhardtii. Our results indicated that UV-B radiation could evidently decrease photosynthesis activity and inhibit electron transport by blocking electron transfer process from the first plastoquinone electron acceptors QA to second plastoquinone electron acceptors QB, but not impair electron transfer from the water oxidizing complex to QA. The psbA gene expression was also altered by UV-B radiation, where up-regulation occurred at 2, 4 and 6h after exposure and down-regulation happened at 12 and 24 h after exposure. These results suggested that UV-B could affects D1 protein normal turnover, so there was not enough D1 for binding with QB, which may affect photosynthetic electron transport and photosynthesis activity. (author)

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)

An evaluation of calculation parameters in the EGSnrc/BEAMnrc Monte Carlo codes and their effect on surface dose calculation

International Nuclear Information System (INIS)

Kim, Jung-Ha; Hill, Robin; Kuncic, Zdenka

2012-01-01

The Monte Carlo (MC) method has proven invaluable for radiation transport simulations to accurately determine radiation doses and is widely considered a reliable computational measure that can substitute a physical experiment where direct measurements are not possible or feasible. In the EGSnrc/BEAMnrc MC codes, there are several user-specified parameters and customized transport algorithms, which may affect the calculation results. In order to fully utilize the MC methods available in these codes, it is essential to understand all these options and to use them appropriately. In this study, the effects of the electron transport algorithms in EGSnrc/BEAMnrc, which are often a trade-off between calculation accuracy and efficiency, were investigated in the buildup region of a homogeneous water phantom and also in a heterogeneous phantom using the DOSRZnrc user code. The algorithms and parameters investigated include: boundary crossing algorithm (BCA), skin depth, electron step algorithm (ESA), global electron cutoff energy (ECUT) and electron production cutoff energy (AE). The variations in calculated buildup doses were found to be larger than 10% for different user-specified transport parameters. We found that using BCA = EXACT gave the best results in terms of accuracy and efficiency in calculating buildup doses using DOSRZnrc. In addition, using the ESA = PRESTA-I option was found to be the best way of reducing the total calculation time without losing accuracy in the results at high energies (few keV ∼ MeV). We also found that although choosing a higher ECUT/AE value in the beam modelling can dramatically improve computation efficiency, there is a significant trade-off in surface dose uncertainty. Our study demonstrates that a careful choice of user-specified transport parameters is required when conducting similar MC calculations. (note)

Process and appliance for determining the nature of transported substances by measuring the radiations transmitted

International Nuclear Information System (INIS)

Wykes, J.S.; Surzyn, P.M.; Croke, G.M.; Adsley, Ian.

1980-01-01

The invention relates to a process for determining the nature of a substance transported, comprising the collimation of the radiation of not less than two energies so that they form beams; the irradiation of the matter transported by the beams, the detection of the non-scattered radiations for the two energies at least, after passing in the transported matter, and the deduction of the nature of the transported matter according to the radiations detected. The radiations are collimated by placing a shield around the gamma source (americium 241 or cesium 137). The detector is protected by a shield so that it prevents any significant interference due to the reactions near the surface provoked by those radiations not of the lowest energy, with detection of those radiations of lesser energy. In a variation, a source of relatively higher energy radiations is placed at a distance from the source of relatively lesser energy radiations. The latter have a component taken from natural ionizing radiation and this component is withdrawn to a predetermined calibration [fr

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.

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

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.

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)

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

Radiation exposures of workers and the public associated with the transport of radioactive material in Germany

International Nuclear Information System (INIS)

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

2004-01-01

Most radioactive material packages transported emit penetrating ionising radiation and radiation exposures of transport workers and the public may occur during their transport. The radiation exposures incurred by transport workers and members of the public can vary significantly depending on a number of factors: most important is the type of radiation emitted (primarily gamma and neutron radiation), the radiation field intensity in the surrounding of a package and conveyance and the duration of exposure to ionising radiation. The information and guidance material on occupational exposures has primarily been derived from a survey and analysis of personal monitoring data provided by a number of commercial transport operators in Germany known as major carrier and handler organisations of fuel cycle and non-fuel cycle material (in terms of the number of pack-ages and the activity carriaged). To some extent advantage was taken of compilations of statistical transport and exposure data collated within other transport safety analysis studies including research projects funded by the European Commission. The exposure data collected cover the time period of the last 4 - 8 years and are most representative for routine transport operations closely related to the movement phase of packaged radioactive material, i.e. receipt, vehicle loading, carriage, in-transit storage, intra-/intermodal transfer, vehicle unloading and delivery at the final destination of loads of radioactive material and packages and the related supervisory and health physics functions. Radiation dose monitoring of members of the public, however, is generally impracticable and, consequently, the information available relies on employing dose assessment models and reflects radiation exposures incurred by hypothetical or critical group individuals of members of the public under normal conditions of transport

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

International Nuclear Information System (INIS)

Koyama, Kinji; Taji, Yukichi; Miyasaka, Shun-ichi; Minami, Kazuyoshi.

1977-07-01

The modular code system RADHEAT is for producing coupled multigroup neutron and gamma-ray cross section sets, analyzing the neutron and gamma-ray transport, and calculating the energy deposition and atomic displacements due to these radiations in a nuclear reactor or shield. The basic neutron cross sections and secondary gamma-ray production data are taken from ENDF/B and POPOP4 libraries respectively. The system (1) generates multigroup neutron cross sections, energy deposition coefficients and atomic displacement factors due to neutron reactions, (2) generates multigroup gamma-ray cross sections and energy transfer coefficients, (3) generates secondary gamma-ray production cross sections, (4) combines these cross sections into the coupled set, (5) outputs and updates the multigroup cross section libraries in convenient formats for other transport codes, (6) analyzes the neutron and gamma-ray transport and calculates the energy deposition and the number density of atomic displacements in a medium, (7) collapses the cross sections to a broad-group structure, by option, using the weighting functions obtained by one-dimensional transport calculation, and (8) plots, by option, multigroup cross sections, and neutron and gamma-ray distributions. Definitions of the input data required in various options of the code system are also given. (auth.)

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)

Germany: Exposure of Transport Workers During the Transport of Most Frequently Transported NORM in Germany

International Nuclear Information System (INIS)

2013-01-01

The German national report to this CRP was focused on the following services according to the research agreement: (1) Status review, analysis and evaluation of the radiation exposure imposed by shipment and expected exposure of the shipment staff of the most relevant NORM in Germany; (2) Development of evaluation criteria and safety requirements to provide adequate safety standards for the transportation of NORM; (3) Development and application of procedures to determine the limits for exempt materials/consignments for transportation according to German Transport Regulations for all NORM. For the analysis and evaluation of the radiation exposure imposed by shipment of NORM for the following materials, a couple of transport scenarios were defined and the dose to transport workers was calculated. - Tantalum raw materials; - Raw phosphate; - Pipe scales and sludge from oil and gas exploitation; - Coal ash; - Waste rock material from uranium mining; - Zircon raw materials; - Titanium dioxide raw materials; - Filter gravel from waterworks

EMSH program for calculating electron and photon transport through a matter at energies of 10 keV-1TeV

International Nuclear Information System (INIS)

Tayurskij, V.A.

1989-01-01

The EMSH program (Electro-Magnetic Shower) for the calculation of 1 keV-1 TeV electron and photon transport through a substance is described. The program is written in FORTRAN for the ES computers. Electron and positron bremsstrahlung, e - e - - and e + e - scattering, positron annihilation, production of e + e - -pairs by photons, photon. Compton scattering, photoelectric effect, photon Rayleigh scattering are taken into account during calculations. The cross sections of these processes are assigned with the 3-10% accuracy. Energy losses for atom ionization and excitation and bremsstrahlung, as well as multiple scattering are taken into account for charged particles. The program is used to calculate electron conversion into positrons, to estimate accelerator radiation background, to simulate electromagnetic showers in electromagntic calorimeters. 37 refs.; 15 figs

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

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

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

International Nuclear Information System (INIS)

1982-01-01

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

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.

Radiation shielding for fusion reactors

International Nuclear Information System (INIS)

Santoro, R.T.

2000-01-01

Radiation shielding requirements for fusion reactors present different problems than those for fission reactors and accelerators. Fusion devices, particularly tokamak reactors, are complicated by geometry constraints that complicate disposition of fully effective shielding. This paper reviews some of these shielding issues and suggested solutions for optimizing the machine and biological shielding. Radiation transport calculations are essential for predicting and confirming the nuclear performance of the reactor and, as such, must be an essential part of the reactor design process. Development and optimization of reactor components from the first wall and primary shielding to the penetrations and containment shielding must be carried out in a sensible progression. Initial results from one-dimensional transport calculations are used for scoping studies and are followed by detailed two- and three-dimensional analyses to effectively characterize the overall radiation environment. These detail model calculations are essential for accounting for the radiation leakage through ports and other penetrations in the bulk shield. Careful analysis of component activation and radiation damage is cardinal for defining remote handling requirements, in-situ replacement of components, and personnel access at specific locations inside the reactor containment vessel. (author)

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)

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)

Radiation exposures of workers resulting from the transport of gamma radiography sources in Germany

International Nuclear Information System (INIS)

Sentuc, F.N.; Schwarz, G.

2006-01-01

Gamma radiation sources are widely used for industrial purposes e.g. for non-destructive material testing. Many of these sources are permanently installed at a facility within instruments e.g. for level or thickness gauging. Other radioactive sources are implemented in portable devices for industrial gamma radiography which have to be carried to the various remote usage sites. In Germany, approximately 20 000 - 25 000 shipments of gamma radiography sources are proceeding annually on public transport routes. Since routine radiation monitoring programmes do not permit task-specific determination of occupational doses e.g. doses incurred during the movement phase and handling related doses, work has been carried out with the objective to determine the radiation exposures of the personnel attributable to transportation. For this purpose, a survey was launched in 2005 collecting data about e.g. the number and conditions of transports, the activity and type of transported radiation sources and the radiation level within the driver's cab to allow a dose assessment to be made for transport workers. The results of this survey covering the most important companies for gamma radiography services in Germany are presented in this paper. (authors)

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.

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

Science.gov (United States)

Longoni, Gianluca; Anderson, Stanwood L.

2009-08-01

The numerical solution of the Linearized Boltzmann Equation (LBE) via the Discrete Ordinates method (SN) requires extensive computational resources for large 3-D neutron and gamma transport applications due to the concurrent discretization of the angular, spatial, and energy domains. This paper will discuss the development RAPTOR-M3G (RApid Parallel Transport Of Radiation - Multiple 3D Geometries), a new 3-D parallel radiation transport code, and its application to the calculation of ex-vessel neutron dosimetry responses in the cavity of a commercial 2-loop Pressurized Water Reactor (PWR). RAPTOR-M3G is based domain decomposition algorithms, where the spatial and angular domains are allocated and processed on multi-processor computer architectures. As compared to traditional single-processor applications, this approach reduces the computational load as well as the memory requirement per processor, yielding an efficient solution methodology for large 3-D problems. Measured neutron dosimetry responses in the reactor cavity air gap will be compared to the RAPTOR-M3G predictions. This paper is organized as follows: Section 1 discusses the RAPTOR-M3G methodology; Section 2 describes the 2-loop PWR model and the numerical results obtained. Section 3 addresses the parallel performance of the code, and Section 4 concludes this paper with final remarks and future work.

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Assessment of radiation maps during activated divertor moving in the ITER building

International Nuclear Information System (INIS)

Ying Dongchuan; Zeng Qin; Qiu Yuefeng; Dang Tongqiang; Wu Yican; Loughlin, Michael

2011-01-01

As the main interface components between plasma and vacuum vessel, the divertor is foreseen to be removed to the hot cell for refurbishment during the 20 years of ITER operation. During this process, the activated divertor will cause a large increase of radiation in the ITER building. 3D analysis has been performed to assess the radiation maps throughout the ITER building for assisting the shielding design for personnel and sensitive equipment. The activation of the divertor has been determined by coupled neutron transport and inventory calculations, radiation maps have been obtained from gamma transport calculations. The neutron and gamma transport calculations have been performed by MCNP5 code with FENDL2.1library. The inventory calculations have been performed by FISPACT2007 code with EAF-2007 library. The results of these 3D decay gamma radiation maps are presented by pictures in this paper, including the biological dose maps and maps of heat deposition in electronic equipment.

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

International Nuclear Information System (INIS)

Pölz, Stefan; Laubersheimer, Sven; Eberhardt, Jakob S; Harrendorf, Marco A; Keck, Thomas; Benzler, Andreas; Breustedt, Bastian

2013-01-01

The basic idea of Voxel2MCNP is to provide a framework supporting users in modeling radiation transport scenarios using voxel phantoms and other geometric models, generating corresponding input for the Monte Carlo code MCNPX, and evaluating simulation output. Applications at Karlsruhe Institute of Technology are primarily whole and partial body counter calibration and calculation of dose conversion coefficients. A new generic data model describing data related to radiation transport, including phantom and detector geometries and their properties, sources, tallies and materials, has been developed. It is modular and generally independent of the targeted Monte Carlo code. The data model has been implemented as an XML-based file format to facilitate data exchange, and integrated with Voxel2MCNP to provide a common interface for modeling, visualization, and evaluation of data. Also, extensions to allow compatibility with several file formats, such as ENSDF for nuclear structure properties and radioactive decay data, SimpleGeo for solid geometry modeling, ImageJ for voxel lattices, and MCNPX’s MCTAL for simulation results have been added. The framework is presented and discussed in this paper and example workflows for body counter calibration and calculation of dose conversion coefficients is given to illustrate its application. (paper)

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)

Study on shielding design method of radiation streaming in a tokamak-type DT fusion reactor based on Monte Carlo calculation

International Nuclear Information System (INIS)

Sato, Satoshi

2003-09-01

three dimensional Monte Carlo calculation is required for the shielding calculation in the tokamak-type DT nuclear fusion reactor with many penetrations. 2) In Chapter 3, radiation streaming through the slit between the blanket modules is described, in Chapter 4, that through the small circular duct in the blanket modules is described, in Chapter 5, and that through the large opening duct in the vacuum vessel is described. The nuclear properties of the blanket, the vacuum vessel and the TF coil are systematically calculated for the various configurations. Based on the obtained results, the analytical formulas of these nuclear properties are deduced, and the guideline is proposed for the shielding design. 3) In Chapter 6, in order to evaluate the decay gamma ray dose rate around the duct due to radiation streaming through the large opening duct in the vacuum vessel, the evaluation method is proposed using the decay gamma ray Monte Carlo calculation. By replacing the prompt gamma-ray spectrum to the decay one in the Monte Carlo code, the decay gamma ray Monte Carlo transport calculation is conducted. The effective variance reduction method is developed for the decay gamma ray Monte Carlo calculation in the over-all tokamak region with drastically reducing the calculation time. Using this method, the shielding calculation is conducted for the ITER duct penetration, and the effectiveness of this method is demonstrated. (author)

A comparison of radiation doses and risks between spent fuel transport/storage and selected non-nuclear activities

International Nuclear Information System (INIS)

Pennington, C.W.

2003-01-01

Spent fuel transport and storage have achieved an exemplary safety record over four decades within both the United States (US) and the global community at large. This paper offers an assessment demonstrating the safety of spent fuel transport and storage packagings relative to currently accepted but unregulated non-nuclear activities and practices within society. Over the last quarter of a century, several spent fuel transport and storage packaging test programmes have produced data that allow calculation of potential releases and population doses resulting from a terrorist attack. The US Department of Energy (DOE) has used this information to develop projected worst-case, low probability population exposures as part of the Final Environmental Impact Statement (FEIS) for the Yucca Mountain repository. The paper discusses potential population exposures from these packagings based on analysis and testing under beyond-design-basis (BDB) events, including missile attacks, and then defines and defends an acceptance criterion for the bounding outcomes of these events, based upon current accepted activities within society that produce high radiation doses to the general public. These activities involve unregulated technologies and practices within society that yield population doses significantly exceeding those that would result from such hypothetical and highly improbable events as a terrorist missile attack on a spent fuel transport or storage packaging. In particular, technologically enhanced natural radiation (TENR) exposures from building materials, farming, and masonry construction are highlighted. Recent landmark work by the US National Academy of Sciences (NAS) and by the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) are cited in support of this assessment, along with work from the US Environmental Protection Agency (EPA). From this compelling evidence, it is concluded that spent fuel transport and storage represent a low

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.

Dose rate calculations for a reconnaissance vehicle

International Nuclear Information System (INIS)

Grindrod, L.; Mackey, J.; Salmon, M.; Smith, C.; Wall, S.

2005-01-01

A Chemical Nuclear Reconnaissance System (CNRS) has been developed by the British Ministry of Defence to make chemical and radiation measurements on contaminated terrain using appropriate sensors and recording equipment installed in a land rover. A research programme is under way to develop and validate a predictive capability to calculate the build-up of contamination on the vehicle, radiation detector performance and dose rates to the occupants of the vehicle. This paper describes the geometric model of the vehicle and the methodology used for calculations of detector response. Calculated dose rates obtained using the MCBEND Monte Carlo radiation transport computer code in adjoint mode are presented. These address the transient response of the detectors as the vehicle passes through a contaminated area. Calculated dose rates were found to agree with the measured data to be within the experimental uncertainties, thus giving confidence in the shielding model of the vehicle and its application to other scenarios. (authors)

Parallel unstructured mesh optimisation for 3D radiation transport and fluids modelling

International Nuclear Information System (INIS)

Gorman, G.J.; Pain, Ch. C.; Oliveira, C.R.E. de; Umpleby, A.P.; Goddard, A.J.H.

2003-01-01

In this paper we describe the theory and application of a parallel mesh optimisation procedure to obtain self-adapting finite element solutions on unstructured tetrahedral grids. The optimisation procedure adapts the tetrahedral mesh to the solution of a radiation transport or fluid flow problem without sacrificing the integrity of the boundary (geometry), or internal boundaries (regions) of the domain. The objective is to obtain a mesh which has both a uniform interpolation error in any direction and the element shapes are of good quality. This is accomplished with use of a non-Euclidean (anisotropic) metric which is related to the Hessian of the solution field. Appropriate scaling of the metric enables the resolution of multi-scale phenomena as encountered in transient incompressible fluids and multigroup transport calculations. The resulting metric is used to calculate element size and shape quality. The mesh optimisation method is based on a series of mesh connectivity and node position searches of the landscape defining mesh quality which is gauged by a functional. The mesh modification thus fits the solution field(s) in an optimal manner. The parallel mesh optimisation/adaptivity procedure presented in this paper is of general applicability. We illustrate this by applying it to a transient CFD (computational fluid dynamics) problem. Incompressible flow past a cylinder at moderate Reynolds numbers is modelled to demonstrate that the mesh can follow transient flow features. (authors)

Radiation induced low-energy electron transport in a tissue environment

International Nuclear Information System (INIS)

Toburen, L.H.; Dingfelder, M.; Ozturk, N.; Christou, C.; Shinpaugh, J.L.; Friedland, W.; Wilson, W.E.; Paretzke, H.G.

2003-01-01

Monte Carlo (MC) track simulation codes are used extensively in radiobiology to quantify the spatial distributions of interactions initiated by the absorption of ionizing radiation. The spatial patterns of ionization and excitation are instrumental for assessing the formation of damage clusters in DNA and chromosomes leading to such biologic endpoints as cellular transformation and mutation. The MC codes rely on an extensive database of elastic and inelastic scattering cross sections to follow the production and slowing of secondary electrons. Because of inherent uncertainties in this database we are exploring the sensitivity of MC results to the details of the cross sections used with emphasis on low-energy electrons, i.e., track ends, that are anticipated to play a dominant role in damage cluster formation. Simulations of electron transport using gas or liquid based interaction cross sections illustrate substantial difference in the spectra of electrons with energies less than about 50 eV. In addition, the electron yields from MC simulations appear to be nearly a factor of five larger than our recent measurements of electron transport spectra in water (ice) at electron energies of about 10 eV. Examples of the changes in electron transport spectra for variations in the electron scattering cross sections used for the MC calculations will be illustrated and compared with an evolving database of measured spectra of electrons from ion induced secondary electron transport in thin foils. These measurements provide guidance for assessment of elastic and elastic cross sections appropriate to condensed phase transport. This work is supported in part by the U.S. Department of Energy, Grant No. DE-FG02-01ER-63233; the National Cancer Institute, Grant No. 1R01CA93351-01A1; and the European Community under Contract No. FIGH-CT-1999-00005

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

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...

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.

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)

Three-dimensional self-consistent radiation transport model for the fluid simulation of plasma display panel cell

International Nuclear Information System (INIS)

Kim, H.C.; Yang, S.S.; Lee, J.K.

2003-01-01

In plasma display panels (PDPs), the resonance radiation trapping is one of the important processes. In order to incorporate this effect in a PDP cell, a three-dimensional radiation transport model is self-consistently coupled with a fluid simulation. This model is compared with the conventional trapping factor method in gas mixtures of neon and xenon. It shows the differences in the time evolutions of spatial profile and the total number of resonant excited states, especially in the afterglow. The generation rates of UV light are also compared for the two methods. The visible photon flux reaching the output window from the phosphor layers as well as the total UV photon flux arriving at the phosphor layer from the plasma region are calculated for resonant and nonresonant excited species. From these calculations, the time-averaged spatial profiles of the UV flux on the phosphor layers and the visible photon flux through the output window are obtained. Finally, the diagram of the energy efficiency and the contribution of each UV light are shown

Development of 2-D/1-D fusion method for three-dimensional whole-core heterogeneous neutron transport calculations

International Nuclear Information System (INIS)

Lee, Gil Soo

2006-02-01

To describe power distribution and multiplication factor of a reactor core accurately, it is necessary to perform calculations based on neutron transport equation considering heterogeneous geometry and scattering angles. These calculations require very heavy calculations and were nearly impossible with computers of old days. From the limitation of computing power, traditional approach of reactor core design consists of heterogeneous transport calculation in fuel assembly level and whole core diffusion nodal calculation with assembly homogenized properties, resulting from fuel assembly transport calculation. This approach may be effective in computation time, but it gives less accurate results for highly heterogeneous problems. As potential for whole core heterogeneous transport calculation became more feasible owing to rapid development of computing power during last several years, the interests in two and three dimensional whole core heterogeneous transport calculations by deterministic method are increased. For two dimensional calculation, there were several successful approaches using even parity transport equation with triangular meshes, S N method with refined rectangular meshes, the method of characteristics (MOC) with unstructured meshes, and so on. The work in this thesis originally started from the two dimensional whole core heterogeneous transport calculation by using MOC. After successful achievement in two dimensional calculation, there were efforts in three-dimensional whole-core heterogeneous transport calculation using MOC. Since direct extension to three dimensional calculation of MOC requires too much computing power, indirect approach to three dimensional calculation was considered.Thus, 2D/1D fusion method for three dimensional heterogeneous transport calculation was developed and successfully implemented in a computer code. The 2D/1D fusion method is synergistic combination of the MOC for radial 2-D calculation and S N -like methods for axial 1

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

International Nuclear Information System (INIS)

Roussin, R.W.

1993-01-01

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

Education and Training in Radiation, Transport and Waste Safety Newsletter, No. 2c, May 2013

International Nuclear Information System (INIS)

2013-05-01

The IAEA Strategic Approach to Education and Training in Radiation, Transport and Waste Safety (2011-2020) provides a framework for establishing a sustainable education and training infrastructure in Member States that addresses national needs for building and maintaining competence in radiation, transport and waste safety that is consistent with IAEA Safety Standards. For this purpose, IAEA's General Conference has encouraged Member States to develop a national strategy for education and training, underlining the fundamental importance of sustainable programmes for building competence in radiation, transport and waste safety, as a key component of safety infrastructure. Furthermore Member States that receive assistance from IAEA are obliged to apply IAEA Safety Standards which require, inter alia, governments to establish a national policy and strategy for safety, including provisions for acquiring and maintaining the necessary competence nationally for ensuring safety. IAEA's Division of Radiation, Transport and Waste Safety is assisting Member States to develop their own national strategies in Europe via the Regional Project RER/9/109 on ''Strengthening Education and training Infrastructure, and Building Competence in Radiation Safety'', which includes, inter alia, Regional Workshops on National Strategies for education and training in radiation transport and waste safety. IAEA's Regional Training Centres (RTCs) in Greece and Belarus are key partners in the European region.

Education and Training in Radiation, Transport and Waste Safety Newsletter, No. 2d, June 2013

International Nuclear Information System (INIS)

2013-06-01

The IAEA Strategic Approach to Education and Training in Radiation, Transport and Waste Safety (2011-2020) provides a framework for establishing a sustainable education and training infrastructure in Member States that addresses national needs for building and maintaining competence in radiation, transport and waste safety that is consistent with IAEA Safety Standards. For this purpose, IAEA's General Conference has encouraged Member States to develop a national strategy for education and training, underlining the fundamental importance of sustainable programmes for building competence in radiation, transport and waste safety, as a key component of safety infrastructure. Furthermore Member States that receive assistance from IAEA are obliged to apply IAEA Safety Standards which require, inter alia, governments to establish a national policy and strategy for safety, including provisions for acquiring and maintaining the necessary competence nationally for ensuring safety. IAEA's Division of Radiation, Transport and Waste Safety is assisting Member States to develop their own national strategies in Latin America via the Regional Project RLA/9/070 on ''Strengthening Education and training Infrastructure, and Building Competence in Radiation Safety'', which includes, inter alia, Regional Workshops on National Strategies for education and training in radiation transport and waste safety. IAEA's Regional Training Centres (RTCs) in Argentina and Brazil are key partners in the Latin-American region.

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.

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

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

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)

Power and Particle Balance Calculations with Impurities in NSTX

Science.gov (United States)

Holland, C. G.; Maingi, R.; Owen, L. W.; Kaye, S. M.

1998-11-01

We reported the development C. Holland, et. al., Bull. Am. Phys. Soc. 42 (1997) 1927. and application R. Maingi et al., Proc. 3rd International Workshop on Spherical Tori, Sept. 3-5, 1997, St. Petersburg, Russia. of a Graphical User Interface to assess the important terms for edge and divertor plasma calculations for NSTX with the b2.5 edge plasma transport code B. Braams, Contrib. Plasma Phys. 36 (1996) 276.. The goals of those calculations were to estimate the worst case peak heat flux for plasma-facing component design, and the radiation requirements to reduce the peak heat flux. In this study we present the first simulations with intrinsic carbon impurity radiation. We find in general that the intrinsic carbon radiation should be sufficient to provide a wide operation window for the NSTX device. Details of the relative importance of heat flux transport mechanisms as determined with the GUI will be presented.

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)

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

Radiation dose evaluation for hypothetical accident with transport package containing Iridium-192 source

International Nuclear Information System (INIS)

Trontl, K.; Bace, M.; Pevec, D.

2002-01-01

The aim of this paper is to evaluate dose rates for a hypothetical accident with transport package containing Iridium-192 source and to design additional shielding necessary for the safe unloading of the container, assuming that during the unloading process the whole contents of a radioactive source is unshielded and that the operation is going to take place at the site where a working area exists in the vicinity of the unloading location. Based on the calculated radiation dose rates, a single arrangement of the additional concrete shields necessary for reduction of the gamma dose rates to the permitted level is proposed. The proposed solution is optimal considering safety on one hand and costs on the other.(author)

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

Time-dependent Flow and Transport Calculations for Project Opalinus Clay (Entsorgungsnachweis)

International Nuclear Information System (INIS)

Kosakowski, G.

2004-07-01

This report describes two specific assessment cases used in the safety assessment for a proposed deep geological repository for spent fuel, high level waste and long-lived intermediate-level waste, sited in the Opalinus Clay of the Zuercher Weinland in northern Switzerland (Project Entsorgungsnachweis, NAG RA, 2002d). In this study the influence of time dependent flow processes on the radionuclide transport in the geosphere is investigated. In the Opalinus Clay diffusion dominates the transport of radionuclides, but processes exist that can locally increase the importance of the advective transport for some time. Two important cases were investigated: (1) glaciation-induced flow due to an additional overburden in the form of an ice shield of up to 400 m thickness and (2) fluid flow driven by tunnel convergence. For the calculations the code FRAC3DVS (Therrien and Sudicky, 1996) was used. FRAC3DVS solves the three-dimensional flow and transport equation in porous and fractured media. For the case of glaciation-induced flow (1) a two-dimensional reference model without glaciations was calculated. During the glaciations the geosphere release-rates are up to a factor of about 1.7 higher compared to the reference model. The influence of glaciations on the transport of cations or neutral species is less than for anions, since the importance of the advective transport for anions is higher due to the lower accessible porosity for anions. The increase in the release rates during glaciations is lower for sorbing compared to non-sorbing radionuclides. The influence of the tunnel convergence (2) on the transport of radionuclides in the geosphere is very small. Due to the higher source term the geosphere release rates are slightly higher if tunnel convergence is considered. In addition to the two assessment cases this report investigates the applicability of the one-dimensional approximation for modelling transport through the Opalinus Clay. For the reference case of the safety

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)

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.

An anisotropic diffusion approximation to thermal radiative transfer

International Nuclear Information System (INIS)

Johnson, Seth R.; Larsen, Edward W.

2011-01-01

This paper describes an anisotropic diffusion (AD) method that uses transport-calculated AD coefficients to efficiently and accurately solve the thermal radiative transfer (TRT) equations. By assuming weak gradients and angular moments in the radiation intensity, we derive an expression for the radiation energy density that depends on a non-local function of the opacity. This nonlocal function is the solution of a transport equation that can be solved with a single steady-state transport sweep once per time step, and the function's second angular moment is the anisotropic diffusion tensor. To demonstrate the AD method's efficacy, we model radiation flow down a channel in 'flatland' geometry. (author)

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

Impurity transport in internal transport barrier discharges on JET

International Nuclear Information System (INIS)

Dux, R.

2002-01-01

In JET plasmas with internal transport barrier (ITB) the behaviour of metallic and low-Z impurities (C, Ne) was investigated. In ITB discharges with reversed shear, the metallic impurities accumulate in cases with too strong peaking of the density profile, while the concentration of low-Z elements C and Ne is only mildly peaked. The accumulation might be so strong, that the central radiation approximately equals the central heating power followed by a radiative collapse of the transport barrier. The radial location with strong impurity gradients (convective barrier) was identified to be situated inside (not at!) the heat flux barrier. Calculations of neo-classical transport were performed for these discharges, including impurity-impurity collisions. It was found, that the observed Z-dependence of the impurity peaking and the location of the impurity 'barrier' can be explained with neo-classical transport. ITB discharges with monotonic shear show less inward convection and seem to be advantageous with respect to plasma purity. (author)

Unconventional application of the two-flux approximation for the calculation of the Ambartsumyan-Chandrasekhar function and the angular spectrum of the backward-scattered radiation for a semi-infinite isotropically scattering medium

Science.gov (United States)

Remizovich, V. S.

2010-06-01

It is commonly accepted that the Schwarzschild-Schuster two-flux approximation (1905, 1914) can be employed only for the calculation of the energy characteristics of the radiation field (energy density and energy flux density) and cannot be used to characterize the angular distribution of radiation field. However, such an inference is not valid. In several cases, one can calculate the radiation intensity inside matter and the reflected radiation with the aid of this simplest approximation in the transport theory. In this work, we use the results of the simplest one-parameter variant of the two-flux approximation to calculate the angular distribution (reflection function) of the radiation reflected by a semi-infinite isotropically scattering dissipative medium when a relatively broad beam is incident on the medium at an arbitrary angle relative to the surface. We do not employ the invariance principle and demonstrate that the reflection function exhibits the multiplicative property. It can be represented as a product of three functions: the reflection function corresponding to the single scattering and two identical h functions, which have the same physical meaning as the Ambartsumyan-Chandrasekhar function ( H) has. This circumstance allows a relatively easy derivation of simple analytical expressions for the H function, total reflectance, and reflection function. We can easily determine the relative contribution of the true single scattering in the photon backscattering at an arbitrary probability of photon survival Λ. We compare all of the parameters of the backscattered radiation with the data resulting from the calculations using the exact theory of Ambartsumyan, Chandrasekhar, et al., which was developed decades after the two-flux approximation. Thus, we avoid the application of fine mathematical methods (the Wiener-Hopf method, the Case method of singular functions, etc.) and obtain simple analytical expressions for the parameters of the scattered radiation

Design analysis of various transportation package options for BN-350 SNF in terms of nuclear radiation safety in planning for long-terms dry storage

International Nuclear Information System (INIS)

Aisabekov, A.Z.; Mukenova, S.A.; Tur, E.S.; Tsyngaev, V.M.

2004-01-01

Full text: This effort is performed under the BN-350 reactor facility decommissioning project. One of the project tasks - spent nuclear fuel handling - includes the following: fuel packaging into sealed canisters, transportation of the canisters in multi-seat metallo-concrete containers and placement of the containers for a long-term dry storage. The goal of this effort is to computationally validate nuclear and radiation safety of the SNF containers placed for storage both under normal storage conditions and probable accident situations. The basic unit structure and design configurations are presented: assemblies, canisters, transportation containers. The major factors influencing nuclear and radiation safety are presented: fuel burn-up, enrichment, fabrication tolerance, types of fuel assemblies, configuration of assemblies in the canister and canisters in the container, background of assemblies placed in the reactor and cooling pool. Conditions under which the SNF containers will be stored are described and probable accident situations are listed. Proceeding from the conservatism principle, selection of the assemblies posing the greatest nuclear hazard is validated. A neutron effective multiplication factor is calculated for the SNF containers under the normal storage conditions and for the case of emergency. The effective multiplication factor is shown to be within a standard value of 0.95 in any situation. Based on the experimental data on assembly and canister dose rates, canisters posing the highest radiation threat are selected. Activities of sources and gamma-radiation spectral composition are calculated. Distribution of the dose rate outside the containers both under the normal storage conditions and accident situations are calculated. The results obtained are analyzed

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)

Radiation shielding quality assurance

Science.gov (United States)

Um, Dallsun

For the radiation shielding quality assurance, the validity and reliability of the neutron transport code MCNP, which is now one of the most widely used radiation shielding analysis codes, were checked with lot of benchmark experiments. And also as a practical example, follows were performed in this thesis. One integral neutron transport experiment to measure the effect of neutron streaming in iron and void was performed with Dog-Legged Void Assembly in Knolls Atomic Power Laboratory in 1991. Neutron flux was measured six different places with the methane detectors and a BF-3 detector. The main purpose of the measurements was to provide benchmark against which various neutron transport calculation tools could be compared. Those data were used in verification of Monte Carlo Neutron & Photon Transport Code, MCNP, with the modeling for that. Experimental results and calculation results were compared in both ways, as the total integrated value of neutron fluxes along neutron energy range from 10 KeV to 2 MeV and as the neutron spectrum along with neutron energy range. Both results are well matched with the statistical error +/-20%. MCNP results were also compared with those of TORT, a three dimensional discrete ordinates code which was developed by Oak Ridge National Laboratory. MCNP results are superior to the TORT results at all detector places except one. This means that MCNP is proved as a very powerful tool for the analysis of neutron transport through iron & air and further it could be used as a powerful tool for the radiation shielding analysis. For one application of the analysis of variance (ANOVA) to neutron and gamma transport problems, uncertainties for the calculated values of critical K were evaluated as in the ANOVA on statistical data.

Status of shielding analysis methods for transport packages

International Nuclear Information System (INIS)

Parks, C.V.; Broadhead, B.L.; Brady, M.C.

1991-01-01

Shielding analysis methods for transport packages are becoming more important to the cask designer because optimized cask designs with higher payloads can yield doses near the limits set by regulatory authorities. Uncertainty arising from generation of radiation sources, selection of cross-section data, and the radiation transport methodology must be considered. Recent comparison studies using popular US codes illustrate calculational discrepancies arising from each of these areas

Radiation doses from radiation sources of neutrons and photons by different computer calculation; Tecniche di calcolo di intensita` di dose da sorgenti di radiazione neutronica e fotonica con l`uso di codici basati su metodologie diverse

Energy Technology Data Exchange (ETDEWEB)

Siciliano, F.; Lippolis, G.; Bruno, S.G. [ENEA, Centro Ricerche Trisaia, Rotondella (Italy)

1995-11-01

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 theoretic background are concerned.

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

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.

Calculation and mapping of gamma radiation field in the pool of Apsara reactor

International Nuclear Information System (INIS)

Singh, Tej; Singh, Kanchhi; Sharma, ARchana; Somakumar, K.; Raina, V.K.; Srinivasan, P.; Prasad, S.K.; Babu, D.A.R.; Sharma, D.N.

2007-12-01

Theoretical simulation of the radiation transport occurring in the Apsara core and bulk shield was carried out using two different radiation transport codes, MCNP and QADCG. The MCNP is a Monte Carlo based statistical method solving Boltzmann transport equation, where as the latter code QADCG is a point kernel based deterministic method with build-up factor correction. The aim of the simulation was to do a dose mapping and estimate the expected value of gamma dose rates at various locations where experimental measurements were conducted. Details regarding the simulation techniques employed by both the MCNP and QADCG software with reference to the Apsara core and shield geometry and source gamma energy distribution in the fuel plates are presented in this report. Different types of particle tallies requested in MCNP and QADCG are discussed. Details of variance reduction methods employed in reducing the statistical uncertainty of Monte Carlo simulation are also mentioned in the report. The statistical errors associated with Monte Carlo based simulation varied between 3% - 6% in most of the energy bins that contribute to the total fluence and hence to the dose rates. It was observed that the experimental values and the theoretically simulated values match each other closely following a similar trend except for certain experimental locations which had photon flux contributions from extraneous sources like the N-16 activity present in water, beam tubes and pool liner towards shielding corner. It is seen that the theoretical values are found to be larger than experimental values by factors ranging from 1.1 to 3 depending on the water shield thickness. This study served in validation of the experimental measurements conducted by GM counter based teletector and dipole based detectors. In addition, the comparison provided a confirmation of the accuracy of the radiation transport simulation techniques used for dose rate evaluation in case of complex source geometries and

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

Implicit Monte Carlo methods and non-equilibrium Marshak wave radiative transport

International Nuclear Information System (INIS)

Lynch, J.E.

1985-01-01

Two enhancements to the Fleck implicit Monte Carlo method for radiative transport are described, for use in transparent and opaque media respectively. The first introduces a spectral mean cross section, which applies to pseudoscattering in transparent regions with a high frequency incident spectrum. The second provides a simple Monte Carlo random walk method for opaque regions, without the need for a supplementary diffusion equation formulation. A time-dependent transport Marshak wave problem of radiative transfer, in which a non-equilibrium condition exists between the radiation and material energy fields, is then solved. These results are compared to published benchmark solutions and to new discrete ordinate S-N results, for both spatially integrated radiation-material energies versus time and to new spatially dependent temperature profiles. Multigroup opacities, which are independent of both temperature and frequency, are used in addition to a material specific heat which is proportional to the cube of the temperature. 7 refs., 4 figs

Dose calculation for electrons

International Nuclear Information System (INIS)

Hirayama, Hideo

1995-01-01

The joint working group of ICRP/ICRU is advancing the works of reviewing the ICRP publication 51 by investigating the data related to radiation protection. In order to introduce the 1990 recommendation, it has been demanded to carry out calculation for neutrons, photons and electrons. As for electrons, EURADOS WG4 (Numerical Dosimetry) rearranged the data to be calculated at the meeting held in PTB Braunschweig in June, 1992, and the question and request were presented by Dr. J.L. Chartier, the responsible person, to the researchers who are likely to undertake electron transport Monte Carlo calculation. The author also has carried out the requested calculation as it was the good chance to do the mutual comparison among various computation codes regarding electron transport calculation. The content that the WG requested to calculate was the absorbed dose at depth d mm when parallel electron beam enters at angle α into flat plate phantoms of PMMA, water and ICRU4-element tissue, which were placed in vacuum. The calculation was carried out by the versatile electron-photon shower computation Monte Carlo code, EGS4. As the results, depth dose curves and the dependence of absorbed dose on electron energy, incident angle and material are reported. The subjects to be investigated are pointed out. (K.I.)

A short history of nomograms and tables used for thermal radiation calculations

Science.gov (United States)

Stewart, Seán. M.; Johnson, R. Barry

2016-09-01

The theoretical concept of a perfect thermal radiator, the blackbody, was first introduced by the German physicist Gustav Robert Kirchhoff in 1860. By the latter half of the nineteenth century it had become the object of intense theoretical and experimental investigation. While an attempt at trying to theoretically understand the behavior of radiation emitted from a blackbody was undertaken by many eminent physicists of the day, its solution was not found until 1900 when Max Planck put forward his now famous law for thermal radiation. Today, of course, understanding blackbody behavior is vitally important to many fields including infrared systems, illumination, pyrometry, spectroscopy, astronomy, thermal engineering, cryogenics, and meteorology. Mathematically, the form Planck's law takes is rather cumbersome meaning calculations made with it before the advent of modern computers were rather tedious, dramatically slowing the process of computation. Fortunately, during those early days of the twentieth century researchers quickly realized Planck's equation, and the various functions closely related to it, readily lend themselves to being given a graphical, mechanical, or numerically tabulated form for their evaluation. The first of these computational aids to appear were tables. These arose shortly after Planck introduced his equation, were produced in the greatest number, and remained unsurpassed in their level of accuracy compared to all other aids made. It was also not long before nomograms designed to aid thermal radiation calculations appeared. Essentially a printed chart and requiring nothing more than a straightedge to use, nomograms were cheap and extremely easy to use. Facilitating instant answers to a range of quantities relating to thermal radiation, a number were produced and the inventiveness displayed in some was quite remarkable. In this paper we consider the historical development of many of the nomograms and tables developed and used by generations

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.

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

Effects of body habitus on internal radiation dose calculations using the 5-year-old anthropomorphic male models

DEFF Research Database (Denmark)

Xie, Tianwu; Kuster, Niels; Zaidi, Habib

2017-01-01

Xtended general purpose Monte Carlo transport code and calculated the absorbed dose and effective dose of five 18F-labelled radiotracers for children of various habitus. For most organs, the S-value of F-18 presents stronger statistical correlations with body weight, standing height and sitting height than BMI...... and SSR. The self-absorbed fraction and self-absorbed S-values of F-18 and the absorbed dose and effective dose of 18F-labelled radiotracers present with the strongest statistical correlations with body weight. For 18F-Amino acids, 18F-Brain receptor substances, 18F-FDG, 18F-L-DOPA and 18F-FBPA, the mean...... absolute effective dose differences between phantoms of different habitus and fixed reference models are 11.4%, 11.3%, 10.8%, 13.3% and 11.4%, respectively. Total body weight, standing height and sitting height have considerable effects on human internal dosimetry. Radiation dose calculations...

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

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.

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

Self-consistent collisional-radiative model for hydrogen atoms: Atom–atom interaction and radiation transport

International Nuclear Information System (INIS)

Colonna, G.; Pietanza, L.D.; D’Ammando, G.

2012-01-01

Graphical abstract: Self-consistent coupling between radiation, state-to-state kinetics, electron kinetics and fluid dynamics. Highlight: ► A CR model of shock-wave in hydrogen plasma has been presented. ► All equations have been coupled self-consistently. ► Non-equilibrium electron and level distributions are obtained. ► The results show non-local effects and non-equilibrium radiation. - Abstract: A collisional-radiative model for hydrogen atom, coupled self-consistently with the Boltzmann equation for free electrons, has been applied to model a shock tube. The kinetic model has been completed considering atom–atom collisions and the vibrational kinetics of the ground state of hydrogen molecules. The atomic level kinetics has been also coupled with a radiative transport equation to determine the effective adsorption and emission coefficients and non-local energy transfer.

Deterministic methods in radiation transport. A compilation of papers presented February 4--5, 1992

Energy Technology Data Exchange (ETDEWEB)

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

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.

Deterministic methods in radiation transport. A compilation of papers presented February 4-5, 1992

Energy Technology Data Exchange (ETDEWEB)

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

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.

Education and Training in Radiation, Transport and Waste Safety Newsletter, No. 2b, May 2013

International Nuclear Information System (INIS)

2013-05-01

The IAEA Strategic Approach to Education and Training in Radiation, Transport and Waste Safety (2011-2020) provides a framework for establishing a sustainable education and training infrastructure in Member States that addresses national needs for building and maintaining competence in radiation, transport and waste safety that is consistent with IAEA Safety Standards. For this purpose, IAEA's General Conference has encouraged Member States to develop a national strategy for education and training, underlining the fundamental importance of sustainable programmes for building competence in radiation, transport and waste safety, as a key component of safety infrastructure. Furthermore Member States that receive assistance from IAEA are obliged to apply IAEA Safety Standards which require, inter alia, governments to establish a national policy and strategy for safety, including provisions for acquiring and maintaining the necessary competence nationally for ensuring safety. IAEA's Division of Radiation, Transport and Waste Safety is assisting Member States to develop their own national strategies in Asia and the Pacific via the Regional project RAS/9/066 on ''Strengthening Education and training Infrastructure, and Building Competence in Radiation Safety'', which includes, inter alia, Regional Workshops on National Strategies for education and training in radiation transport and waste safety. IAEA's Regional Training Centres (RTCs) in Malaysia and Syrian Arabic Republic are key partners in the Asian and the Pacific region.

Education and Training in Radiation, Transport and Waste Safety Newsletter, No. 2a, May 2013

International Nuclear Information System (INIS)

2013-05-01

The IAEA Strategic Approach to Education and Training in Radiation, Transport and Waste Safety (2011-2020) provides a framework for establishing a sustainable education and training infrastructure in Member States that addresses national needs for building and maintaining competence in radiation, transport and waste safety that is consistent with IAEA Safety Standards. For this purpose, IAEA's General Conference has encouraged Member States to develop a national strategy for education and training, underlining the fundamental importance of sustainable programmes for building competence in radiation, transport and waste safety, as a key component of safety infrastructure. Furthermore Member States that receive assistance from IAEA are obliged to apply IAEA Safety Standards which require, inter alia, governments to establish a national policy and strategy for safety, including provisions for acquiring and maintaining the necessary competence nationally for ensuring safety. IAEA's Division of Radiation, Transport and Waste Safety is assisting Member States to develop their own national strategies in Africa via the Regional project RAF/9/04 on ''Strengthening Education and training Infrastructure, and Building Competence in Radiation Safety'', which includes, inter alia, Regional Workshops on National Strategies for education and training in radiation transport and waste safety. AFRA Regional Designated Centres, in Algeria, Ghana and Morocco, equivalent to the IAEA's Regional Training Centres (RTCs) present in all the other regions, are key partners in the African region.

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.

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

Science.gov (United States)

Chapoutier, Nicolas; Mollier, François; Nolin, Guillaume; Culioli, Matthieu; Mace, Jean-Reynald

2017-09-01

In the context of the rising of Monte Carlo transport calculations for any kind of application, AREVA recently improved its suite of engineering tools in order to produce efficient Monte Carlo workflow. Monte Carlo codes, such as MCNP or TRIPOLI, are recognized as reference codes to deal with a large range of radiation transport problems. However the inherent drawbacks of theses codes - laboring input file creation and long computation time - contrast with the maturity of the treatment of the physical phenomena. The goals of the recent AREVA developments were to reach similar efficiency as other mature engineering sciences such as finite elements analyses (e.g. structural or fluid dynamics). Among the main objectives, the creation of a graphical user interface offering CAD tools for geometry creation and other graphical features dedicated to the radiation field (source definition, tally definition) has been reached. The computations times are drastically reduced compared to few years ago thanks to the use of massive parallel runs, and above all, the implementation of hybrid variance reduction technics. From now engineering teams are capable to deliver much more prompt support to any nuclear projects dealing with reactors or fuel cycle facilities from conceptual phase to decommissioning.

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

Gray and multigroup radiation transport models for two-dimensional binary stochastic media using effective opacities

International Nuclear Information System (INIS)

Olson, Gordon L.

2016-01-01

One-dimensional models for the transport of radiation through binary stochastic media do not work in multi-dimensions. Authors have attempted to modify or extend the 1D models to work in multidimensions without success. Analytic one-dimensional models are successful in 1D only when assuming greatly simplified physics. State of the art theories for stochastic media radiation transport do not address multi-dimensions and temperature-dependent physics coefficients. Here, the concept of effective opacities and effective heat capacities is found to well represent the ensemble averaged transport solutions in cases with gray or multigroup temperature-dependent opacities and constant or temperature-dependent heat capacities. In every case analyzed here, effective physics coefficients fit the transport solutions over a useful range of parameter space. The transport equation is solved with the spherical harmonics method with angle orders of n=1 and 5. Although the details depend on what order of solution is used, the general results are similar, independent of angular order. - Highlights: • Gray and multigroup radiation transport is done through 2D stochastic media. • Approximate models for the mean radiation field are found for all test problems. • Effective opacities are adjusted to fit the means of stochastic media transport. • Test problems include temperature dependent opacities and heat capacities • Transport solutions are done with angle orders n=1 and 5.

Small-sized linear accelerator of 2.5 MeV electrons with a local radiation shield for custom examination of freight transported by motor transport

International Nuclear Information System (INIS)

Baklanov, A.V.; Gavrish, Yu.N.; Klinov, A.P.; Krest'yaninov, A.S.; Nikolaev, V.M.; Fomin, L.P.; Linkenbach, H.A.; Geus, G.; Knospel, W.

2001-01-01

A new development of a small-sized linear accelerator of 2.5 MeV electrons with a local radiation protection is described. The accelerator is intended for movable facilities of radiation custom of the freight transported by motor transport. Main constructive solutions, mass and dimension characteristics and results of preliminary tests of the accelerator parameters and characteristics of radiation protection are presented [ru

Development of Neutron and Photon Shielding Calculation System for Workstation (NPSS-W)

International Nuclear Information System (INIS)

Shimizu, Yoshio; Nojiri, Ichiro; Odajima, Akira; Sasaki, Toshihisa; Kurosawa, Naohiro

1998-01-01

In plant designs and safety evaluations of nuclear fuel cycle facilities, it is important to evaluate the direct radiation and the skyshine (air-scattered photon radiation) from facilities reasonably. The Neutron and Photon Shielding Calculation System for Workstation (NPSS-W) was developed. The NPSS-W can carry out the shielding calculations of the photon and the neutron easily and rapidly. The NPSS-W can easily calculate the radiation source intensity by ORIGEN-S and the dose equivalent rate by SN transport calculational codes, which are ANISN and DOT3.5. The NPSS-W consists of five modules, which named CAL1, CAL2, CAL3, CAL4, CAL5). Some kinds of shielding calculational systems are calculated. The user's manual of NPSS-W, the examples of calculations for each module and the output data are appended. (author)

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

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.

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

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)

Simulation of radiation transport using MCNP for a teletherapy machine; Simulacion del transporte de radiacion usando MCNP para una maquina de teleterapia

Energy Technology Data Exchange (ETDEWEB)

Flores O, F.E.; Mireles G, F.; Davila R, J.I.; Pinedo V, J.L.; Risorios M, C.; Lopez del Rio, H. [UAZ, Unidad Academica de Estudios Nucleares, 98068 Zacatecas (Mexico)

2008-07-01

The MCNP code is used to simulate the radiation transport taking as tools the transport physics of each particle, either photon, neutron or electron, and the generation of random numbers. Developed in the Los Alamos National Laboratory, this code has been used thoroughly with great success, because the results of the simulations are broadly validated with representative experiments. In the one present work the room of radiotherapy of the Institute Zacatecano of the Tumor it is simulated, located in the city of Zacatecas where one is Theratron 780C machine manufactured by MSD Nordion, with the purpose of estimating the contribution to the dose that would be received in different points of the structure, included three directly under the source. Three results of analytical calculations for points located at different distances from the source are presented, and they are compared against those obtained by the simulation. Its are also presented results for the simulation of 10 points more distributed around the source. (Author)

XSDRNPM-S biasing of MORSE-SGC/S shipping-cask calculations

International Nuclear Information System (INIS)

Hoffman, T.J.; Tang, J.S.

1982-06-01

This report describes implementation of a systematic approach for biasing a Monte Carlo radiation transport calculation. In particular, the adjoint fluxes from a one-dimensional discrete ordinates calculation with the XSDRNPM-S code are used to generate biasing parameters for the multigroup Monte Carlo code, MORSE-SGC/S. Application of this biasing procedure to several deep penetration spent fuel shipping cask problems is also reported. The results obtained for neutron and gamma-ray transport indicate that relatively inexpensive Monte Carlo calculations are possible for dry and water filled shipping cask problems using these procedures. 5 tables

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

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)

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

Application of the Monte Carlo technique to the study of radiation transport in a prompt gamma in vivo neutron activation system

International Nuclear Information System (INIS)

Chan, A.A.; Beddoe, A.H.

1985-01-01

A Monte Carlo code (MORSE-SGC) from the Radiation Shielding Information Centre at Oak Ridge National Laboratory, USA, has been adapted and used to model radiation transport in the Auckland prompt gamma in vivo neutron activation analysis facility. Preliminary results are presented for the slow neutron flux in an anthropomorphic phantom which are in broad agreement with those obtained by measurement via activation foils. Since experimental optimization is not logistically feasible and since theoretical optimization of neutron activation facilities has not previously been attempted, it is hoped that the Monte Carlo calculations can be used to provide a basis for improved system design

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)

A functional method for estimating DPA tallies in Monte Carlo calculations of Light Water Reactors

International Nuclear Information System (INIS)

Read, Edward A.; Oliveira, Cassiano R.E. de

2011-01-01

There has been a growing need in recent years for the development of methodology to calculate radiation damage factors, namely displacements per atom (dpa), of structural components for Light Water Reactors (LWRs). The aim of this paper is to discuss the development and implementation of a dpa method using Monte Carlo method for transport calculations. The capabilities of the Monte Carlo code Serpent such as Woodcock tracking and fuel depletion are assessed for radiation damage calculations and its capability demonstrated and compared to those of the Monte Carlo code MCNP for radiation damage calculations of a typical LWR configuration. (author)

Hybrid Monte-Carlo method for ICF calculations

International Nuclear Information System (INIS)

Clouet, J.F.; Samba, G.

2003-01-01

) conduction and ray-tracing for laser description. Radiation transport is usually solved by a Monte-Carlo method. In coupling diffusion approximation and transport description, the difficult part comes from the need for an implicit discretization of the emission-absorption terms: this problem was solved by using the symbolic Monte-Carlo method. This means that at each step of the simulation a matrix is computed by a Monte-Carlo method which accounts for the radiation energy exchange between the cells. Because of time step limitation by hydrodynamic motion, energy exchange is limited to a small number of cells and the matrix remains sparse. This matrix is added to usual diffusion matrix for thermal and radiative conductions: finally we arrive at a non-symmetric linear system to invert. A generalized Marshak condition describe the coupling between transport and diffusion. In this paper we will present the principles of the method and numerical simulation of an ICF hohlraum. We shall illustrate the benefits of the method by comparing the results with full implicit Monte-Carlo calculations. In particular we shall show how the spectral cut-off evolves during the propagation of the radiative front in the gold wall. Several issues are still to be addressed (robust algorithm for spectral cut- off calculation, coupling with ALE capabilities): we shall briefly discuss these problems. (authors)

Independent verification of monitor unit calculation for radiation treatment planning system.

Science.gov (United States)

Chen, Li; Chen, Li-Xin; Huang, Shao-Min; Sun, Wen-Zhao; Sun, Hong-Qiang; Deng, Xiao-Wu

2010-02-01

To ensure the accuracy of dose calculation for radiation treatment plans is an important part of quality assurance (QA) procedures for radiotherapy. This study evaluated the Monitor Units (MU) calculation accuracy of a third-party QA software and a 3-dimensional treatment planning system (3D TPS), to investigate the feasibility and reliability of independent verification for radiation treatment planning. Test plans in a homogenous phantom were designed with 3-D TPS, according to the International Atomic Energy Agency (IAEA) Technical Report No. 430, including open, blocked, wedge, and multileaf collimator (MLC) fields. Test plans were delivered and measured in the phantom. The delivered doses were input to the QA software and the independent calculated MUs were compared with delivery. All test plans were verified with independent calculation and phantom measurements separately, and the differences of the two kinds of verification were then compared. The deviation of the independent calculation to the measurements was (0.1 +/- 0.9)%, the biggest difference fell onto the plans that used block and wedge fields (2.0%). The mean MU difference between the TPS and the QA software was (0.6 +/- 1.0)%, ranging from -0.8% to 2.8%. The deviation in dose of the TPS calculation compared to the measurements was (-0.2 +/- 1.7)%, ranging from -3.9% to 2.9%. MU accuracy of the third-party QA software is clinically acceptable. Similar results were achieved with the independent calculations and the phantom measurements for all test plans. The tested independent calculation software can be used as an efficient tool for TPS plan verification.

Directions in Radiation Transport Modelling